[Federal Register Volume 61, Number 119 (Wednesday, June 19, 1996)]
[Notices]
[Pages 31171-31192]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10619]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from May 24, 1996, through June 7, 1996. The last
biweekly notice was published on June 5, 1996 (61 FR 28604).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
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 July 19, 1996, 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
[[Page 31172]]
petition for leave to intervene shall be filed in accordance with the
Commission's ``Rules of Practice for Domestic Licensing Proceedings''
in 10 CFR Part 2. Interested persons should consult a current copy of
10 CFR 2.714 which is available at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC and at
the local public document room for the particular facility involved. If
a request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: 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. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
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.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: April 25, 1996
Description of amendment request: The proposed amendment would
change the definition of Operable-Operability, revise Technical
Specifications (TSs) and associated Bases Section for TSs 3.5.F.1,
``Core and Containment Cooling systems,'' TSs 3.9.B.1, 3.9.B.2,
3.9.B.3, 3.9.b.4, ``Auxiliary Electrical System,'' and TSs 3.7.B.1.a,
c, and e, and 3.7.b.2.a, c, and e, ``Standby Gas Treatment System and
Control Room High Efficiency Air Filtration System,'' and delete TSs
4.5.F.1, ``Core and Containment Cooling Systems,'' and 3.7.B.1.f,
``Standby Gas Treatment System and Control Room High Efficiency Air
Filtration 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 amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Operation of PNPS [Pilgrim Nuclear Power Station] in accordance
with the proposed license amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated because of the following:
Definition of ``Operable-Operability''
Definitions perform a supporting function for other sections of
the TS. The definition of ``Operable-Operability'' affects the
manner
[[Page 31173]]
in which the requirements for a Limiting Condition for Operation
(LCO) and its associated remedial actions are applied when a support
system is inoperable. This definition re-affirms the principle that
a system is operable when it is capable of performing its specified
function and when all necessary support systems are also capable of
performing their related support functions. The corollary is that a
system is inoperable when it is not capable of performing its
specified function or when a necessary support system is not capable
of performing its related support function.
No changes are being made to the plant design, system
configuration, or method of operation. The proposed change does not
affect the ability of the AC power sources to perform their required
safety functions nor affect the ability of the features they support
to perform their respective safety functions. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
EDG [Emergency Diesel Generator]
An Individual Plant Examination (IPE) for Internal Events was
submitted to the NRC in response to Generic Letter 88-20 in
September 1992. The IPE was used to quantify the overall impact of
the proposed 14 day allowed outage time on core damage frequency.
Part III provides the results of a comprehensive Probabilistic
Safety Assessment (PSA) of the impact of the proposed AOTs [allowed
outage times] for the EDGs and Startup and Shutdown transformers. As
shown in Part III, there is not a significant increase in risk due
to the proposed change. Thus the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The existing specification 3.9.B.1 is being separated into two
segments (a and b) because of the proposed and different AOTs for
the Startup and Shutdown transformers. As a result of the PSA, the
AOT for the Startup transformer (a) is reduced from 7 days to 72
hours, while the AOT for the Shutdown transformer (b) remains at 7
days. The reduction of the AOT from 7 days to 3 days is based on the
relative risk importance of the Startup transformers support to the
balance of plant systems. Similarly, an additional reduction from 72
hours to 48 hours is proposed in the AOT for a simultaneous loss of
both the Startup transformer and an EDG (TS 3.9.B.4.b) based upon
the Startup transformer's contribution to risk in relation to the
EDG 14-day AOT risk assessment analysis and that two power sources
have been removed from the associated bus. The AOT reductions
represent a measurable decrease in risk as assessed in the PSA.
Thus, the probability or consequences of an accident previously
evaluated are not significantly increased.
The current technical specifications allow one EDG to be out of
service for three days based on the availability of the SUT [startup
transformer] and SDT [shutdown transformer] and the fact that each
EDG carries sufficient engineered safeguards equipment to cover all
design basis accidents. With one EDG out of service and a Loss of
Offsite Power (LOOP) condition, the capability to power vital and
auxiliary system components remains available via the other EDG, and
for one train of ESF equipment via the SDT for all operating,
transient and accident conditions. Increasing the EDG AOT to 14 days
provides flexibility in the maintenance and repair of the EDGs. The
EDG unavailability will be monitored and trended in accordance with
the Maintenance Rule. The PSA analyses supports the change to a 14
day AOT for the EDGs based on an insignificant increase in overall
risk. Implementation of the proposed change is expected to result in
less than a one percent increase in the baseline core damage
frequency (2.84E-05/yr), which is considered to be insignificant
relative to the underlying uncertainties involved with probabilistic
safety assessments. Additional conditions are added to the Standby
Liquid Control, Standby Gas Treatment, and Control Room High
Efficiency Air Filtration systems requiring the EDG associated with
these systems to remain operable while in the 14 day EDG AOT. Thus,
the 14 day EDG AOT does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Eliminating the 4.5.F.1 requirement for daily testing of the
operable diesel generator when the redundant diesel generator
becomes inoperable is consistent with the guidance provided in
Generic Letter 93-05. The change does not affect the ability of the
emergency diesel generator to perform on demand, and by actually
lowering the number of demands to demonstrate operability, reduces
the probability of equipment failure. The redundant EDG will remain
in service during the entire period of inoperability of the out-of-
service EDG. If a common cause failure cannot be ruled out, the
redundant EDG will be tested to assure operability. The proposed
revisions do not involve a significant change to the plant design or
operation, only to the manner in which remaining equipment is
confirmed to be operable, which is consistent with NRC guidance.
Thus operation of PNPS in accordance with the proposed license
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The 3.9.B.1 and 2 requirements to demonstrate both EDGs and
associated emergency buses operable are deleted. This change is
based on the NRC guidance provided in item 10.1 of Generic Letter
93-05, ``Line-Item Technical Specification Improvements to Reduce
Surveillance Requirements for Testing During Power Operation.''
Revising the methods for verifying EDG and emergency bus operability
does not physically alter the plant or have an affect on the
probability or consequences of an accident previously evaluated.
Deleting the testing requirements for an EDG when the other EDG is
inoperable does not increase the probability or consequences of an
accident previously evaluated because the reliability program and
routinely performed TS surveillances continue to provide the added
assurance sought by the testing. The elimination of this testing
will serve to improve the overall reliability of the EDGs. Since the
proposed change does not affect the design or negatively affect the
performance of the EDGs, the change will not result in a significant
increase in the consequences or probability of an accident
previously analyzed.
SGT [Standby Gas Treatment] and CRHEAF [Control Room High
Efficiency Air Filtration]
During normal plant operation, with one SGT or CRHEAF subsystem
inoperable, the inoperable subsystem must be restored to operable
status in 7 days. In this condition, the remaining operable SGT or
CRHEAF subsystem is adequate to perform the required radioactivity
release control function. However, the overall system reliability is
reduced because a single failure in the operable subsystem could
result in the radioactivity release control function not being
adequately performed. The 7 day completion time is based on
consideration of such factors as the availability of the operable
redundant SGT subsystem and the low probability of a DBA [design
basis accident] occurring during this period.
If the SGT or CRHEAF subsystem cannot be restored to operable
status within 7 days when in the Run, Startup, or Hot Shutdown MODE,
the plant must be brought to a MODE in which the LCO does not apply.
To achieve this status, the plant must be brought to at least Hot
Shutdown within 12 hours and to Cold Shutdown within 36 hours. The
allowed completion times are reasonable, based on operating
experience, to reach the required plant conditions from full power
conditions in an orderly manner and without challenging plant
systems.
Current TS governing refueling operations restrict fuel movement
if one train of SGTS or one train of CRHEAF are inoperable. In this
condition the remaining operable SGT and CRHEAF trains are adequate
to perform the required radioactivity release control functions.
However, the overall system reliability is reduced because a single
failure in the operable train could result in the radioactivity
release control function of the systems not being adequately
performed. New requirements are added that require if one train of
SGT or CRHEAF is inoperable, the redundant train of SGT or CRHEAF
must be demonstrated to be operable within 2 hours. This
substantiates the availability of the operable trains. Fuel handling
is limited only to the following 7 days and if the inoperable train
is not returned to an operable condition within that time frame, the
operable SGT train is placed in operation or fuel handling
activities are suspended. For CRHEAF, after 7 days, the operable
subsystem is demonstrated operable in accordance with existing
surveillances on a daily basis. The proposed changes do not modify
system design, use, or configuration in a manner different from
their original design and therefore do not involve a significant
increase in the consequences or probability of an accident
previously analyzed.
The revisions to make the SGT and CRHEAF TS sections similar in
wording are made to enhance usability and alleviate possible
confusion. These changes are strictly editorial, have no impact, and
do not alter
[[Page 31174]]
technical content or meaning of the specifications. These editorial
changes do not involve a significant increase in the probability or
consequences of an accident previously analyzed.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The operation of PNPS in accordance with the proposed license
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated because of the
following:
Definition of ``Operable-Operability''
The revised definition redefines the AC power needs to allow
either onsite or offsite power available for systems/subsystems to
be considered operable. This does not compromise the level of safety
already afforded to such systems/subsystems because the functional
operability requirements continue to be assured through the
technical specifications applicable to such systems/subsystems. AC
power availability continues to be assured through existing and
proposed surveillances and action statements applicable to AC power
systems. Reducing the need for both onsite and offsite power sources
in order to consider operable, the systems/subsystems powered by
these AC power sources, provides additional operational flexibility
by allowing redundant systems/subsystems to still be considered
``operable'' within the requirements of their functional operability
requirements. No new change or modes of plant operation are
involved. Therefore, operation in accordance with the revised
definition does not introduce any new or different kind of accident
from any accident previously evaluated.
EDG
The proposed amendment will extend the action completion/allowed
outage time for an inoperable emergency diesel generator from 72
hours to 14 days. The EDGs are designed as backup AC power sources
for essential safety systems in the event of loss of offsite power.
The proposed AOT does not change the conditions, operating
configurations or minimum amount of operating equipment assumed in
the safety analysis for accident mitigation. The EDGs and AC
equipment are not accident initiators. No change is being made in
the manner in which the EDG's provide plant protection. No new modes
of plant operation are involved. An extended AOT for one EDG does
not increase the probability of occurrence of a new or different
kind of accident previously evaluated. The PSA results concluded
that the risk contribution of the EDG AOT extension is
insignificant.
The current Pilgrim Technical Specifications requiring immediate
and daily testing of the redundant operable EDG is based on the
assumption that the increased testing provides additional assurance
that the equipment is available should it be needed. Industry
experience indicates that repetitive testing can place demands and
wear on the EDG without necessarily providing additional confidence
of availability. Also, the new surveillance requires verification
that offsite power is available and that a common cause failure is
not present. These actions provide assurance that the required
emergency buses can be energized with no loss of functions to
mitigate accident or transient conditions. In addition, Pilgrim has
implemented an EDG reliability program to maintain reliability of
EDGs. The proposed change does not introduce any new mode of plant
operation or new accident precursors, involve any physical
alterations to plant configurations, or make changes to system set
points that could initiate a new or different kind of accident.
Therefore, operation in accordance with the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
The AOT for an inoperable Startup Transformer is reduced from 7
days to 72 hours based upon the PSA that was performed to
quantitatively assess the risk impact of the proposed amendment. The
proposed reduction in AOT improves overall AC power source
availability because the SUT will potentially be inoperable for
shorter time periods. Therefore, reducing the AOT does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
SGT and CRHEAF
The SGT system is designed to filter radioactive materials from
the secondary containment following a postulated DBA or fuel
handling accident prior to release to the environment to ensure
compliance with 10 CFR 100 limits.
The CRHEAF is designed to filter intake air for the control room
atmosphere during conditions when normal intake air may be
contaminated.
The proposed revisions do not affect the ability of the SGTS or
CRHEAF to perform their intended function, do not create the
possibility of a new or different kind of accident from the loss of
coolant or fuel handling accidents previously analyzed, and do not
modify system configuration, use, or design. Therefore, operating
Pilgrim in accordance with this change will not create the
possibility of a new or different kind of accident from any accident
previously analyzed.
The revisions to make the SGT and CRHEAF TS sections similar in
wording are made to enhance usability and alleviate possible
confusion. These changes are strictly editorial, have no impact, and
do not alter technical content or meaning of the specifications.
These editorial changes do not create the possibility of a new or
different kind of accident from any previously analyzed.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The operation of PNPS in accordance with the proposed license
amendment will not involve a significant reduction in a margin of
safety because of the following:
Definition of ``Operable-Operability''
The implementation of the ``Operability'' definition clarifies
the relationship between AC power supplies and the operability
status of the equipment requiring AC power. No change is being made
in which the plant systems relied upon in the safety analyses
provide plant protection. Plant safety margins are maintained
through the limitations established in the TS LCOs. Since there will
be no significant reduction to the physical design or operation of
the plant there will be no significant reduction to any of these
margins.
EDG
Operation of PNPS in accordance with the proposed license
amendment will not involve a significant reduction in a margin of
safety. As shown in Part III [of the application dated April 25,
1996], incorporation of the proposed change involves an
insignificant reduction in the margin of safety.
The proposed changes do not significantly reduce the basis for
any technical specification related to the establishment of, or the
maintenance of, a safety margin nor do they require physical
modifications to the plant. Additional conditions are added to the
Standby Liquid Control, Standby Gas Treatment, and Control Room High
Efficiency Air Filtration systems requiring the diesel generator
associated with the redundant operable trains of these systems to
remain operable while in the 14 day EDG AOT. Moreover, the PSA
results showed that the risk contribution of extending the AOT for
an inoperable EDG is insignificant. The reduction in the AOT for the
SUT could improve availability, therefore, reducing overall risk.
Likewise the proposed changes in the deletion of testing have no
impact on the safety margin.
As previously stated, implementation of the proposed changes is
expected to result in an insignificant increase in: (1) power
unavailability to the emergency buses (given that a loss of offsite
power has occurred), and (2) core damage frequency. Implementation
of the proposed changes does not increase the consequences of a
previously analyzed accident nor significantly reduce a margin of
safety. Functioning of the EDGs and the manner in which limiting
conditions of operation are established are unaffected.
SGT and CRHEAF
SGT and CRHEAF contribute to the margin of safety by supporting
the secondary containment system during fuel handling by mitigating
the consequences of a fuel handling event. Allowing fuel movement to
continue as established in the LCOs does not involve a significant
reduction in the margin of safety because the first line of defense,
the other SGT and CRHEAF trains will be operable. The proposed
change will allow placing the Operable SGT subsystem in operation,
or in the case of CRHEAF, conducting daily testing, as an
alternative to suspending movement of irradiated fuel. This
alternative is less restrictive than the existing requirement,
however, the proposed requirements ensure that the remaining
subsystem is operable, that no failures that could prevent actuation
have occurred, and that any failure would be readily detected. The
proposed change does not result in a significant reduction in a
margin of safety because it allows operations which have the
potential for releasing radioactive material to the secondary
containment to continue only if the system designed to mitigate the
[[Page 31175]]
consequences of this release is functioning. Proper operation of
only one SGT or one CRHEAF subsystem is sufficient to mitigate the
consequences of any analyzed accident. Therefore, this change does
not change any of the assumptions in the accident analysis and does
not involve a significant reduction in a margin of safety.
The revisions to make the SGT and CRHEAF TS sections similar in
wording are made to enhance usability and alleviate possible
confusion. These changes are strictly editorial, have no impact, and
do not alter technical content or meaning of the specifications.
These editorial 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Jocelyn A. Mitchell, Acting
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: April 22, 1996
Description of amendment request: The licensee is proposing to
change the technical specifications to reflect a revision to the
overload cutoff limit on the manipulator crane inside the containment
at the Haddam Neck Plant. Due to a change in fuel design and supplier,
the heaviest fuel assembly design starting in Cycle 20 will be the
Westinghouse-supplied LOPAR design. Therefore, the heaviest combination
beginning in Cycle 20 will be the Westinghouse LOPAR fuel assembly with
a full-length rod cluster control assembly (RCCA) inserted. It will now
be used as the standard for the overload cutoff limit on the
manipulator crane.
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 proposed change will revise the method of determining the
overload cutoff limit for the manipulator crane. The actual absolute
value of the cutoff limit will not be increased and will not affect
the [probability] of any plant accidents.
Since there is no actual increase in the absolute overload
cutoff limit, there will be no adverse effects to the crane, cables,
or associated hardware. Therefore, there is no impact on the crane's
ability to perform its intended function. Even though the net
lifting forces on an individual assembly have increased 25 pounds,
the limit is within the recommended Westinghouse guidelines with
respect to fuel handling and will not result in potential damage to
assembly grids during fuel handling activities.
As such, CYAPCO [Connecticut Yankee Atomic Power Company] has
concluded that these changes do not involve an 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.
The changes conservatively revise the method of determining the
overload cutoff limit for the manipulator crane. There is no impact
on the basic functioning of plant systems or equipment. Therefore,
the change does not create a malfunction that is different from
those previously evaluated.
As such, the proposed changes described above 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 revisions in the methodology for determining the
overload cutoff limit for the manipulator crane is conservative and
in accordance with vendor standards. The changes do not adversely
affect any equipment credited in the safety analysis. Also, the
changes do not adversely affect the probability or consequences of
any plant accident, including the fuel handling accident or offsite
doses associated with those accidents.
As such, the proposed changes have no significant impact on 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: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Project Director: Phillip F. McKee
Duke Power Company, Docket Nos. 50-413 and 50-414, Catawba Nuclear
Station, Units 1 and 2, York County, South Carolina
Date of amendment request: December 14, 1995, as supplemented by
letter dated May 16, 1996
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) to improve the TS Action
Statements and Surveillance Requirements for diesel generators in
accordance with the recommendations and guidance in Generic Letter 93-
05, Generic Letter 94-01, NUREG-1366, and NUREG-1431. The proposed
amendments would also incorporate technical and administrative 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:
Criterion 1
Operation of the facilities in accordance with the requested
amendments will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Improvements to the LCOs [limiting condition for operation] and
surveillance requirements for the emergency diesel generators do not
affect their capability to provide emergency power to plant vital
instruments and safety related equipment. In fact, these
improvements make the diesel generators more reliable since they
significantly reduce the amount of wear and stress due to excessive
and unnecessary testing. The proposed monthly testing of the diesel
generator continues to ensure that the system is ready for service
when needed. The fast starts and fast loadings continue to ensure
that the timing and loading requirements for engineered safety
features actuation are met. The proposed changes do not affect any
of the design basis accident analyses previously evaluated.
Therefore, these proposed changes do not involve any increase in the
probability or consequences of any accident previously evaluated.
The proposed changes are fully consistent with the recommendations
and guidance contained in GL [Generic Letter] 93-05, GL 94-01,
NUREG-1366, NUREG-1431, and are compatible with plant operating
experience.
Criterion 2
Operation of the facilities in accordance with the requested
amendments will not create the possibility of a new or different
kind of accident from any accident previously evaluated. The
proposed changes in fact improve the reliability of the diesel
generators by eliminating unnecessary wear and stress. Improved
reliability decreases the failure probability which also decreases
the probability of an accident not previously evaluated. None of the
requested amendments increase the common mode failure probability
thus would not increase the chance of both EDG's [emergency diesel
[[Page 31176]]
generators] for a particular nuclear unit being out of service
simultaneously. The proposed changes are fully consistent with the
recommendations and guidance contained in GL 93-05, GL 94-01, NUREG-
1366, NUREG-1431, and are compatible with plant operating
experience.
Criterion 3
Operation of the facilities in accordance with the requested
amendments will not involve a significant reduction in a margin of
safety. The proposed monthly testing of the diesel generators
continues to ensure that the system is ready for service when
needed. The fast starts and fast loadings continue to ensure that
the timing and loading requirements for engineered safety features
actuation are met. The proposed changes improve the reliability of
the diesel generators. Implementation of the Maintenance Rule also
ensures continued reliability of the diesel generators. No margin of
safety is decreased as a result of these TS changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi and Docket
No. 40-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: April 18, 1996, as supplemented by
letter dated June 4, 1996
Description of amendment request: The licensee has proposed to (1)
amend Limiting Condition for Operation (LCO) 3.10.6 and Surveillance
Requirement 3.10.6.3, and (2) add a Surveillance Requirement 3.10.6.4
of the Technical Specifications (TSs) for the Grand Gulf Nuclear
Station, Unit 1, and the River Bend Station, Unit 1, to allow another
method of fuel movement and loading in the core when control rods are
removed or withdrawn from defueled core cells. Currently, LCO 3.10.6
allows only fuel loading as part of the approved spiral reloading
sequence to prevent fuel loading into core cells in which the control
rod has been removed or withdrawn. This amendment request does not
withdraw this approved method, revise the frequency of performing the
surveillance during fuel loading, or alter the method of verifying the
fuel is being loaded in compliance with the approved method. Grand Gulf
Unit 1 and River Bend Unit 1 are both General Electric (GE) Boiling
Water Reactor (BWR)-6 plants, the latest version of the GE design
series.
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:
Entergy Operations, Inc. [(EOI)] propose[d] to change the
current Grand Gulf Nuclear Station (GGNS) and River Bend Station
(RBS) Technical Specifications [(TSs)]. The specific proposed change
is to add an additional method of performing fuel loading into LCO
3.10.6, ``Multiple Control Rod Withdrawal - Refueling''. The
proposed change would allow fuel loading [in the core] if a positive
means of assuring fuel assemblies cannot be loaded into a core cell
with a withdrawn or removed control rod is in effect. [Currently,
the TSs for both plants allow fuel assembles to be loaded in
compliance with an approved spiral reload sequence which is used to
ensure the reactivity additions are minimized. Spiral loadings
encompass reloading a core cell on the edge of a continuous fueled
region.]
The Commission has provided standards for determining whether a
no significant hazards consideration exists as stated in 10 CFR
50.92(c). A proposed amendment to an operating license involves no
significant hazards consideration if operation of the facility in
accordance with the proposed amendment would not: (1) involve a
significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a
new or different kind of accident from any accident previously
evaluated; or (3) involve a significant reduction in a margin of
safety.
Entergy Operations, Inc. [EOI] has evaluated the no significant
hazards consideration in its request for this license amendment and
determined that no significant hazards consideration results from
this change. In accordance with 10 CFR 50.91(a), Entergy Operations,
Inc. [EOI] is providing the analysis of the proposed amendment
against the three standards in 10 CFR 50.92(c). A description of the
no significant hazards consideration determination follows:
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The refueling interlocks (i.e., the refueling equipment and one-
rod-out interlocks) allowed to be bypassed by Technical
Specification [TS] LCO 3.10.6 are explicitly assumed in the analysis
of the control rod removal error or fuel loading error during
refueling. This analysis evaluates the consequences of control rod
withdrawal during refueling. Criticality and, therefore, subsequent
prompt reactivity excursions are prevented during the insertion of
fuel, provided all control rods are fully inserted during the fuel
insertion. The refueling interlocks accomplish this by preventing
loading fuel into the core with any control rod withdrawn, or by
preventing withdrawal of a rod from the core during fuel loading.
LCO 3.10.6 allows multiple control rod withdrawals, control rod
removals, associated control rod drive (CRD) removal, or any
combination of these, and the ``full in'' position indication input
to the refueling interlocks is allowed to be bypassed for each
withdrawn control rod if all fuel has been removed from the cell.
This supports the GGNS Updated Final Safety Analyses Report (UFSAR)
and RBS Updated Safety Analyses Report (USAR) analyses since, with
no fuel assemblies in the core cell, the associated control rod has
no reactivity control function and does not need to remain inserted.
Prior to reloading fuel into the cell, however, the associated
control rod must be inserted to ensure that an inadvertent
criticality does not occur, as evaluated in the analysis.
The Technical Specification [TS] requirements prohibiting fuel
loading was placed in the Technical Specifications [TSs] for GGNS
and RBS as part of the originally enforced Technical Specification
[TS] requirements to resolve NRC concerns identified in IE
Information Notice No. 83-35, ``Fuel Movement with Control Rods
Withdrawn at BWRs,'' (IEN 83-35). IEN 83-35 details instances where
fuel assemblies were loaded into core cells while the control rod
was withdrawn and discusses that the General Electric Company (GE)
had issued Service Information Letter (SIL) No. 372.
SIL No. 372 discusses a potential event where 8 fuel assemblies
are loaded into 2 [two] adjacent core cells where the control rods
are withdrawn and no action is taken to recover from the errors. In
this SIL GE identified that the probability of such an event
occurring was extremely low but potentially slightly higher than
10-6 probability of the event even further to where it need not
be considered credible (i.e., below 10-6 per reactor year), GE
recommended that the additional administrative control of
prohibiting loading fuel with withdrawn rods be enforced.
The proposed change will only provide an additional way to meet
the intent of the original GE recommendation. [The currently
approved method is listed in LCO 3.10.6 and Surveillance Requirement
3.10.6.3.]. The proposed change will provide the additional
allowance to perform fuel loading only if an additional positive
means of assuring fuel assemblies cannot be loaded into a core cell
with a withdrawn or removed control rod is in effect. The positive
means will entail a physical barrier such that, even if refueling
procedures were violated and an attempt was made to load a fuel
assembly into a core cell with a withdrawn or removed control rod,
the action would be prevented. This requirement provides sufficient
additional restrictions to meet the intent of the GE recommendation
to add additional administrative controls to prevent the postulated
event from occurring.
The probability of an inadvertent criticality occurring will
continue to be precluded by
[[Page 31177]]
the same number of layers of administrative controls [as the
currently approved method]; therefore, the proposed change does not
significantly increase the probability or consequences of an
accident previously evaluated.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The administrative changes in the Technical Specification [TS]
requirements do not involve a change in the design of the plant. The
proposed requirements will continue to ensure that fuel is not
loaded into a core cell that is associated with a removed or
withdrawn control rod.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
The margin of safety associated with criticality events during
fuel handling is provided by the event being a non credible event.
The proposed change will only provide an additional means to meet
the same intent of ensuring that the event is of such low
probability as to be considered non credible. The proposed change
will provide the additional allowance to perform fuel loading only
if an additional positive means of assuring fuel assemblies cannot
be loaded into a core cell with a withdrawn or removed control rod
is in effect. The positive means will entail a physical barrier such
that even if refueling procedures were violated and an attempt was
made to load a fuel assembly into a core cell with a withdrawn or
removed control rod the action would be prevented. This requirement
provides sufficient additional restrictions to ensure that the event
is of such low probability as to be considered non credible.
The probability of an inadvertent criticality occurring will
continue to be precluded by the same number of layers of
administrative controls [as the currently approved method];
therefore, this change does not reduce the level of safety imposed
by the current Technical Specification [TS] requirements.
Therefore, the proposed changes do not cause 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: (1) Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120, for Grand Gulf
Nuclear Station and (2) Government Documents Department, Louisiana
State University, Baton Rouge, LA 70803, for River Bend Station.
Attorney for licensee: (1) Nicholas S. Reynolds, Esquire, Winston
and Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502,
for Grand Gulf Nuclear Station and (2) Mark Wetterhahn, Esq., Winston &
Strawn, 1400 L Street, N.W., Washington, DC 20005, for River Bend
Station.
NRC Project Director: William D. Beckner
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: May 9, 1996
Description of amendment request: The amendment request would allow
allow the licensee to perform the surveillance of the relief mode of
operation of each of the 20 safety/relief valves (S/RVs) on the 4 main
steam lines without physically lifting the disk off the seat at power.
The proposed changes are to Surveillance Requirements (SRs) 3.4.4.3,
Safety/Relief Valves, 3.5.1.7, Automatic Depressurization System
Valves, and 3.6.1.6.1, Low-Low Set Valves, of the Technical
Specifications, and the changes would state that the required operation
of the valve to verify is that the relief-mode actuator strokes when
the valve is manually actuated. Each S/RV is a Dikkers, 8 X 10, direct-
acting, spring loaded, safety valve with attached pneumatic actuator
for relief-mode operation. Eight of the S/RVs use the relief mode to
perform the Automatic Depressurization System (ADS) function. Also, six
S/RVs, two of which are also ADS S/RVs, use the relief mode to perform
the Low-Low Set valve function. The licensee also proposed changes to
the Bases of the Technical Specifications that are associated with the
above proposed 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: The Dikkers S/RV provides
pressure relief based on the principle of vertically moving the stem
that attaches directly to the valve disk. The force that provides the
stem movement is provided by one of two sources; the vessel pressure
directly against the force of the stem spring (safety mode), or the
pneumatic actuator arm against the force of the stem spring (relief
mode). ASME Boiler and Pressure Vessel Code requires testing the safety
mode of operation once every five year operating cycle. Once a safety
valve is installed, the safety mode is never tested while the S/RV is
installed in the plant. The testing of the relief mode of operation for
a direct-acting S/RV provides verification that the control functions
of electrical and pneumatic connections have been properly reconnected,
and that the actuator arm will provide the necessary force to operate
the S/RV.
This proposed change provides verification of proper control
connections by requiring the pneumatic and electrical controls to
cycle the actuator arm on each S/RV after installation in the
drywell. The test population of S/RVs removed each outage for safety
setpoint testing will be tested in the relief mode. This testing
will demonstrate that the installed S/RVs will function properly in
the relief mode. The remaining installed S/RVs will continue to be
tested for proper system function. As presently required by GGNS
Technical Specifications and administrative procedures, proper
operation of the solenoid control block will be demonstrated by
providing an open signal to each S/RV, with a check to verify that
each solenoid valve repositions. Verification of proper solenoid
valve operation, in addition to the proper relief-mode operation of
the test population, provides assurance that the S/RV will perform
as expected when control air pressure is applied to the solenoid
valve control block.
Entergy Operations, Inc. is proposing that the Grand Gulf
Nuclear Station Operating License be amended to perform the
surveillance of each safety relief valve (S/RV) relief mode of
operation without physically lifting the disk off the seat at power.
During the refueling outage, a sample population of the S/RVs
will be removed for safety-mode setpoint testing in accordance with
the GGNS IST program, using ASME Boiler and Pressure Vessel Code,
Section XI. Each of these removed S/RVs will be tested in the relief
mode to verify that the pneumatic actuator functions correctly, and
this test sample will be used to provide assurance that the
installed S/RV pneumatic actuators will function properly. After the
test sample of S/RVs has been replaced with recertified spares, and
S/RV controls have been connected, the upper stem nut that couples
the valve stem to each newly- installed S/RV's pneumatic actuator
will be moved up the stem to allow an uncoupled actuation of the
relief-mode actuator. Control air pressure to each actuator will be
reduced from normal system pressure to prevent damaging the
pneumatic relief-mode actuator. The actuator will be remotely
operated from the control room, as required by current test methods,
and visual verification will be performed for proper actuator
response and range of motion. After proper actuator operation has
been verified, the upper stem nut will be returned to its operating
stem location. Verification of proper system logic controls and
function for every installed S/RV will continue to be performed, as
required by Technical Specifications.
The commission has provided standards for determining whether a
no significant hazards consideration exists as stated in 10 CFR
50.92(c). A proposed amendment to an operating license involves no
significant hazards if the operation of the facility in accordance
with the proposed amendment would not: (1) involve a significant
increase
[[Page 31178]]
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.
Entergy Operations has evaluated the no significant hazards
considerations in its request for a license amendment. In accordance
with 10 CFR 50.91(a), Entergy Operations, Inc. is providing the
following analysis of the proposed amendment against the three
standards in 10 CFR 50.92:
a. No significant increase in the probability or consequences of
an accident previously evaluated results from this change.
Each refueling outage, a test sample of the population of S/RVs
is removed from the plant to perform testing as required by ASME
Boiler and Pressure Vessel Code, Section XI. These S/RVs will be
stroked in the relief mode during as-found testing, and are
therefore verified to operate properly when each S/RV stem is raised
by the relief-mode pneumatic actuator. This proposed surveillance
verifies proper S/RV relief-mode operation of all installed S/RVs
based upon this test sample. This testing, in conjunction with
replacement of each S/RV prior to the end of its expected service
life, provides reasonable assurance that the installed S/RVs will
perform as well as the test population of S/RVs.
After the S/RVs have been replaced in the plant, and after all
controls are reconnected, the relief-mode actuator on each newly-
installed S/RV will be uncoupled from the S/RV stem, and stroked.
This actuator stroke will verify that no damage has occurred to the
relief-mode actuator during S/RV transportation from its storage
location to its operating location. The direct coupling of the valve
stem to disk provides assurance that proper relief actuation will
occur when the actuator is operated. The safety-mode components are
completely encased within the valve body and bonnet, which provides
a rugged structure to prevent damage to these components. The
remaining installed S/RVs will continue to be tested for proper
control system function as previously required by Technical
Specifications. The direct coupling of the S/RV stem to disk
provides assurance that proper relief-mode actuation will occur when
the actuator is operated. The safety mode of the GGNS S/RVs is not
affected by a malfunction of the relief-mode components.
Blockage of each S/RV discharge line will be prevented by the
same Foreign Material Exclusion (FME) controls that exist for other
reactor vessel and support systems. These FME controls, combined
with the horizontal orientation of the S/RV discharge piping mating
surfaces, provide reasonable assurance that discharge line blockage
will not occur.
Therefore, no significant increase in the probability or
consequences of an accident previously evaluated results from this
proposed change.
b. This change would not create the possibility of a new or
different kind of accident from any previously analyzed.
The proposed change demonstrates that each S/RV will perform its
intended relief-mode function, which is the intent of the present
surveillance. The relief mode of S/RV operation is demonstrated to
be operable based upon successful performance of a test population,
S/RV component service life, and existing Technical Specification
surveillances. No new failure mechanisms to the relief- mode of
operation are introduced, as the proposed surveillance verifies
relief actuator operability. Plant FME controls, combined with the
horizontal orientation of the S/RV discharge piping mating flange,
provides reasonable assurance that discharge line blockage will not
occur. This proposed change does not add any new systems,
structures, or components, nor does it introduce new S/RV operating
modes.
Therefore, this change would not create the possibility of a new
or different kind of accident from any previously analyzed.
c. This change would not involve a significant reduction in the
margin of safety.
This proposed change will verify that the relief mode of all
installed S/RVs will operate properly based upon demonstrated relief
mode performance of a sample of S/RVs. The failure mode of the S/RV
relief function would require a failure of either the pneumatic
actuator, lifting linkage, or solenoid block. Each of these items
has been verified to have a service life exceeding the replacement
cycle of each S/RV. Therefore, proper operation of a sample
population of S/RVs provides reasonable assurance that the remaining
S/RVs would perform identically, within the original margin of
expected S/RV operability. In addition, each S/RVFEs solenoid block
and control functions will continue to be tested and cycled each
refueling outage. The removal of the valve stroke surveillance for
all S/RVs does not increase the possibility of valve malfunction,
since valve stroke is verified during the as-found testing of the
sample population of S/RVs. This proposed surveillance test reduces
the number of S/RV actuations, and therefore, reduces challenges to
the system both mechanically and thermally. Also, the proposed
alternative method of testing reduces the possibility of a stuck-
open S/RV, since this proposed method will not stroke the S/RVs with
the reactor pressurized during reactor power operations.
Therefore, this change would not involve a significant reduction
in the margin of safety.
Based on the above evaluation, Entergy Operations, Inc. has
concluded that operation in accordance with the proposed amendment
involves no significant hazards considerations.
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: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: May 31, 1996
Description of amendment request: The amendment would provide an
alternative method to compensate for inoperable refueling equipment
interlocks. The alternative method would be to insert a control rod
withdrawal block and verify that all control rods are fully inserted;
however, the control rods required to be inserted would not apply to
those control rods withdrawn in accordance with LCO 3.10.6, ``Multiple
Control Rod Withdrawal -Refueling.'' The amendment would add an
additional Required Action for Limiting Condition for Operation (LCO)
3.9.1, ``Refueling Equipment Interlocks,'' of the Technical
Specifications (TSs) for Grand Gulf Nuclear Station, Unit 1 (GGNS). The
alternative method then could be used to respond to inoperable
interlocks instead of only the current method of halting in-vessel fuel
movement with equipment associated with the inoperable interlock.
The proposed change does not remove the current Required Action
method for LCO 3.9.1 and does not change the surveillance requirements
on the refueling equipment. The licensee has also provided changes to
the Bases of the TSs for the proposed amendment.
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. The licensee has proposed the amendment for the TSs for
both GGNS and River Bend Station (RBS). References made to the RBS TSs
and to RBS in the licensee's analysis of no significant hazards
consideration have been removed and replaced by [...]. The licensee's
analysis is presented below:
Entergy Operations, Inc. proposes to change the current Grand
Gulf Nuclear Station (GGNS) [...] Technical Specifications. The
specific proposed change adds additional acceptable Required Actions
to the Actions of LCO 3.9.1, ``Refueling Equipment Interlocks,''
[for inoperable interlocks]. The additional Required Actions will
add an alternative [method] to [the current method of] suspending
fuel movement in the reactor vessel when the refueling interlocks
are inoperable. The requested alternative is to insert a control rod
withdrawal block
[[Page 31179]]
immediately and verify all control rods required to be inserted are
fully inserted. [The control rods required to be inserted would not
apply to control rods withdrawn in accordance with LCO 3.10.6,
``Multiple Control Rod Withdrawal--Refueling.'']
The Commission has provided standards for determining whether a
no significant hazards consideration exists as stated in 10 CFR
50.92(c). A proposed amendment to an operating license involves no
significant hazards consideration if operation of the facility in
accordance with the proposed amendment would not: (1) involve a
significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a
new or different kind of accident from any accident previously
evaluated; or (3) involve a significant reduction in a margin of
safety.
Entergy Operations, Inc. has evaluated the [criteria for] no
significant hazards consideration in its request for this license
amendment and determined that no significant hazards consideration
results from this change. In accordance with 10 CFR 50.91(a),
Entergy Operations, Inc. is providing the analysis of the proposed
amendment against the three standards in 10 CFR 50.92(c). A
description of the no significant hazards consideration
determination follows:
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The refueling interlocks are explicitly assumed in the GGNS
Updated Final Safety Analyses Report (UFSAR) [...] analysis of the
control rod removal error or fuel loading error during refueling.
This analysis evaluates the probability and consequences of control
rod withdrawal during refueling. Criticality and, therefore,
subsequent prompt reactivity excursions are prevented during the
insertion of fuel, provided all control rods are fully inserted
during the fuel insertion. The refueling interlocks accomplish this
by preventing loading fuel into the core with any control rod
withdrawn, or by preventing withdrawal of a rod from the core during
fuel loading.
When the refueling interlocks are inoperable the current method
of preventing the insertion of fuel when a control rod is withdrawn
is to prevent fuel movement. This method is currently required by
the Technical Specifications. An alternate method to ensure that
fuel is not loaded into a cell with the control rod withdrawn is to
prevent control rods from being withdrawn and verify that all
control rods required to be inserted are fully inserted. The
proposed actions will require that a control rod block be placed in
effect thereby ensuring that control rods are not subsequently
inappropriately withdrawn. Additionally, following placing the
control rod withdrawal block in effect, the proposed actions will
require that all required control rods be verified to be fully
inserted. This verification is in addition to the requirements to
periodically verify control rod position by other Technical
Specification requirements. These proposed actions will ensure that
control rods are not withdrawn and cannot be inappropriately
withdrawn because an electrical or hydraulic block to control rod
withdrawal is in place. Like the current requirements the proposed
actions will ensure that unacceptable operations are blocked (e.g.,
loading fuel into a cell with a control rod withdrawn [would be
blocked]).
The proposed additional acceptable Required Actions provide the
same level of assurance that fuel will not be loaded into a core
cell with a control rod withdrawn as the current Required Action or
the Technical Specification Surveillance Requirement.
Therefore, the proposed change does not significantly increase
the probability or consequences of an accident previously evaluated.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The change in the Technical Specification requirements does not
involve a change in plant design. The proposed requirements will
continue to ensure that fuel is not loaded into the core when a
control rod is withdrawn except following the requirements of LCO
3.10.6, ``Multiple Control Rod Removal--Refueling,'' which is
unaffected by this change.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
As discussed in the Bases for the affected Technical
Specification requirements, inadvertent criticality is prevented
during the insertion of fuel provided all control rods are fully
inserted during the fuel insertion. The refueling interlocks
function to support the refueling procedures by preventing control
rod withdrawal during fuel movement and the inadvertent loading of
fuel when a control rod is withdrawn.
The proposed change will allow the refueling interlocks to be
inoperable and fuel movement to continue only if a control rod
withdrawal block is in effect and all required control rods are
verified to be fully inserted. These proposed Required Actions
provide the same level of protection as the refueling interlocks by
preventing a configuration which could lead to an inadvertent
criticality event. The refueling procedures will continue to be
supported by the proposed required actions because control rods
cannot be withdrawn and as a result fuel cannot be inadvertently
loaded when a control rod is withdrawn.
Therefore, the proposed changes do not cause 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: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: May 31, 1996, as supplemented by letter
dated May 2, 1996.
Description of amendment request: The amendment request would
revise the current reactor vessel material surveillance program
schedule for GGNS. This is the schedule for withdrawing surveillance
capsules from the reactor vessel for testing to measure the impact of
neutron irradiation of the vessel material and is required by Section
III.B.3 of Appendix H, ``Reactor Vessel Material Surveillance Program
Requirements,'' of 10 CFR Part 50. The schedule must be approved by the
Nuclear Regulatory Commission (NRC) before implementation.
For GGNS, there are three surveillance capsules inside the reactor
vessel, each of which contains specimens of the reactor vessel
material. The first capsule was removed from the reactor vessel on May
7, 1995, during the 7th refueling outage. Because no useful data is
expected from testing the material specimens in the first capsule, the
request would allow the first capsule to be placed back into the
vessel.
As part of revising the schedule, the licensee is also renumbering
the three surveillance capsules so that the capsule removed at the 7th
refueling outage becomes the third capsule when it is placed back in
the vessel. The proposed change would, however, not extend the time
that the next capsule (the renumbered first capsule) would be withdrawn
from the GGNS reactor vessel.
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:
Entergy Operations, Inc., proposes to change the withdrawal
schedule for the reactor vessel material surveillance capsules [and
renumber the capsules]. The revised schedule for withdrawal of the
surveillance capsules is withdrawal of the first capsule at 24
Effective Full Power Years. The withdrawal schedule for the second
capsule is to be determined at a later date. The third capsule which
was withdrawn on May 7, 1995 is to be returned to reactor vessel
during
[[Page 31180]]
the Fall, 1996 outage and retained as a standby. [The current
schedule for withdrawal of the three capsules is 8 and 24 Effective
Full Power Years for the first two capsules, and the third capsule
is a spare with no specific schedule for withdrawal.]
The Commission has provided standards for determining whether a
no significant hazards consideration exists as stated in 10 CFR
50.92(c). A proposed amendment to an operating license involves no
significant hazards consideration if operation of the facility in
accordance with the proposed amendment would not: (1) involve a
significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a
new or different kind of accident from any accident previously
evaluated; or (3) involve a significant reduction in a margin of
safety.
In consideration of the October 4, 1995, decision of the Atomic
Safety and Licensing Board concerning an amendment request from
Perry Nuclear Power Plant, Entergy Operations, Inc. has evaluated
the no significant hazards consideration in its request for a change
to the withdrawal schedule required by 10 CFR 50, Appendix H, and
determined that no significant hazards consideration results from
this change. In accordance with 10 CFR 50.91(a), Entergy Operations,
Inc. is providing the analysis of the proposed amendment against the
three standards in 10 CFR 50.92(c):
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The change revises the withdrawal schedule for the reactor
vessel material surveillance capsules and returns a withdrawn
capsule to the reactor vessel. The capsules [only contain specimens
of the reactor vessel material and] are not an initiator of any
previously analyzed accident. The withdrawal or return of the
surveillance capsule does not effect the probability or consequences
of any previously analyzed accident. Extending the time for
withdrawal of the first capsule and returning the withdrawn capsule
to the vessel do not adversely affect the pressure temperature limit
curves for the reactor vessel. Regulatory Guide 1.99 [, ``Effects of
Residual Elements on Predicted Radiation Damage to Reactor Vessel
Materials,''] is currently used to prepare the pressure temperature
limit curves and is inherently conservative for boiling water
reactors (BWRs)[, as GGNS]. The current pressure temperature limit
curves will continue to be adhered to. Additionally, [GGNS]
participates in the supplemental test program designed to
significantly increase the amount of BWR surveillance data. [This
program has supplemental capsules which were installed in the Cooper
and Oyster Creek Nuclear Power Plants, which contain the limiting
GGNS weld and plate vessel material, and which will be withdrawn in
1996, 2000, and 2002.] This program will be used to complement the
GGNS surveillance program such that postponement of the capsule
withdrawals will have minimal impact on the understanding of the
irradiation effects on the GGNS vessel.
[The licensee stated in its May 2, 1996, letter that testing of
the specimens in the removed capsule may not provide useful
indicators of the damage to the vessel material because the low
neutron fluence on the vessel and the good material chemistry will
result in a minimal null-ductility temperature shift. Testing the
material specimens will destroy them; however, placing the capsule
back in the vessel will allow the specimens to have more irradiation
until useful data could be obtained from testing the specimens.]
Therefore, the proposed change does not significantly increase
the probability or consequences of an accident previously evaluated.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Returning the withdrawn capsule to the vessel and postponing the
withdrawal of the first capsule do not contribute to the possibility
of a new or different kind of accident or [plant] malfunction from
those previously analyzed [in the Updated Final Safety Analysis
Report for GGNS]. Failure of the reactor vessel is not a credible
accident since the vessel itself is a highly reliable component.
This change does not affect that determination. The potential for
reactor vessel cracking will be adequately assessed by the proposed
withdrawal schedule.
[The licensee stated in its May 2, 1996, letter that testing of
the specimens in the removed capsule may not be useful indicators of
the damage to the vessel material because the low neutron fluence on
the vessel and good material chemistry will result in a minimal
shift.]
In addition, the results from the supplemental test program will
provide indication of the condition of the vessel until the data
from the first GGNS capsule[, withdrawn and tested,] are available.
The proposed change provides the same level of confidence in the
integrity of the vessel. The pressure temperature curves are
currently controlled by the Technical Specifications and are
determined using the conservative methodology in Regulatory Guide
1.99. Therefore, the possibility of failure of the reactor vessel is
not increased. The proposed change does not involve a change in the
design of the plant. The current pressure temperature limit curves
are inherently conservative and will continue to be adhered to.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
The current pressure temperature limit curves [for the reactor
vessel] are inherently conservative and provide sufficient margin to
ensure the integrity of the reactor vessel. The [proposed] changes
do not adversely affect these curves. The supplemental test program
will be used to complement the GGNS surveillance program such that
postponement of the capsule withdrawal [and testing] will have
minimal impact on the understanding of irradiation effects on the
GGNS vessel. The capsules removed in 1996 as part of the
supplemental program will have a [neutron] fluence higher than the
25% of the design life fluence used in establishing the original
GGNS [reactor vessel material surveillance program] schedule;
therefore, the use of the supplemental test program results will
meet the intent of the original test schedule.
Therefore, the proposed changes do not result in a significant
reduction in the margin of safety.
Based on the above evaluation, Entergy Operations, Inc. has
concluded that operation in accordance with the proposed change
involves no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Dates of amendment request: March 21, 1996, and May 13, 1996
Description of amendment request: The licensee proposed to change
the Turkey Point Units 3 and 4 Technical Specifications (TS) to
relocate the requirements of the Radiological Effluent Technical
Specifications (RETS) to other documents.
The proposed amendments would relocate the LIMITING CONDITIONS FOR
OPERATION (LCO) and SURVEILLANCE REQUIREMENTS associated with the RETS
in accordance with GL 89-01, NUREG-1301, and NUREG-1431, Rev. 1. The
definition in TS 1.15, ``Members of the Public,'' would be deleted
since it is already located in 10 CFR Part 20 and has been inserted
into the Offsite Dose Calculation Manual (ODCM). The definitions for
the ODCM and Process Control Program (PCP) would be relocated to the
Administrative Controls section of the TS. TS 3/4.3.3.5 and the
radioactive gaseous effluent portion of TS 3/4.3.3.6 and associated
tables, instrumentation operational conditions, remedial actions and
surveillance requirements would be controlled through the ODCM or PCP
and associated procedures. Technical
[[Page 31181]]
Specification Administrative Control sections would contain the
programmatic controls for the ODCM and PCP. The remaining portion of TS
3.3.3.6 would retain the operational conditions, remedial actions, and
surveillance requirements for the explosive gas monitor
instrumentation.
The procedural details of the current TS on radioactive effluents
and radiological environmental monitoring would be deleted. Associated
operational conditions, remedial actions and surveillance requirements
presently in the Technical Specifications would be controlled through
the ODCM or PCP.
Administrative changes to the TS were also proposed due to
paragraph and section numbering changes and relocations associated with
the proposed technical changes.
New sections TS 6.8.4f and 6.8.4g were proposed to provide
programmatic controls for the Radiological Effluents Controls Program
and the Radiological Environmental Monitoring Program.
TS 6.9.1.3 and TS 6.9.1.4 would be simplified and the reporting
details now contained in these specifications would be relocated to the
ODCM or PCP with the exception of the requirement to report licensee-
initiated changes to the PCP in the Annual Radioactive Effluent Release
Report.
New record retention requirements changes for the ODCM and PCP
would be added to TS 6.10.3q.
In summary, as provided in the guidance, the current technical
content of the specifications which would be transferred to the ODCM or
the PCP. New programmatic controls for radioactive effluents and
radioactive effluent monitoring would be added to the TS, as well as
further clarification to the definitions of the ODCM and PCP. The
Technical Specification requirements for Gas Decay Tanks and Explosive
Gas Mixture would be relocated to the Plant Systems section of the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1)Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The changes being proposed are administrative in nature in that
they relocate Technical Specification requirements associated with
RETS from the Technical Specifications to the ODCM or PCP. These
changes are in accordance with the recommendations contained in GL
89-01, NUREG 1301, and NUREG 1431 Rev. 1. The only change being made
to existing requirements or commitments are administrative in
nature. The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident, nor do they affect
any assumptions or conditions in any of the accident analyses. Since
the accident analyses remain bounding, their probability or
consequences are not adversely affected. Therefore, the probability
or consequences of an accident previously evaluated are not
affected.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The changes being proposed are administrative in nature in that
they relocate Technical Specification requirements associated with
RETS from the Technical Specifications to the ODCM or PCP. These
changes are in accordance with the recommendations contained in GL
89-01, NUREG 1301, and NUREG 1431, Rev. 1. The only change being
made to existing requirements or commitments are administrative in
nature. The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment used to
mitigate the consequences of an accident.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be
created.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The changes being proposed are administrative in nature in that
they relocate Technical Specification requirements associated with
RETS from the Technical Specifications to the ODCM or PCP. These
changes are in accordance with the recommendations contained in GL
89-01, NUREG 1301, and NUREG 1431, Rev. 1. The only change being
made to existing requirements or commitments are administrative in
nature. All technical content is preserved. The operating limits and
functional capabilities of the affected systems, structures, and
components are unchanged by the proposed amendments.
Therefore, a significant reduction in a margin of safety would
not be involved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: Frederick J. Hebdon
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Dates of amendment request: May 28, 1996
Description of amendment request: The licensee proposed to change
the Turkey Point Units 3 and 4 Technical Specifications (TS) to change
the licensed qualifications of the Operations Manager. The proposed
change would delete the qualification option that the Operations Manger
could have held a Senior Reactor Operator License on a boiling water
reactor and replace it with an option that this individual could have
completed the Turkey Point Nuclear Plant Senior Management Operation
Training Course (i.e., certified at an appropriate simulator for
equivalent senior operator knowledge level).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The change being proposed is administrative in nature, addresses
organizational personnel qualification issues, and does not affect
assumptions contained in plant safety analyses, the physical design
and/or operation of the plant, or Technical Specifications that
preserve safety analysis assumptions.
The individual Florida Power & Light Company (FPL) chooses to
fill the position of Operations Manager will have extensive
educational and management- level nuclear power experience meeting
the criteria of ANSI N18.1-1971. The Operations Supervisor and
Nuclear Plant Supervisors maintain SRO licenses on Turkey Point. The
current Technical Specifications do not require the Operations
Manager to hold an SRO License at Turkey Point. The current
Technical Specifications permit the Operations Manager to have held
an SRO License on another plant. The proposed change will continue
to require that the Operations Manager has completed the Turkey
Point Nuclear Plant Senior Management Operations Training Course if
the incumbent did not previously hold an SRO license. The Turkey
Point Nuclear Plant Senior Management Operations Training Course
ensures that the Operations Manager has the training on plant-
specific systems
[[Page 31182]]
and procedures at Turkey Point and a knowledge level equivalent to
the license requirements for operations management.
The on-shift Operations' organization is, and will continue to
be, supervised and directed by the Operations Supervisor, who is
currently required by Technical Specification 6.2.2.h. to hold an
SRO License.
Additionally, the proposed changes do not impact or change, in
any way, the minimum on-shift manning or qualifications for those
individuals responsible for the actual licensed operation of the
facility as required by 10 CFR 50.54(l).
Based on the above, the proposed changes do not affect the
probability or consequences of accidents previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The change being proposed is administrative in nature, addresses
personnel qualification issues, does not affect assumptions
contained in plant safety analyses, the physical design and/or
operation of the plant, or Technical Specifications that preserve
safety analysis assumptions.
The proposed changes address organizational and qualifications
issues related to the criteria used for assignment of individuals to
the Operations organization off-shift management chain of command.
Since the proposed change does not impact or change, in any way, the
minimum on-shift manning or qualifications for those individuals
responsible for the actual licensed operation of the facility,
operation of the facility in accordance with the proposed amendment
would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed change addresses organizational and qualification
issues related to the criteria used for assignment of individuals to
the Operations organization off-shift management chain of command.
The proposed change does not impact or change, in any way, the
minimum on-shift manning or qualifications for those individuals
responsible for the actual licensed operation of the facility.
FPL's operating organization at Turkey Point Plant is shown on
Figure 1-2, Appendix A of the NRC-approved FPL Topical Quality
Assurance Report (TQAR). Since changes to the TQAR are governed by
10 CFR Sec. 50.54(a)(3), any changes to the TQAR that reduce
commitments previously accepted by the NRC require approval by the
NRC prior to implementation.
While the Operations Manager is responsible for the plant's
operating organization, his responsibilities also include management
of the plant's Health Physics and Chemistry departments. The
Operations organization is supervised and directed by the Operations
Supervisor, who is required by Technical Specification 6.2.2.h. to
hold a Senior Reactor Operator License. The Turkey Point Units 3 and
4 Technical Specifications do not require that the Operations
Manager maintain an SRO License (nor even that the incumbent has
ever held a Senior Reactor Operator License at Turkey Point). The
Turkey Point Technical Specification 6.3.1, FACILITY STAFF
QUALIFICATIONS, will ensure that, other than license certification,
the individual filling the Operations Manager position has the
requisite education, training, and experience for the management
position.
As a result, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: Frederick J. Hebdon
GPU Nuclear Corporation and Saxton Nuclear Experimental
Corporation, Docket No. 50-146, Saxton Nuclear Experimental
Facility (SNEF), Bedford County, Pennsylvania
Date of amendment request: February 2, 1996, as supplemented on
February 28, April 24 and May 24, 1996.
Description of amendment request: The proposed amendment would (1)
increase the scope of work permitted within the exclusion area at the
SNEF to include action preparatory to major component and facility
decommissioning limited to asbestos removal, removal of defunct plant
electrical services, and installation of decommissioning support
facilities and systems such as heating, ventilation, and air
conditioning,
(2) eliminate administrative access controls requiring that the
grating covering the auxiliary compartment stairwell and rod room
door remain locked except for authorized entry, and (3) revise the
facility layout diagram to allow the exclusion area to consist of,
at a minimum, the containment vessel, and at a maximum, extend to
the SNEF outer security fence, and to include on the diagram the
footprint of the proposed decommissioning support facilities.
Basis for proposed no significant hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve a significant hazards
considerations because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The SNEF ended power operation in May 1972, and the reactor core
has been removed. In its present condition, the only accidents
applicable to the site are fire, flooding, and radiological hazard.
The additional activities associated with the expansion of the
permissible work scope will not involve a significant increase in
the probability or consequences of a fire. There is no effect on the
probability or consequences of flooding nor would there be a
significant increase in the probability or consequences of an
offsite radiological hazard. The relocation of administratively
controlled accesses in accordance with the revised wording and the
proposed clarification of the facility layout diagram would have no
affect on analyzed accidents. Activities associated with the
construction of the decommissioning support facilities and the
existence of the completed buildings depicted on the revised figure
will not involve a significant increase in the probability or
consequences of a fire, flood, or radiological hazard. The proposed
changes identified by this technical specification change request do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
For the reasons discussed in 1 above, the possibility of a new
or different kind of accident from any accident previously evaluated
will not be created by the performance of the activities delineated
in the proposed revised technical specifications. There is similarly
no possibility of a new or different kind of accident from any
accident previously evaluated that would result from relocation of
administratively controlled accesses within the containment vessel;
from the flexibility to relocate/modify the exclusion area fence or
from the identification of the footprint, construction and existence
of the completed decommissioning support facilities.
3. Involve a significant reduction in a margin of safety.
For the reasons discussed in 1 above, none of the proposed
changes involve a significant reduction in a margin of safety.
The NRC staff has reviewed the analysis of the licensees and, based
on this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Saxton Community Library, 911
Church Street, Saxton, Pennsylvania 16678 Attorney for the Licensee:
Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge,
2300 N Street, NW, Washington, DC 20037
[[Page 31183]]
NRC Project Director: Seymour H. Weiss
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 20, 1996
Description of amendment request: The proposed amendment would
revise the Facility Operating License No. NPF-47 and Appendix C to the
license to reflect the name change from Gulf States Utilities Company
to Entergy Gulf States, Inc.
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:
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The proposed change documents changing the legal name of the
company. The proposed change will not affect any other obligations.
The company will still own all of the same assets, serve the same
customers, and all existing obligations and commitments will
continue to be honored.
Therefore, the proposed change does no significantly increase
the probability or consequences of an accident previously evaluated.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The administrative changes in the Operating License requirements
do not involve any change in the design of the plant.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
The proposed change is administrative in nature, as described
above, therefore, this change does not reduce the level of safety
imposed by any current requirements.
Therefore, the proposed changes do not cause 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: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: April 25, 1996
Description of amendment request: The change modifies the
calibration requirement for the source range monitors and intermediate
range monitors by noting that the sensors are excluded.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Pursuant to 10 CFR 50.92, NNECO has reviewed the proposed change
and concludes that the change does not involve a significant hazards
consideration (SHC) since the proposed change satisfies the criteria
in 10 CFR 50.92(c). That is, the proposed change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
By removing the requirement for sensor calibration the function
and safety performance of these systems will not be affected.
Existing surveillances, operator verification of overlap and system
interlocks ensure correct system performance without sensor
calibration.
Therefore, based on the above, the proposed change to the
Technical Specifications does not involve a significant increase in
the probability or consequences of any previously analyzed accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
This change does not cause the source range monitors (SRM) or
the intermediate range monitors (IRM) to function any differently
than intended by design and, therefore, does not create the
possibility of a new or different kind of accident. The Technical
Specification change deletes a Technical Specification requirement
which could not literally be complied with for one component and
that has no effect on the functional performance of the SRMs or
IRMs.
Therefore, this change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
This change corrects a Technical Specification requirement which
could not literally be complied with for one component and that has
no effect on the functional performance of the SRMs or IRMs.
Instrument calibrations and functional checks are still performed
during each refueling outage to assure adequate system performance.
Therefore, this change has no impact on the margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and Waterford Library, ATTN: Vince Juliano, 49 Rope
Ferry Road, Waterford, CT 06385.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: February 14, 1996
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant (DCPP), Unit Nos. 1 and 2, to revise 30 TS and add two new
TS surveillance requirements to support implementation of extended fuel
cycles at DCPP, Unit Nos. 1 and 2. The specific TS changes proposed
include those for 9 trip actuating device tests, 12 fluid system
actuation tests, and 11 miscellaneous tests. Two of the fluid system
actuation tests are proposed new TS surveillance requirements. The TS
changes also include the addition of a new frequency notation, ``R24,
REFUELING INTERVAL,'' to Table 1.1 of the TS. Also, a revision that
applies to all subsequent TS changes involves revising the Bases
section of TS 4.0.2 to change the surveillance frequency from an 18-
month surveillance interval to at least once each refueling interval.
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
[[Page 31184]]
consequences of an accident previously evaluated.
The surveillance interval notation addition in TS Table 1.1 and
the updated TS 4.0.2 Bases section are administrative changes that
do not affect the probability or consequences of accidents.
The 30 proposed TS surveillance interval increases from 18 to 24
months do not alter the intent or method by which the inspections,
tests, or verifications are conducted, do not alter the way any
structure, system, or component functions, and do not change the
manner in which the plant is operated. The surveillance,
maintenance, and operating histories indicate that the equipment
will continue to perform satisfactorily with longer surveillance
intervals. Few surveillance and maintenance problems were
identified. No problems recurred, with the exception of those
associated with the pressurizer heater emergency breakers, which
will continue to be surveilled on a quarterly frequency until they
are replaced.
There are no known mechanisms that would significantly degrade
the performance of the evaluated equipment during normal plant
operation. All potential time-related degradation mechanisms have
insignificant effects in the timeframe of interest (24 months +25
percent, or 30 months). Based on the past performance of the
equipment, the probability or consequences of accidents would not be
significantly affected by the proposed surveillance interval
increases.
The 24-month surveillance intervals for the two new TS proposed
to verify that the CCW [component cooling water] and ASW [auxiliary
saltwater] pumps will start automatically are based on an evaluation
of historical operation, maintenance, and surveillance data for the
pumps. These historical data are available because the pumps have
been operated, maintained, and tested on 18- month intervals in
accordance with procedures since initial plant startup. These new
surveillances represent additional TS requirements to ensure the CCW
and ASW pumps start when required. No known degradation mechanisms
would significantly affect the ability of the pumps to start over
the timeframe of interest (30 months maximum). Based on the past
performance of the equipment, these proposed new TS would not affect
the probability or consequences of accidents.
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.
The surveillance interval notation addition in TS Table 1.1 and
the updated TS 4.0.2 Bases section are administrative changes that
do not affect the type of accidents possible.
For the 30 proposed TS changes involving surveillance interval
increases from 18 to 24 months, the surveillance and maintenance
histories indicate that the equipment will continue to effectively
perform its design function over the longer operating cycles.
Additionally, the increased surveillance intervals do not result in
any physical modifications, affect safety function performance or
the manner in which the plant is operated, or alter the intent or
method by which surveillance tests are performed. Only a few
problems have been identified and generally have not recurred. All
potential time-related degradations have insignificant effects in
the timeframe of interest. The proposed surveillance interval
increases would not affect the type of accidents possible.
The 24-month surveillance intervals for the two new TS proposed
to verify starting of the CCW and ASW pumps are based on an
evaluation of historical operation, maintenance, and surveillance
data. These new TS represent additional requirements to ensure the
CCW and ASW pumps start when required. No known degradation
mechanisms would significantly affect the ability of the pumps to
start over the timeframe of interest. These proposed new TS would
not affect the type of accidents possible.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The surveillance interval notation addition in TS Table 1.1 and
the updated TS 4.0.2 Bases section are administrative changes that
do not affect the margin of safety.
For the 30 proposed TS changes involving 18- to 24-month
surveillance interval increases, evaluation of historical
surveillance and maintenance data indicates there have been only a
few problems experienced with the evaluated equipment.
There are no indications that potential problems would be cycle-
length dependent or that potential degradation would be significant
for the timeframe of interest and, therefore, increasing the
surveillance interval will have little, if any, impact on safety.
There is no safety analysis impact since these changes will have no
effect on any safety limit, protection system setpoint, or limiting
condition for operation, and there are no hardware changes that
would impact existing safety analysis acceptance criteria. Safety
margins would not be significantly affected by the proposed
surveillance interval increases.
As previously noted, the 24-month surveillance intervals for the
two new TS are based on an evaluation of historical data, represent
additional requirements, and are not believed to be significantly
affected by potential time-dependent degradation. As such, these
proposed new TS would not affect any margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: May 9, 1996
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 by revising Technical Specifications (TS)
3/4.3.2, ``Engineered Safety Features Actuation System
Instrumentation,'' and 3/4.6.2, ``Containment Spray System.'' The
changes would clarify the description of the initiation signal required
for operation of the containment spray system at Diablo Canyon Power
Plant (DCPP) and correctly incorporate changes made in previous license
amendments. All of the changes are administrative in nature.
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.
Revising the description of the containment spray (CS)
initiating signal clarifies the design of the plant and provides
uniformity across the Technical Specifications (TS) associated with
the CS initiation function. The enhanced description does not affect
system operation or performance, nor the probability of any event
initiators. The changes do not affect any engineered safety feature
actuation setpoints or accident mitigation capabilities.
The administrative changes to TS 3/4.3.2, Table 4.3-2, correct
the column headings and restore test frequency notation. The changes
only revise the TS to correspond with previously issued license
amendments (LAs).
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.
[[Page 31185]]
The administrative changes in the description of the CS
initiating signal provide uniformity across the TS associated with
the spray system. There are no design, operation, maintenance, or
testing changes associated with the administrative changes.
The administrative changes to TS 3/4.3.2, Table 4.3-2, correct
the column headings and restore test frequency notation. The changes
only revise the TS to correspond with previously issued LAs.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The administrative changes in CS signal description are not
associated with any design, operation, maintenance, or testing
revisions.
The administrative changes to TS 3/4.3.2, Table 4.3-2, correct
the column headings and restore test frequency notation. The changes
only revise the TS to correspond with previously issued LAs.
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of amendment request: May 20, 1996 (TS 373)
Description of amendment request: The proposed amendment revises
the technical specifications to incorporate a 24-hour delay in
implementing the action requirements due to a missed surveillance
requirement when the action requirements provide a restoration time
that is less than 24 hours. This change also clarifies that the time
limit of the action requirements applies from the point in time it is
identified a surveillance has not been performed and not at the time
that the allowed surveillance interval was exceeded. The licensee
claims this amendment is consistent with generic guidance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment to TS definition 1.0.LL is in accordance
with the guidance of GL 87-09 and NUREG 1433, Revision 1. The
proposed change will allow BFN to continue operation for an
additional 24 hours after discovery of a missed surveillance. The
change being proposed does not affect the precursor for any accident
or transient analyzed in Chapter 14 of the BFN Updated Final Safety
Analysis Report. The proposed change does not reflect a revision to
the physical design and/or operation of the plant. Therefore,
operation of the facility in accordance with the proposed change
does not affect the probability or consequences of an accident
previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment to TS definition 1.0.LL is in accordance
with the guidance of GL 87-09 and NUREG 1433, Revision 1. The
proposed change will allow the plant to continue operation for an
additional 24 hours after discovery of a missed surveillance. The
change being proposed will not change the physical plant or the
modes of operation defined in the facility license. The change does
not involve the addition or modification of equipment, nor do they
alter the design or operation of plant systems. Therefore, operation
of the facility in accordance with the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed amendment to TS definition 1.0.LL is in accordance
with the guidance of GL 87-09 and NUREG 1433, Revision 1. The
proposed change does not affect plant safety analysis or change the
physical design or operation of the plant. The proposed change will
allow the plant up to 24 hours to perform a missed surveillance. The
overall effect is a net gain in plant safety by avoiding unnecessary
shutdowns and the associate system transients due to missed
surveillance. Therefore, operation of the facility in accordance
with 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: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: May 8, 1996
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
5.3, ``Reactor,'' and TS 5.4, ``Fuel Storage,'' by removing the
enrichment limit for reload fuel and imposing fuel storage restrictions
on the spent fuel storage racks and the new fuel storage racks. The
revised TS are structured consistent with the Westinghouse Standard
Technical Specifications and the fuel storage restrictions are based on
the criticality analyses used to support TS Amendment 92 dated March 7,
1991.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to determine that no significant hazards
exist. The proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The criticality analysis which was performed in support of
Technical Specification Amendment 92, dated March 7, 1991,
demonstrated that adequate margins to criticality can be maintained
with fuel enrichments up to 49.2 grams of U235 per axial
centimeter stored in the New Fuel Storage Racks and enrichments up
to 52.3 grams of U235 per axial centimeter stored in the Spent
Fuel Storage Racks.
The bounding cases of the analysis demonstrated that keff
remains less than 0.95 in the Spent Fuel Storage Racks and the New
Fuel Storage Racks if flooded with unborated water. The bounding
cases of the analysis also demonstrated that keff remains less
than 0.98 in the New Fuel Storage Racks if moderated by optimally
misted moderator. Therefore, the 49.2 grams of U235 per axial
centimeter enrichment is acceptable for storage in the New Fuel
Storage Racks and 52.3 grams of U235 per axial centimeter for
storage in the Spent Fuel Storage Racks.
The only other accident that needs to be considered is a fuel
handling accident. Since the mass of the fuel assembly would not be
appreciably altered by the increased fuel
[[Page 31186]]
enrichment, the probability of this accident occurring is not
changed. The consequences of a fuel handling accident also would not
be affected by the use of higher fuel enrichment since the fission
product inventories in a fuel assembly are not a significant
function of initial fuel enrichment. This accident was analyzed in
the criticality analysis which was performed in support of Technical
Specification Amendment 92, dated March 7, 1991.
It should be noted that any changes in the nuclear properties of
the reactor core that may result from higher fuel enrichments would
be analyzed in the appropriate reload analysis.
The administrative relocation of information to licensee
controlled documents (i.e., USAR) conforms to NRC policy for the
content of technical specifications and does not increase the
probability or consequences of an accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
As discussed above, the only safety issue significantly affected
by the proposed change is the criticality analysis of the Spent Fuel
Storage Racks and the New Fuel Storage Racks. Since it has been
demonstrated that kG2eff remains below 0.95 and
0.98, respectively, in those areas, no new or different accident
would be created through the use of fuel enrichments up to 52.3
grams of U235 per axial centimeter at the Kewaunee Nuclear
Power Plant. Administrative controls will ensure that only fuel
enriched to 49.2 grams of U235 per axial centimeter or less
will be placed into the New Fuel Storage Racks.
The relocation of information to licensee controlled documents
does not create the possibility of a new or different kind of
accident.
3. Involve a significant reduction in the margin of safety.
Since the criticality analyses have shown that increasing the
allowable weight percent enrichment to 52.3 grams of U235 per
axial centimeter would not increase keff above 0.95 in the
Spent Fuel Storage Racks and increasing the allowable weight percent
enrichment to 49.2 grams of U235 per axial centimeter would not
increase keff above 0.98 in the New Fuel Storage Racks, it is
concluded that this proposed change would not reduce the margin of
safety. Any changes in the nuclear properties of the reactor core
that may result from higher fuel enrichments would be analyzed in
the appropriate reload analysis to ensure compliance with applicable
reload considerations and requirements.
Relocation of information to licensee controlled documents is an
administrative action and therefore does not reduce the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497
NRC Project Director: Gail H. Marcus
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.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 17, 1996
Description of amendment request: The proposed amendments would
modify Technical Specification Section 3/4.4.5, Steam Generators, 3/
4.4.6, Reactor Coolant System Leakage, and associate Bases to allow the
installation of tube sleeves as an alternative to plugging to repair
defective steam generator tubes.
Date of individual notice in the Federal Register: May 29, 1996 (61
FR 26936)
Expiration date of individual notice: June 28, 1996
Local Public Document Room location: Wharton County Junior
College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX
77488 Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: April 24, 1996
Brief description of amendment request: The proposed amendment
would modify Technical Specifications (TSs) 5.3.1 and 6.9.3.2 to
reflect use of new fuel obtained from ABB/Combustion Engineering, and
to incorporate staff-approved core reload analysis computer programs
(codes). Date of individual notice in Federal Register: May 1, 1996 (61
FR 19326)
Expiration date of individual notice: May 31, 1996
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
[[Page 31187]]
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: January 5, 1996, as
supplemented by letters dated April 19, May 1, and May 10, 1996.
Brief description of amendments: The amendments revise the
operating licenses and Technical Specification (TS) Section 1.26 to
increase the authorized rated thermal power. The amendments also revise
TS 4.1.1.4, 3.1.3.4, and 3.2.6 (Figure 3.2-1) to lower the allowable
reactor coolant system cold leg temperature limits for each of the
three Palo Verde Nuclear Generating Station units, and TS 3.4.2.1 and
3.4.2.2 to lower the pressurizer safety valve setpoints for Units 1 and
3 to support the increased power operation. The Unit 2 pressurizer
safety valve setpoints in TS 3.4.2.1 and 3.4.2.2 were revised in
Amendment 78, approved March 28, 1995, to the same values being
requested for Units 1 and 3 in this submittal.
Date of issuance: May 23, 1996
Effective date: May 23, 1996, to be implemented for Unit 1 within
30 days of issuance; to be implemented for Unit 2 within 30 days of
issuance; to be implemented for Unit 3 within 45 days as of the date of
issuance, except for the pressurizer safety valve setpoints change
which are effective prior to startup from Unit 3's sixth refueling
outage.
Amendment Nos.: Unit 1 - 108; Unit 2 - 100; Unit 3 - 80
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7544) The April 19, May 1, and May 10, 1996, supplemental letters
provided additional clarifying information and did not change the
initial no significant hazards consideration determination. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated May 23, 1996. No significant hazards
consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: January 31, 1996.
Brief description of amendment: This amendment revises the
Technical Specifications Section 4.4 to allow the use of 10 CFR Part
50, Appendix J, Option B, Performance-Based Containment Leakage Rate
Testing.
Date of issuance: May 28, 1996
Effective date: May 28, 1996
Amendment No. 169
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7545) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 28, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: November 15, 1995, as
supplemented by letters dated March 15, and April 10, 1996
Brief description of amendments: The amendments revise the
Technical Specifications and the associated Bases to increase the
setpoint tolerance of the main steam safety valves (MSSVs) from plus or
minus 1% to plus or minus 3%, to incorporate a requirement to reset the
as-left MSSV lift settings to within plus or minus 1% following
surveillance testing, and to delete two obsolete footnotes.
Date of issuance: May 31, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 146 and 140
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65676). The March 15 and April 10, 1996 letters provided clarifying
information that did not change the scope of the November 15, 1995
application and the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated May 31, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: January 12, 1995, as
supplemented by letter dated June 29, 1995
Brief description of amendments: The amendments revise and clarify
portions of Technical Specification Section 6.0, ``Administrative
Controls.''
Date of issuance: May 30, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 145 and 139
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 24, 1995 (60
FR 58109) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 30, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: April 3, 1996
Brief description of amendments: The amendments revise the
Technical Specifications and the associated Bases to provide that if
neither Train A or Train B of the hydrogen igniter is operable in any
one containment region, there is an allowance of 7 days to restore one
hydrogen igniter to operable status, or be in hot shutdown within the
next 6 hours.
Date of issuance: June 3, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 147 and 141
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 16, 1996 (61 FR
16649) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 3, 1996 No significant
hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
[[Page 31188]]
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 19, 1995, as supplemented by letter
dated December 7, 1995
Brief description of amendment: The amendment revised the
recombiner surveillance requirements to conform with the staff guidance
provided in NUREG-1432, ``Standard Technical Specifications Combustion
Engineering Plants.''
Date of issuance: June 5, 1996
Effective date: June 5, 1996
Amendment No.: 119
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
180) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 5, 1996. No significant hazards
consideration comments received: No
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: January 4, 1996
Brief description of amendments: These amendments rectify a
discrepancy in Technical Specification 3.5.3, and provide assurance
that administrative controls for High Pressure Safety Injection pumps
remain effective in the lower operational modes.
Date of Issuance: May 30, 1996
Effective Date: May 30, 1996
Amendment Nos.: 143 and 183
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5813) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 30, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: November 22, 1995
Brief description of amendments: These amendments upgrade existing
TS 3/4.4.6.1 for the Reactor Coolant System Leakage Detection Systems
by adopting the Standard Technical Specifications for Combustion
Engineering Plants to both St. Lucie Units.
Date of Issuance: May 30, 1996
Effective Date: May 30, 1996
Amendment Nos.: 144 and 84
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1629) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 30, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: March 28, 1996 (TSCR 234)
Brief description of amendment: The amendment modifies Technical
Specification pages 3.1-5 and 3.1-16 to indicate 40 percent of the
rated reactor thermal power as the anticipatory reactor scram bypass
setpoint on turbine trip or generator load rejection.
Date of Issuance: June 4, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 184
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18167) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated June 4, 1996 No significant
hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Houston Lighting & Power Company, City Public Service Board of San
Antonio Central Power and Light Company, City of Austin, Texas,
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County,
Texas
Date of amendment request: January 22, 1996, as supplemented April
4 and May 2, 1996
Brief description of amendment: The amendment modified the steam
generator tube plugging criteria in TS 3/4.4.5, Steam Generators, the
allowable primary-to-secondary leakage in TS 3/4.4.6.2, Operational
Leakage, and the associated Bases. These changes allowed the
implementation of alternate steam generator tube plugging criteria for
the tube support plate/tube intersections for Unit 1.
Date of issuance: May 22, 1996
Effective date: May 22, 1996
Amendment No.: 83
Facility Operating License No. NPF-76. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 16, 1996 (61 FR
16651) as corrected April 22, 1996 (61 FR 17735). The additional
information contained in the supplemental letter dated May 2, 1996, was
clarifying in nature and thus, within the scope of the initial notice
and did not affect the staff's proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated May 22, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,
Linn County, Iowa
Date of application for amendment: July 21, 1995, as supplemented
August 8, 1995 and December 15, 1995
Brief description of amendment: The amendment made administrative
changes to various sections of the DAEC Technical Specifications (TS).
The amendment replaced the surveillance condition when an Emergency
Service Water pump or loop is inoperable with an OPERABILITY
verification of the opposite train's Emergency Diesel Generator (EDG).
The amendment modified the TS to allow credit for demonstration of EDG
OPERABILITY that occurred within the previous 24 hours. The amendment
revised the format and language of TS Section 5.5
[[Page 31189]]
to clarify the requirements and state the capacity of the spent fuel
pool and vault storage in order to remove ambiguities in the wording
and to be more consistent with the Improved Standard TS guidance. The
amendment revised the list of Operations Committee responsibilities
(Section 6.5.1.6) to eliminate Committee review of procedures
implementing Security and Emergency Plans.
Date of issuance: June 5, 1996
Effective date: June 5, 1996
Amendment No.: 214
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49938) and February 2, 1996 (61 FR 3953) The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
June 5, 1996. No significant hazards consideration comments received:
No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of application for amendments: May 4, 1995, as supplemented
November 27, 1995, and March 1, 1996
Brief description of amendments: The amendments revise the
pressurizer and main steam safety valve lift setting tolerance from
plus or minus 1 percent to plus or minus 3 percent (as-found setpoint
only), revise the safety limit curves, reformat Section 2, and correct
typographical errors.
Date of issuance: May 21, 1996 Effective date: May 21, 1996, with
full implementation within 30 days
Amendment Nos.: Unit 1 - 123, Unit 2 - 116
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47621) The November 27, 1995, and March 1, 1996, letters provided
clarifying information in response to NRC staff questions. This
information was within the scope of the original application and did
not change the staff's initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated May 21, 1996. No
significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: March 13, 1996
Brief description of amendments: These amendments delete the
requirement in Technical Specifications (TS) 4.0.5a for NRC written
approval prior to implementation of relief from ASME Code requirements
by deleting ``...(g),.except where specific written relief has been
granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).'' Also,
the amendments add the ASME Section XI definition of ``Biennially or
every 2 years - At least once per 731 days,'' in TS 4.0.5b.
Date of issuance: May 28, 1996
Effective date: May 28, 1996
Amendment Nos.: Unit 1 - 112; Unit 2 - 110
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18173) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 28, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: April 3, 1996
Brief description of amendments: These amendments revise the
combined Technical Specifications (TS) for the Diablo Canyon Nuclear
Power Plant, Unit Nos. 1 and 2 to revise Technical Specifications 3/
4.7.5, ``Control Room Ventilation System;'' 3/4.7.6, ``Auxiliary
Building Safeguards Air Filtration System;'' and 3/4.9.12, ``Fuel
Handling Building Ventilation System'' to clarify the testing
methodology utilized by PG&E to determine the operability of the
charcoal and high efficiency particulate air (HEPA) filters in the
engineering safeguards features (ESF) air handling units at the Diablo
Canyon Power Plant (DCPP).
Date of issuance: May 28, 1996
Effective date: May 28, 1996
Amendment Nos.: Unit 1 - 113; Unit 2 - 111
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18173) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 28, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of application for amendment: May 8, 1996, as supplemented May
10, 1996, and May 29, 1996, and June 3, 1996.
Brief description of amendment: This amendment modifies the
Technical Specifications to correct several typographical errors that
were implemented in the Improved Technical Specifications at Ginna
Station per Amendment No. 61.
Date of issuance: June 3, 1996
Effective date: As of date of issuance.
Amendment No.: 65
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes (61 FR 24965, dated May 17,
1996). That notice provided an opportunity to submit comments on the
Commission's proposed no significant hazards consideration
determination. No comments have been received. The notice published May
17, 1996, also provided for a hearing by June 17, 1996, but indicated
that if a Commission makes a final no significant hazards consideration
determination, any such hearing would take place after issuance of the
amendment. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 3, 1996.
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
[[Page 31190]]
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: February 9, 1996 as superseded
by letter dated March 22, 1996.
Brief description of amendment: The amendment revises Technical
Specification (TS) 1.7, 4.6.1.1, 3.6.1.3, 4.6.1.3, 6.8.4 and the
associated Bases section to directly reference Regulatory Guide 1.163,
``Performance-Based Containment Leak Test Program,'' as required by 10
CFR 50, Appendix J, Option B for the Type A containment integrated leak
rate tests and the Type B and C local leak tests.
Date of issuance: May 28, 1996
Effective date: May 28, 1996, to be implemented within 30 days from
the date of issuance.
Amendment No.: 111
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18174) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 28, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: January 30, 1996
Brief description of amendments: The amendments modify the
Technical Specifications to increase the minimal allowable reactor
coolant system total flow rate.
Date of issuance: June 5, 1996
Effective date: June 5, 1996
Amendment Nos.: 201 and 182
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7559) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 5, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: April 24, as supplemented by
letter dated May 29, 1996.
Brief description of amendment: The amendment would modify the WNP-
2 technical specifications to support Cycle 12 operation, reflect use
of new fuel obtained from ABB/Combustion Engineering, and incorporate
staff-approved core reload analysis computer programs (codes). Date of
issuance: June 4, 1996 Effective date: June 4, 1996, to be implemented
within 30 days of issuance.
Amendment No.: 146
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 1, 1996 (61 FR
19326). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 4, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
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
[[Page 31191]]
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 July 19, 1996, 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. 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. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-001, 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).
Commonwealth Edison Company, Docket No. 50-249, Dresden Nuclear
Power Station, Unit No. 3
Date of application for amendment: May 22, 1996
Brief description of amendment: The amendment authorizes, on a one-
time temporary basis, operation of Dresden, Unit 3, with the structural
steel members in the Low Pressure Coolant Injection (LPCI) corner rooms
outside the Updated Final Safety Analysis Report (UFSAR) design
parameters, but capable of performing their intended safety function.
Following a reactor scram on May 15, 1996, Commonwealth Edison Company
(ComEd) performed a Safety Evaluation (SE) in accordance with the
requirements of 10 CFR 50.59 to determine if the current configuration
of the corner room structural steel members had reduced the margin of
safety as described in the UFSAR. The SE determined that the
configuration does not reduce the margin of safety with respect to the
stress allowables for the structural steel if subjected to a Safe
Shutdown Earthquake (SSE). An unreviewed safety question was determined
to exist because stress allowables for the structural steel subjected
to an Operating Basis Earthquake (OBE) were found outside the UFSAR
requirements; however, the current configuration of the corner room
structural steel members has not
[[Page 31192]]
significantly reduced the margin of safety as described in the UFSAR.
Date of Issuance: May 31, 1996 Effective date: May 31, 1996
Amendment No.: 144
Facility Operating License No. DPR-25. The amendment revised the
license.
Press release issued requesting comments as to proposed no
significant hazards consideration: Yes. Joliet Herald News on May 25,
1996, and the Morris Daily Herald on May 29, 1996. Comments received:
No comments were received on the proposed no significant hazards
consideration determination; however, comments were received concerning
the licensee's timeliness and decision-making in restoring the UFSAR
design margin to the structural steel members installed the LPCI corner
rooms at Dresden Unit 3.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the State of Illinois and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated May 31, 1996.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
NRC Project Director: Robert A. Capra
Dated at Rockville, Maryland, this 12th day of June 1996.
For the Nuclear Regulatory Commission
John A. Zwolinski,
Deputy Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation
[Doc. 96-15398 Filed 6-18-96; 8:45 am]
BILLING CODE 7590-01-F
