[Federal Register Volume 61, Number 109 (Wednesday, June 5, 1996)]
[Notices]
[Pages 28604-28626]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10605]
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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 11, 1996, through May 23, 1996. The last
biweekly notice was published on May 22, 1996 (61 FR 25696).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards onsideration
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 5, 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
[[Page 28605]]
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: May 1, 1996
Description of amendment request: The proposed amendment will
relocate the administrative controls related to the quality assurance
review and audit requirements of Section 6 from the Pilgrim Station
Technical Specifications to the Boston Edison Quality Assurance Manual.
This change is in accordance with the guidance contained in NRC
Administrative Letter 95-06, ``Relocation of Technical Specification
Administrative Controls Related to Quality Assurance.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The change will relocate the administrative controls related to
the quality assurance review and audit requirements from the
technical specifications to the quality assurance plan. These
changes are administrative in nature and do not impact initiators of
analyzed events, accident mitigation capabilities, or transient
events. The quality assurance program is a logical candidate for
such relocation due to the controls imposed by such regulations as
Appendix B to 10 CFR [Part] 50, the existence of NRC approved
quality assurance plans and commitments to industry quality
assurance standards, and the established quality assurance program
change control process in 10 CFR 50.54(a). Therefore, the changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The change will relocate the administrative controls related to
the quality assurance review and audit requirements from the
technical specifications to the quality assurance plan. The quality
assurance program is a logical candidate for such relocation due to
the controls imposed by such regulations as Appendix B to 10 CFR
[Part] 50, the existence of NRC approved quality assurance plans and
commitments to industry quality assurance standards, and the
established quality assurance program change control process in 10
CFR 50.54(a). The proposed changes do not involve a physical
alteration of the plant or changes in methods governing plant
operation. The changes will not impose or eliminate any new or
different requirements. Therefore the changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The change will relocate the administrative controls related to
the quality assurance review and audit requirements from the
technical specifications to the quality assurance plan. These
changes are administrative in nature. The quality assurance program
is a logical candidate for such relocation due to the controls
imposed by such regulations as Appendix B to 10 CFR [Part] 50, the
existence of NRC approved quality assurance plans and commitments to
industry quality assurance standards, and the established quality
assurance program change control process in 10 CFR 50.54(a). The
proposed change will not reduce a margin of safety because it has no
impact on any safety analysis assumptions. Therefore, the operation
of PNPS [Pilgrim Nuclear Power Station] in accordance with the
proposed license amendment will not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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
[[Page 28606]]
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: May 1, 1996
Description of amendment request: The proposed amendment will
reflect the implementation of 10 CFR Part 50, Appendix J, Option B at
the Pilgrim Nuclear Power Station.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes do not involve any physical or operational
changes to structures, systems or components. The proposed changes
provide a mechanism within the TS [Technical Specifications] for
implementing a performance-based leakage rate test program which was
promulgated by the revision to 10CFR50 to incorporate Option B into
Appendix J. The TS Limiting Conditions for Operation (LCO) remain
unaffected by these changes. Thus, the safety design basis for the
accident mitigation functions of the primary containment is
maintained. Therefore, these changes will not increase the
probability or consequences of an accident previously evaluated.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Revising surveillance requirement acceptance criteria and
frequencies does not physically modify the plant and does not modify
the operation of any existing equipment. Further, the TS LCOs remain
unaffected by these changes.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
The proposed changes do not involve a significant reduction in
the margin of safety, nor do they affect a safety limit, an LCO, or
the manner in which plant equipment is operated. The NRC letter
dated November 2, 1995, recognizes that changes similar to the
proposed changes are required to implement Option B of 10CFR50,
Appendix J. In NUREG-1493, ``Performance-Based Containment Leak-Test
Program,'' which forms the basis for the Appendix J revision, the
NRC concludes that adoption of performance-based test intervals for
Appendix J testing will not significantly 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 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
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: May 1, 1996
Description of amendment request: The proposed amendment would
modify the definition of ``Core Alteration,'' and the Limiting
Condition for Operation, Surveillance conditions and Bases section
associated with Technical Specification (TS) 3.7.C, ``Secondary
Containment.''
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:
Proposed Change 1: Definition of ``Alteration of the
Reactor Core
The definition, ``Alteration of the Reactor Core'', is being
revised so that the term will apply only to those activities that
create the potential for a reactivity excursion and, therefore,
warrant special precautions or controls in the TS. The proposed
definition includes normal control rod movement in the definition,
but excludes control rod drive movement (such as rod removal from
the core) when all four fuel bundles surrounding a control rod are
removed. The proposed change does not increase the probability or
consequences of an accident because the proposed definition, by
identifying activities with the potential for causing a reactivity
excursion, ensures that the additional precautions and controls in
the TS are implemented at all appropriate times. In addition, the
movement of components excluded by this definition is not assumed in
the initiation of any analyzed event. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Proposed Change 2: Secondary Containment
The current specifications are revised to specify more clearly
when secondary containment is required, what actions to take if
secondary containment is inoperable, and time frames for completing
the actions. These revisions enhance the existing specification and
serve to make it more definitive by encompassing the conditions
currently specified by TS and supplementing them to specify other
conditions when secondary containment is required.
Surveillances 4.7.C.1.a and b were only necessary during initial
and Cycle 1 operations. Removing obsolete information from the
existing specifications, re-numbering and re-arranging the wording
is an administrative change.
These changes are administrative in nature and do not impact
initiators of analyzed events, accident mitigation capabilities, or
transient events. Therefore, the changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The 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:
Proposed Change 1: Definition of ``Alteration of the
Reactor Core
The definition change specifies more accurately which component
movements constitute a ``Core Alteration''. This change does not
involve a physical alteration of the plant (no new or different type
of equipment will be installed) or changes in methods governing
normal plant operation. The proposed changes will allow movement of
some components (camera, lights, etc.) during times when ``Core
Alterations'' have been halted since these components will not
affect core reactivity. Removal of a control rod involves unlatching
and withdrawal/insertion from over-vessel handling equipment. These
activities necessitate, by design, the removal of the adjacent four
fuel assemblies. With this configuration (no fuel in the cell;
handling the associated control rod), the proposed change will allow
movement of a ``reactivity control component'' while not imposing
requirements unique to ``Core Alterations'' (note: other
requirements, such as those for handling loads over irradiated fuel,
will remain applicable). The reactivity effects of this control rod
movement are more than compensated for by the initial removal of the
fuel assemblies. Therefore, this change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Proposed Change 2: Secondary Containment
The proposed change does not eliminate or relax any existing TS
condition. Rather, it better defines when secondary containment is
required, provides action statements for inoperability and removes
obsolete
[[Page 28607]]
requirements (from first operating cycle). This change does not
involve a physical change to structures, systems or components, and
the safety design bases for the accident mitigating function of the
secondary containment is maintained. Therefore, these changes will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in 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:
Proposed Change 1: Definition of ``Alteration of the
Reactor Core
The proposed definition more accurately identifies those
activities with the potential for causing a reactivity excursion.
The more accurate identification of ``Core Alterations'' will ensure
that when there is a potential for reactivity excursions,
appropriate precautions are applied. The components now excluded
from the proposed definition are those that do not have the
capability for adversely impacting core reactivity. The proposed
change has no impact on safety analysis assumptions. Therefore, the
change will not involve a significant reduction in a margin of
safety.
Proposed Change 2: Secondary Containment
The proposed additions of applicability conditions provide a
more precise understanding of when secondary containment integrity
is required and what actions to take if it becomes inoperable. The
change does not eliminate any existing conditions. The deletion of
surveillances applicable only for the first operating cycle and re-
numbering and re-arranging the remaining surveillance wording is an
administrative change and has no impact on the operation of the
plant or mitigation of accidents. Therefore, the operation of the
facility in accordance with this 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: 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
Carolina Power & Light Company, et al., Docket No. 50-325,
Brunswick Steam Electric Plant, Unit 1, Brunswick County, North
Carolina
Date of amendment request: April 8, 1996
Description of amendment request: The licensee has proposed to
revise the Technical Specifications (TS) to include the following
changes: 1. The Minimum Critical Power Ratio (MCPR) Safety Limit
specified in TS 2.1.2 from 1.07 to 1.09 for Unit 1 Cycle 11 operation;
TS 5.3.1 to reflect the new fuel type (GE13) that will be inserted
during Unit 1 Refueling Outage 10; 2. The acceptable range of sodium
pentaborate concentration for the standby liquid control system shown
in TS Figure 3.1.5-1 to reflect changes to poison material
concentration needed to achieve reactor shutdown based on the new GE13
fuel type.
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.
Proposed Change 1
The proposed amendment will allow the loading and use of GE13
fuel assemblies in the Brunswick Unit 1 reactor core. The use of
GE13 fuel assemblies requires that the safety limit minimum critical
power ratio value also be revised. The safety limit minimum critical
power ratio is established to maintain fuel cladding integrity
during operational transients. The GE13 fuel assembly design has
been analyzed using methods that have been previously approved by
the Nuclear Regulatory Commission and documented in General Electric
Nuclear Energy's reload licensing methodology Topical Report (NEDE-
24011-P-A-11, ``General Electric Standard Application for Reactor
Fuel (GESTAR II)'' dated November 1995).
The proposed revision of the safety limit minimum critical power
ratio does not alter any plant safety-related equipment, safety
function, or plant operations that could change the probability of
an accident. The change does not affect the design, materials, or
construction standards applicable to the fuel bundles in a manner
that could change the probability of an accident.
A methodology that has been previously reviewed and accepted by
the Nuclear Regulatory Commission was used to derive both the
existing and updated safety limit minimum critical power ratio
value. The same methodology and criteria have been applied to derive
the existing safety limit minimum critical power ratio of 1.07 as
that used to derive the updated safety limit minimum critical power
ratio value of 1.09. The updated safety limit minimum critical power
ratio assures that fuel cladding protection equivalent to that
provided with the existing safety limit minimum critical power ratio
value is maintained. This ensures that the consequences of
previously evaluated accidents are not significantly increased.
Proposed Change 2
The standby liquid control system provides a means of reactivity
control that is independent of the normal reactivity control system.
The standby liquid control system must be capable of assuring that
the reactor core can be placed in a subcritical condition at any
time during reactor core life. Technical Specification Figure 3.1.5-
1 specifies the acceptable range of concentrations and volumes for
sodium pentaborate solution used as a neutron absorber (i.e., for
reactivity control). The portion of the sodium pentaborate
concentration range shown in Technical Specification Figure 3.1.5-1
applicable to the lower range of tank volumes is being revised to
increase the required concentration of sodium pentaborate solution.
This change is needed to account for the additional shutdown
reactivity needed based on the planned use of GE13 fuel assemblies
as reload fuel for the Unit 1 reactor core. Since the standby liquid
control system is independent from the normal means of controlling
reactor core reactivity and not used to control core reactivity
during normal plant operations, the proposed revision to the sodium
pentaborate concentration curve for the standby liquid control
system does not alter any plant safety-related equipment, safety
function, or plant operations that could change the probability of
an accident.
The current volume-concentration range of sodium pentaborate
used in the standby liquid control system will achieve a sufficient
concentration of boron in the reactor vessel to ensure reactor
shutdown. Based on the increased reactivity of the new GE13 reload
fuel assemblies, the required sodium pentaborate volume-
concentration range is being revised to ensure sufficient neutron
absorbing solution is available to achieve reactor shutdown;
therefore, the consequences of an accident previously evaluated are
not significantly increased.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Proposed Change 1
The GE13 fuel assembly has been designed and complies with the
acceptance criteria contained in General Electric Nuclear Energy's
standard application for reactor fuel (GESTAR-II), which provides
the latest acceptance criteria for new General Electric fuel
designs. The GE13 fuel assembly complies with GESTAR-II acceptance
criteria that have been previously reviewed and accepted by the
Nuclear Regulatory Commission. The similarity of the GE13 fuel
design to the previously accepted GE11 fuel design, in conjunction
with the increased critical power capability of the GE13 fuel
design, ensure that no new mode or condition of plant operation is
being authorized by the loading and use of the
[[Page 28608]]
GE13 fuel type. The proposed revision of the safety limit minimum
critical power ratio from 1.07 to 1.09 does not modify any plant
controls or equipment that will change the plant's responses to any
accident or transient as given in any current analysis. Therefore,
the proposed change to allow the loading and use of the GE13 fuel
type and the revision of the safety limit minimum critical power
ratio value from 1.07 to 1.09 will not create the possibility for a
new or different kind of accident from any accident previously
evaluated.
Proposed Change 2
As discussed above, the standby liquid control system provides a
means of reactivity control that is independent of the normal
reactivity control system and is capable of assuring that the
reactor core can be placed in a subcritical condition at any time
during reactor core life. The proposed revision to the sodium
pentaborate concentration range does not modify the standby liquid
control system or its controls, does not modify other plant systems
and equipment, and does not permit a new or different mode of plant
operation. As such, the proposed revision to the minimum pentaborate
concentration value does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
Proposed Change 1
As previously discussed, the GE13 fuel assembly design has been
analyzed using methods that have been previously approved by the
Nuclear Regulatory Commission and documented in General Electric
Nuclear Energy's reload licensing methodology Topical Report (NEDE-
24011-P-A-11, ``General Electric Standard Application for Reactor
Fuel (GESTAR II)'' dated November 1995). The safety limit minimum
critical power ratio value is selected to maintain the fuel cladding
integrity safety limit (i.e., that 99.9 percent of all fuel rods in
the core are expected to avoid boiling transition during operational
transients). Appropriate operating limit minimum critical power
ratio values are established, based on the safety limit minimum
critical power ratio value, to ensure that the fuel cladding
integrity safety limit is maintained. The operating limit minimum
critical power ratio values are incorporated in the Core Operating
limits Report as required by Technical Specification 6.9.3.1. The
new GE13 safety limit minimum critical power ratio value of 1.09 is
based on the same fuel cladding integrity safety limit criteria [as]
that for the GE11 safety limit minimum critical power ratio value of
1.07 (i.e., that 99.9 percent of all fuel rods in the core are
expected to avoid boiling transition during operational transients);
therefore, the proposed change does not result in a significant
reduction in the margin of safety.
Proposed Change 2
As previously stated, the purpose of the standby liquid control
is to inject a neutron absorbing solution into the reactor in the
event that a sufficient number of control rods cannot be inserted to
maintain subcriticality. Sufficient solution is to be injected such
that the reactor will be brought from maximum rated power conditions
to subcritical over the entire reactor temperature range from
maximum operating to cold shutdown conditions. General Electric
methodology establishes a fuel type dependent standby liquid control
system shutdown margin to account for calculational uncertainties.
General Electric calculations show that an in-vessel concentration
of 660 ppm will provide a standby liquid control system minimum
shutdown margin in excess of the 3.2%[delta]k value required for the
GE13 fuel. To achieve an in-vessel concentration of 660 ppm, the
acceptable range of standby liquid control system tank
concentrations is being revised for the lower range of tank volumes.
Thus, the proposed revision of the standby liquid control system
sodium pentaborate volume-concentration range ensures that there
will not be a significant reduction in the amount of available
shutdown margin and, therefore, not a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Eugene V. Imbro
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: February 27, 1996
Description of amendment request: The proposed license amendment
would modify the Action Statement of Technical Specification (TS)
3.7.1.1.1. Currently, the TS action statement requires that with the
self actuation function on one or more main steam line code safety
valves associated with an operating loop inoperable, the licensee must
restore the inoperable valve to operable status within 4 hours.
Otherwise, the plant must be in hot standby within the next 6 hours and
in hot shutdown within the following 30 hours. The proposed change will
allow continued power operation at reduced power levels with main steam
safety valves inoperable. The proposed change is consistent with the
philosophy of the Westinghouse Standard Technical Specifications.
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 to the Action Statement of LCO [Limiting
Condition for Operation] 3.7.1.1.1 will allow indefinite operation
at less than or equal to 75% power in the event that the self
actuation function of no more than one safety valve per steam
generator is inoperable, and allow indefinite operation at less than
or equal to 50% power in the event that the self actuation function
of no more than two safety valves per steam generator is inoperable.
The requirement to reduce power will ensure that there is no
increase in the consequences of a loss of load accident. The
proposed change is consistent with the methodology in the
Westinghouse Standard Technical Specifications. The methodology is
conservative, since the PORVs [power operated relief valves] cannot
affect the time of reactor trip on high pressurizer pressure. Thus,
it is concluded that the change does not increase the consequences
of any previously evaluated accident.
The change only specifies a power reduction in the event that
the self actuation function of steam generator safety valves is
inoperable. It does not affect the probability of any accident. The
change by itself does not affect the likelihood of an inoperable
safety valve.
2. [The proposed change does not create] the possibility of a
new or different kind of accident from any previously evaluated.
The change only specifies a power reduction in the event that
the self actuation function of steam generator safety valves is
inoperable. This does not create the potential for a new or
different kind of accident. The lower power level assures that peak
steam generator pressure and RCS [reactor coolant system] pressure
will remain below 110% of design. This provides assurance that no
equipment failure will occur due to overpressurization. Thus, the
change does not create the possibility for a new or different kind
of accident.
3. [The proposed change does not involve] a significant
reduction in a margin of safety.
The allowable power levels have been selected, consistent with
the Westinghouse Standard Technical Specifications, to assure that
steam generator and RCS pressure will remain below 110% of design.
Thus, there is no reduction in a margin of safety for overpressure
protection.
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
[[Page 28609]]
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
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: March 7, 1996
Description of amendment request: The licensee will be replacing a
locally operated (manual) containment sump suction isolation valve, RH-
V-808A, with a remote manually operated (motor operated) valve, RH-MOV-
808A during the upcoming Cycle 19 refueling outage. As a result,
changes are being requested to the Haddam Neck Plant Technical
Specifications to reflect this design change.
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 technical specification change to Section 3/4.4.6.2
and its bases are the replacement of the designation RH-V-808A with
RH-MOV-808A. There are no changes to the requirements of this
specification and this change is therefore an administrative change.
The changes to Section 3/4.5.1 will make the requirements for RH-
MOV-808A identical to those of RH-MOV-22. RH-V-808A is being
converted to a motor operated valve (MOV). This MOV will make the
ability to establish a suction path from the containment to the
Residual Heat Removal (RHR) System single failure proof from the
control room. Both RH-MOV-22 and RH-MOV-808A will be opened to
establish containment sump recirculation post-loss of coolant
accident (LOCA). This will provide added assurance that core cooling
will be maintained in the switch from injection to containment sump
recirculation following a LOCA. The requirement for RH-MOV-808A to
be closed and its hand wheel locked can not cause an accident. The
credit for operation of RH-MOV-808A to ensure that the establishment
of containment sump recirculation is single failure proof is
equivalent to the current crediting of RH-V-808A with the only
difference being that operation of the valve can now be performed
from the control room. Also, since both RH-MOV-22 and RH-MOV-808A
will be procedurally opened during establishment of containment sump
recirculation, the elimination of the requirement to lock open the
breaker for RH-MOV-22 will not affect the consequences of a LOCA.
The proposed changes that reflect the conversion of RH-V-808A to a
MOV and the proposed changes in how the valve is used do not
increase the consequences of a LOCA.
2. [The proposed change does not create] the possibility of a
new or different kind of accident from any previously evaluated.
The proposed changes will require RH-MOV-808A to be closed with
the hand wheel locked. This provides assurance that the valve is in
the required position. Also, RH-MOV-808A will be capable of remote
manual operation during the monthly surveillance which provides
assurance that the valve can be repositioned if necessary. The
proposed opening of RH-MOV-808A at the same time as RH-MOV-22 is
opened, provides greater assurance that a suction path is available
to the RHR pumps as well as lowering the total effective piping
resistance from the containment sump to the pump suction. Therefore,
the proposed changes do not introduce the possibility of a new or
different kind of accident.
3. [The proposed change does not involve] a significant
reduction in a margin of safety.
The proposed changes make RH-MOV-808A identical to RH-MOV-22
with the exception that RH-MOV-808A will not get a closure signal on
Safety Injection Actuation. Both RH-MOV-22 and RH-MOV-808A are
containment isolation valves in a closed system. For closed systems,
the containment isolation requirement is that the valves be either:
a) automatic, b) locked closed, or c) capable of remote manual
operation. RH-MOV-808A and RH-MOV-22 are both capable of remote
manual operation and therefore do not need automatic closure when
they are opened as part of the technical specification required
surveillance. Therefore, the proposed changes can 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: 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
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: March 28, 1996
Description of amendment request: The proposed license amendment
will add an additional footnote to Limiting Condition for Operation
(LCO) 3.4.2.1 and revise an existing footnote for LCO 3.4.2.2.
Currently, the footnote for LCO 3.4.2.2 requires the pressurizer code
safety valve as-found lift setting to be within +3 percent and -1
percent of the setpoint. The proposed change will relax the negative
as-found lift tolerance to -3 percent. The as-left lift tolerance will
remain as plus or minus 1 percent. The same footnote will be added to
LCO 3.4.2.1.
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 relax the pressurizer safety valve
negative as-found lift tolerance to -3 percent. The as-left lift
tolerance will remain as plus or minus 1 percent. This proposed
technical specification change will allow for the full use of the
plus or minus 3 percent as-found acceptance criterion for valve
testing consistent with 1989 ASME Section XI, Subsection IWV. The
relaxing of the as-found lift tolerance can not cause an accident.
The relaxing of the tolerance will allow the safety valve setpoint
to be closer to the Power Operated Relief Valve (PORV) setpoint and
could result in a slightly lower pressure for overheating events.
The analysis that takes credit for the increase in pressure to the
PORV setpoint is the Loss of Load analysis. The minimum departure
from nucleate boiling ratio (DNBR) was reanalyzed without taking any
credit for the transient increase in pressure. The minimum DNBR
still remains above the acceptance criterion as well as above the
limiting minimum DNBR predicted for all Updated Final Safety
Analysis Report Chapter 15 accidents. Also, the relaxed tolerance in
conjunction with a lower safety valve blowdown, yet still
conservative, results in a slightly higher average pressure for a
valve lift/reset cycle. This means that pressurizer overfill will
not be predicted for the limiting transient, loss of feedwater.
Thus, the proposed relaxation of as-found lift tolerance does not
increase the probability or consequences of the design basis
accidents previously evaluated.
2. [The proposed change does not create] the possibility of a
new or different kind of accident from any previously analyzed.
The proposed relaxation of the lift tolerance still requires the
safety valve lift setpoint to be above both the PORV setpoint and
the pressurizer high pressure reactor trip setpoint. In addition,
the as-left setpoint is not being changed. The relaxed tolerance in
combination with a conservative safety valve blowdown still will
preclude the prediction of water relief from the pressurizer. This
means that the proposed change does not introduce the possibility of
a new or different kind of accident.
[[Page 28610]]
3. [The proposed change does not involve] a significant
reduction in a margin of safety.
The proposed relaxation of the as-found lift tolerance for valve
testing is consistent with 1989 ASME Section XI, Subsection IWV. The
as-left lift tolerance will remain plus or minus 1 percent. In
addition, the design basis analyses still meet their acceptance
criteria with the -3 percent lift tolerance. Therefore, the proposed
change can not cause a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457.
Attorney for licensee: 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
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: April 16, 1996
Description of amendment request: The licensee is proposing to
revise the Technical Specifications to permit the Haddam Neck Plant to
remain in Mode 1, 2, 3, or 4 with the average water temperature of the
ultimate heat sink (UHS) greater than 90 deg. additional action has
been added which would require the plant to be placed in at least Hot
Standby within 6 hours and in Cold Shutdown within the following 30
hours upon identifying that the average water temperature of the UHS is
greater than 95 deg.F. In addition, the licensee is proposing to
include a new surveillance requirement for monitoring the average
circulating water inlet temperature to be within its limits when the
average water temperature of the UHS exceeds 89 deg.F.
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 addition to the Action Statement of LCO 3.7.12 of
an 8 hour period to monitor the average water temperature of the UHS
does not involve an increase in the probability of an accident
previously evaluated. The probability of an accident previously
evaluated is not increased by a short-term increase in the average
water temperature of the UHS. An evaluation of the service water
loads associated with the loss-of-offsite power and a coincident
worst case single failure of a diesel generator to start (resulting
in the loss of two of the four service water pumps) determined that
there is adequate margin to accomplish plant cooldown at a service
water inlet temperature of 95 deg.F. The recirculation phase of a
LOCA [loss-of-coolant accident] was evaluated to verify that
adequate flow would be available to the RHR [residual heat removal]
heat exchangers. The most limiting assumptions for the recirculation
phase are offsite power is available and one RHR heat exchanger
service water isolation valve fails to open. The injection phase of
a LOCA was evaluated to verify that adequate flow would be available
to the CAR [containment air recirculation] fan cooling coils. The
most limiting assumption for the injection phase is a loss-of-
offsite power. The results of these evaluations determined that
there is adequate service water flow to accomplish plant cooldown
with average water temperature of the UHS up to 95 deg.F. CYAPCO
[Connecticut Yankee Atomic Power Company] also proposes to include
an additional surveillance requirement to monitor the average water
temperature of the UHS at least once per hour if the average water
temperature of the UHS exceeds 89 deg.F. This additional
surveillance requirement ensures increased operator awareness as the
average water temperature of the UHS approaches the 90 deg.F LCO
limit. Based on the above, there is no significant increase in the
consequences of any 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 proposed technical specification changes do not create the
possibility of a new or different kind of accident from those
previously evaluated. The addition of an 8 hour time period to
monitor the average water temperature of the UHS increases from 6 to
14 hours the amount of time that is allowed before the plant must
proceed to Hot Standby should the average water temperature of the
UHS increase above 90 deg.F. This extension of the time allowed for
the plant to be in Hot Standby does not change the plant
configuration. CYAPCO also proposes to include an additional
surveillance requirement to monitor the average water temperature of
the UHS at least once per hour if the average water temperature of
the UHS exceeds 89 deg.F. This additional surveillance requirement
ensures increased operator awareness as the average water
temperature of the UHS approaches the 90 deg.F LCO limit.
As such, the changes do not create the possibility of a new or
different kind of accident from those previously evaluated.
3. [The proposed change does not] involve a significant
reduction in a margin of safety.
The proposed technical specification changes do not involve a
significant reduction in any margin of safety. The addition of an 8
hour time period to monitor the average water temperature of the UHS
increases from 6 to 14 hours the time required before the plant must
proceed to Hot Standby should the average water temperature of the
UHS temperature [exceed] 90 deg.F. An evaluation has been performed
to demonstrate that the risk significance associated with the
increased action time is very low. In addition, safe shutdown
capability has been demonstrated for service water inlet
temperatures as high as 95 deg.F. The addition of a surveillance
requirement to monitor the average water temperature of the UHS at
least once per hour if the average water temperature of the UHS
exceeds 89 deg.F is an additional requirement, limitation, or
restriction not currently within the technical specifications.
Therefore, these changes do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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
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 proposed amendment will allow
the use of the performance-based containment leakage testing
requirements described in 10 CFR Part 50, Appendix J, Option B.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The changes involved in this license amendment request revise
the testing criteria for the containment penetrations. The revised
criteria will be based on the guidance in Regulatory Guide 1.163,
``Performance-Based Containment Leak-Test Program.'' This guidance
allows for the use of relaxed testing frequencies for containment
penetrations that have performed satisfactorily on a historical
basis. The Containment Leakage Rate Testing Program considers the
type of service, the design of the penetration, and the safety
impact of the penetration in determining the
[[Page 28611]]
testing interval of each penetration. The NRC Staff has reviewed the
potential impact of performance-based testing frequencies for
containment penetrations during the development of the Option B
regulation. The NRC Staff review is documented in NUREG-1493,
``Performance-Based Containment Leakage-Test Program.'' The review
concluded that reducing the frequency of Type A tests (Integrated
Leakage Rate Tests) from three per 10 years to one per 10 years
leads to an imperceptible increase in risk. For Type B and C testing
(Local Leakage Rate Tests), the change in testing frequency should
not have significant impact since this leakage contributes less than
0.1 percent of the overall risk based on the existing regulations.
The use of Option B will allow the extension of testing intervals
with a minimal impact on the radiological release rates since most
penetration leakage is continually well below the specified limits.
In the accident risk evaluation, the NRC Staff noted that the
accident risk is relatively insensitive to the containment leakage
rate because the accident risk is dominated by accident sequences
that result in failure of or bypass of the containment. The use of a
performance-based testing program will continue to provide assurance
that the accident analysis assumptions remain bounding. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously analyzed.
Removal of the surveillance accuracy requirement in Section
4.6.1.2.c will not affect the probability of an accident previously
analyzed since a similar requirement is contained in ANSI/ANS-56.8-
1994, ``Containment System Leakage Testing Requirements.'' ANSI/ANS-
56.8-1994 will be used to develop the technical methods and
techniques for the Containment Leakage Rate Test Program as stated
in Regulatory Guide 1.163. The technical methods and techniques in
ANSI/ANS-56.8-1994 have been determined to be acceptable to the NRC
Staff.
Changes to the Administrative Section describe the containment
testing program only and cannot increase the probability or
consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed license amendment does not change the operation or
equipment of the plant. The change in the test frequency is
dependent on the establishment of a Containment Leakage Test
Program. This test program will ensure the performance history of
each penetration is satisfactory prior to the changing of any test
frequency. Since the performance history of the penetration will be
known, there is no possibility of the implementation of the program
creating a new or different kind of accident than previously
analyzed. Since there is no change to the equipment or the operation
of the plant, there is no possibility of creating a new or different
kind of accident than previously analyzed. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously analyzed.
Removal of the surveillance accuracy requirement in Section
4.6.1.2.c will not create the possibility of a new or different kind
of accident from those previously analyzed since a similar
requirement is contained in ANSI/ANS-56.8-1994, ``Containment System
Leakage Testing Requirements.'' ANSI/ANS-56.8-1994 will be used to
develop the technical methods and techniques for the Containment
Leakage Rate Test Program as stated in Regulatory Guide 1.163. The
technical methods and techniques in ANSI/ANS-56.8-1994 have been
determined to be acceptable to the NRC staff.
Changes to the Administrative Section describe the containment
testing program only and cannot create a different accident from any
previously analyzed.
3. Involve a significant reduction in a margin of safety.
During the development of 10 CFR Part 50, Appendix J, Option B,
the NRC staff determined the reduction in safety associated with the
implementation of the performance-based testing program. The results
of this review are documented in NUREG-1493. The review concluded
that reducing the frequency of Type A tests (Integrated Leakage Rate
Tests) from three per 10 years to one per 10 years leads to an
imperceptible increase in risk. For Type B and C testing (Local
Leakage Rate Tests), the increase in testing frequency should not
have significant impact since this leakage contributes less than 0.1
percent of the overall risk based on the existing regulations. The
use of Option B will allow the extension of testing intervals with a
minimal impact on the radiological release rates since most
penetration leakage is continually well below the specified limits.
In the accident risk evaluation, the NRC Staff noted that the
accident risk is relatively insensitive to the containment leakage
rate because the accident risk is dominated by accident sequences
that result in failure of or bypass of the containment. The use of a
performance based testing program will continue to provide assurance
that the accident analysis assumptions remain bounding. Therefore,
this change does not involve a significant reduction in the margin
of safety.
Removal of the surveillance accuracy requirement in Section
4.6.1.2.c will not involve a significant reduction in the margin of
safety since a similar requirement is contained in ANSI/ANS-56.8-
1994, ``Containment System Leakage Testing Requirements.'' ANSI/ANS-
56.8-1994 will be used to develop the technical methods and
techniques for the Containment Leakage Rate Test Program as stated
in Regulatory Guide 1.163. The technical methods and techniques in
ANSI/ANS-56.8-1994 have been determined to be acceptable to the NRC
Staff.
Changes to the Administrative Section describe the containment
testing program only and do not reduce the margin of safety.
Moreover, the Commission has provided guidance concerning the
application of standards in 10 CFR 50.92 by providing certain
examples (51 FR 7751, March 6, 1986) of amendments that are
considered not likely to involve an SHC [significant hazards
consideration]. Although the proposed change is not enveloped by a
specific example, it has been shown that the proposed change is not
an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: February 6, 1996
Description of amendment request: The proposed amendment would
delete the requirement to perform additional operability testing of
safety system train components when a required component in the
redundant train becomes inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes remove the requirement for testing which is
in addition to the normal surveillance interval. The affected
equipment is subject to periodic surveillance testing required by
the Technical Specifications. Removing the requirement for
additional testing cannot alter any plant operating conditions,
operating practices, equipment settings, or equipment capabilities.
Therefore, changing an AOT [allowable outage time] or a surveillance
interval cannot increase the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
The proposed changes remove the requirement for testing which is
in addition to the normal surveillance interval. The affected
equipment is subject to periodic surveillance testing required by
the Technical Specifications. Removing the requirement for
additional testing cannot alter any plant operating conditions,
operating practices, equipment settings, or equipment capabilities.
Therefore, changing an AOT or a surveillance interval cannot create
the possibility of a new or different
[[Page 28612]]
kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed changes remove the requirement for testing which is
in addition to the normal surveillance interval, in effect extending
the surveillance interval. An excessive surveillance interval
extension could reduce the margin of safety by reducing assurance
that required equipment will function as designed; an overly
restrictive surveillance interval could also reduce the margin of
safety by imposing unnecessary testing wear, equipment
manipulations, and system transients on the plant.
The existing requirements to perform cross-train testing were
based on the operating experience available when they were added to
the TS. Typically this was done during the initial plant licensing
in 1971. The recently published Standard Technical Specifications
(NUREG 1432) do not include cross-train testing requirements for the
Engineered Safety Features components. It has been judged by the NRC
and by the industry, that cross-train testing is unnecessary, and
that testing at normal surveillance intervals is adequate to assure
equipment operability. This recent judgment is based on a much
larger accumulation of operating experience than was available at
the time Palisades was licensed. There are no special features of
the Palisades plant which would invalidate these more recent
judgments of optimal testing requirements. Therefore, operation of
the facility in accordance with the proposed changes will not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: Mark Reinhart
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North 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
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: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas
Date of amendment request: April 29, 1996
Description of amendment request: The proposed amendment relocates
several cycle specific operating parameters from the technical
specifications to the Core Operating Limits Report per Generic Letter
88-16. The parameters being relocated by this change include the
variable low reactor coolant system pressure trip (VLPT) and the
variable low reactor coolant system pressure-temperature protective
limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1. Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The removal of the cycle-dependent variable low RCS pressure-
temperature protective limits and the VLPT setpoint from technical
speciications and placing them into the COLR has no impact on plant
safety and is considered to be administrative in nature. The
proposed change does not affect the safety analyses, physical
design, or operation of the plant. Technical specifications will
continue to require operation within the core protective and
operational limits for each reload cycle as calculated by the
approved reload design methodologies. The appropriate actions
required if limits are violated will remain in the technical
specifications. The reload report presents the results of cycle-
specific evaluations of accident analyses and transients addressed
in the ANO-1 Safety Analysis Report. The cycle-specific 10CFR50.59
evaluation of the reload
[[Page 28613]]
report demonstrates that changes in fuel cycle design and the
corresponding COLR do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2. Does not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change to relocate the variable low RCS pressure-
temperature protective limits and the VLPT setpoint from the
technical specifications to the COLR is administrative in nature. No
change to the design configuration or method of operation of the
plant is made by this proposed change, and therefore, no new
transient initiator has been created. Technical specifications will
continue to require operation within the required core protective
and operating limits and appropriate actions will be taken if the
limits are exceeded. Because plant operation will continue to be
limited by the cycle-specific COLR limits that are established using
NRC-approved methodologies, these relocations will have no impact on
plant safety.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3. Does Not Involve a Significant Reduction in the
Margin of Safety.
Existing technical specification operability and surveillance
requirements are not reduced by the proposed change to relocate the
variable low RCS pressure-temperature protective limits and the VLPT
setpoint to the COLR. The proposed changes are administrative in
nature and do not relate to or modify the safety margins defined in
and maintained by the technical specifications. The cycle-specific
COLR limits for future reload fuel cycles will continue to be
developed based on NRC approved methodologies. Each future reload
undergoes a 10CFR50.59 evaluation to assure that operation of the
plant within the cycle-specific limits will not involve a
significant reduction in a margin of safety.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: May 6, 1996
Description of amendment request: The amendment would reflect that
the name of Mississippi Power & Light Company (MP&L) has been changed
to Entergy Mississippi, Inc. The amendment revises Operating License
NPF-29 and Antitrust Conditions for the Grand Gulf Nuclear Station,
Unit 1 (GGNS) to (1) add the phrase ``(now renamed Entergy Mississippi,
Inc.)'' after the name of Mississippi Power & Light Company (MP&L), (2)
replace the name of Mississippi Power & Light Company (MP&L) by the
name Entergy Mississippi, Inc., and (3) replace a footnote by the
statement: ``Amendment ---- resulted in a name change for Mississippi
Power & Light Company (MP&L) to Entergy Mississippi, Inc.''.The
proposed amendment involves only a change in company name. It does not
involve any changes to the Technical Specifications for GGNS, or to any
requirements or limiting conditions for operation on any equipment or
any systems in the plant.
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 current Grand
Gulf Nuclear Station Facility Operating License and Antitrust
Conditions. The specific proposed change is to reflect that the name
of one of the companies owning Grand Gulf Nuclear Station has
legally changed from Mississippi Power & Light Company to Entergy
Mississippi, Inc.
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 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 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 unaffected.
[The proposed change does not affect any of the existing
requirements or commitments on equipment or systems that are
designed for the safe operation of the plant. It does not affect the
design or operation of the plant.]
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 to the Operating License [and
Antitrust Condition] requirements [to change the name of Mississippi
Power & Light] do not involve any change in the design or operation
of the plant. The company will still own all of the same assets,
serve the same customers, and all existing obligations and
commitments will continue unaffected.
[The proposed changes do not affect equipment or systems that
could caused an accident at 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 [in name] is administrative in nature, as
described above; therefore, this change does not reduce the level of
safety imposed by any current requirements. [The proposed changes do
not affect any equipment or systems at the plant.] The company will
still own all of the same assets, serve the same customers, and all
existing obligations and commitments will continue unaffected.
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. herefore, 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
[[Page 28614]]
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: May 8, 1996
Description of amendment request: The amendment request would
replace the current frequency requirements in Surveillance Requirement
(SR) 3.6.1.3.5, on the leakage rate testing for each containment purge
valve with resilient seals, in the Technical Specifications for Grand
Gulf Nuclear Station, Unit 1 (GGNS). The proposed changes would place
these purge valves on a performance-based leakage testing frequency,
instead of the current once every 184 days and once within 92 days
after opening the valve.The proposed changes do not change the limiting
conditions for operation, the required actions for inoperability, or
the other surveillance requirements on these primary containment
isolation valves.
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:
In accordance with 10 CFR 50.92, Entergy Operations, Inc. has
evaluated the proposed change to the Operating License of GGNS and
has determined that the operation of the facility in accordance with
the proposed amendment would not involve any significant hazards
considerations. In accordance with 10 CFR 50.91(a), Entergy
Operations, Inc. is providing the following analysis of the proposed
amendment against the three [following] standards of 10 CFR
50.92(c):
1) The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
This change deletes the augmented testing requirement for these
containment isolation valves and allows the surveillance intervals
to be set in accordance with the Appendix J testing program.
[Appendix J to 10 CFR Part 50 defines primary containment leakage
testing requirements for water-cooled power reactors as GGNS and
these requirements include frequency of testing for the primary
containment isolation valves.] This change does not affect the
system function or design. The purge valves are not an initiator of
any previously analyzed accident. Leakage rates do not affect the
probability of the occurrence of any accident. Operating history has
demonstrated that these valves do not degrade and cause leakage as
previously anticipated. Because these valves have been demonstrated
to be reliable, these valves can be expected to perform the
containment isolation function as assumed in the accident analyses.
Therefore, there is no significant increase in the consequences
of any previously evaluated accident.
2) The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Extending the test intervals has no influence on, nor does it
contribute in any way to, the possibility of a new or different kind
of accident or malfunction from those previously analyzed. No change
has been made to the design, function or method of performing
leakage testing [or to the design and function of these valves].
Leakage acceptance criteria have not changed. No new accident modes
are created by extending the testing intervals. No safety-related
equipment or safety functions are altered as a result of this
change.
[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 only margin of safety that has the potential of being
impacted by the proposed changes involves the offsite dose
consequences of postulated accidents which are directly related to
the containment leakage rate. The proposed change does not alter the
method of performing the tests nor does it change the leakage
acceptance criteria. Sufficient data has been collected to
demonstrate that the resilient seals do not degrade at an
accelerated rate.
[Also, the proposed change would test these valves in accordance
with the Appendix J testing program at the plant. Appendix J to 10
CFR Part 50 defines primary containment leakage testing requirements
for water-cooled power reactors as GGNS and these requirements
include frequency of testing for the primary containment isolation
valves.]
Because of this demonstrated reliability, this change will
provide sufficient surveillance to determine an increase in the
unfiltered leakage prior to the leakage exceeding that assumed in
the accident analysis.
Therefore, the proposed change does not result in a significant
reduction in a margin of safety.
Based on the above evaluation, Entergy Operation, 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 9, 1996
Description of amendment request: The amendment request would (1)
increase the safety limit minimum critical power ratio (MCPR) for two
loop operation and single loop operation to 1.10 and 1.11,
respectively, and (2) add a General Electric topical report to the list
of documents describing the analytical methods used to determine the
core operating limits. The proposed changes are to Section 2.1.1,
Reactor Core Safety Limits, and Section 5.6.5, Core Operating Limits
Report (COLR), respectively, of the Technical Specifications (TSs).
The licensee also proposed changes to the Bases of the TSs
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:
Entergy Operations, Inc. proposes to change the current Grand
Gulf Nuclear Station [GGNS] Technical Specifications. The specific
change is to modify the Minimum Critical Power Ratio (MCPR) safety
limits reported in Technical Specification 2.1.1.2, the list of
references in Technical Specification 5.6.5, and associated Bases
changes. The proposed change is necessary in order to switch reload
fuel vendors. [General Electric GE11 fuel is being added to the core
in place of Siemens Power Corporation (SPC) fuel.]
The Commission has provided standards for determining whether no
significant hazards considerations exists as stated in 10 CFR 50.92
(c). A proposed amendment to an operating license involves no
significant hazards 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; (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 no significant
hazards consideration in its request for this license amendment and
determined that no significant hazards considerations result 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.
[[Page 28615]]
The Minimum Critical Power Ratio (MCPR) safety limit is defined
in the Bases to Technical Specification 2.1.1 as that limit which
``ensures that during normal operation and during Anticipated
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in
the core do not experience transition boiling.'' The MCPR safety
limit is re-evaluated for each reload and, for GGNS [Operating]
Cycle 9, the analyses have concluded that a two-loop MCPR safety
limit of 1.10 based on the application of the generic GE MCPR
methodology is necessary to ensure that this acceptance criterion is
satisfied. For single-loop operation, a MCPR safety limit of 1.11
based on the generic GE MCPR methodology was determined to be
necessary. Core MCPR operating limits are developed to support the
Technical Specification 3.2 requirements and ensure these safety
limits are maintained in the event of the worst-case transient.
Since the MCPR safety limit will be maintained at all times,
operation under the proposed changes will ensure at least 99.9% of
the fuel rods in the core do not experience transition boiling.
Therefore, The Minimum Critical Power Ratio (MCPR) safety limit
change does not affect the probability or consequences of an
accident.
The implementation of GE's GESTAR-II approved methodology has no
effect on the probability or consequences of any accidents
previously evaluated. One exception to GESTAR is that the mis-
oriented and mis-located bundle events will continue to be analyzed
as accidents subject to the acceptance criteria in the current
licensing basis. The design of the GE11 fuel bundles is such that
the bundles are not likely to be mis-oriented or mis-located and the
normal administrative controls will be in effect for assuring proper
orientation and location. Therefore, the probability of a fuel
loading error is not increased. This analysis ensures that
postulated dose releases will not exceed a small fraction (10
percent) of 10CFR100 limits.
Therefore, the consequences of accidents previously evaluated
are unchanged.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The GE11 fuel to be used in [Operating] Cycle 9 is of a design
compatible with fuel present in the core and used in the previous
cycle. Therefore, the GE11 fuel will not create the possibility of a
new or different kind of accident. The proposed changes do not
involve any new modes of operation, any changes to setpoints, or any
plant modifications. They introduce revised MCPR safety limits that
have been proved to be acceptable for Cycle 9 operation. Compliance
with the applicable criterion for incipient boiling transition
continues to be ensured. The proposed MCPR safety limits do not
result in the creation of any new precursors to an accident.
Therefore, the proposed changes do not create the possibility of
a new or different type of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
The MCPR safety limits have been evaluated to ensure that during
normal operation and during AOOs [abnormal operating occurrences],
at least 99.9% of the fuel rods in the core do not experience
transition boiling. Therefore, the implementation of the proposed
changes in the MCPR safety limit ensure there is no reduction in the
margin of safety.
As with the current SPC methodology, GGNS will implement only
the NRC-approved revisions to GE's GESTAR methodology. This GE
methodology is similar to those SPC reports currently listed in TS
5.6.5 and it will be applied in a similar, conservative fashion. One
exception to GESTAR is that the mis-oriented and mis-located bundle
events will continue to be analyzed as accidents subject to the
acceptance criteria in the current licensing basis. This analysis
ensures that postulated dose releases will not exceed a small
fraction (10 percent) of 10CFR100 limits. On this basis, the
implementation of this GE methodology 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: 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., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 6, 1996
Description of amendment request: The proposed change will amend
the Allowable Values of parameters in Table 3.3-4 of Waterford Steam
Electric Station, Unit 3, (Waterford 3) Technical Specifications (TSs)
to make it consistent with the identical parameters in Table 2.2-1 of
TSs for Waterford 3. The proposed change will add Mode 4 to the
surveillance requirements of Table 4.3-2, Item 5.c (Safety Injection
System Automatic Actuation Logic) that was inadvertently removed.
Finally, the proposed change removes a reference to TS 3.3.3.2 in
Surveillance Requirements TS 4.10.2.2 and 4.10.4.2 since Incore
Detectors has been removed from the TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes described herein are administrative changes
necessary to correct administrative errors. The proposed changes
will have no affect on any design basis accidents nor will these
changes affect any material condition of the plant. Therefore, the
proposed changes will not involve a significant increase in the
probability or consequences of any accident previously evaluated.
The proposed changes are purely administrative. There are no new
system or design changes associated with this proposal. Therefore,
the proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change will have no impact on any protective
boundary, safety limit, or margin to safety. The proposed change
corrects inconsistencies in the TS and is purely administrative in
nature. Therefore, the proposed change will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: May 7, 1996 (TSCR 247)
Description of amendment request: The proposed change to the
technical specifications would adopt the provisions of the Standard
Technical Specifications (STS), NUREG-1433, Rev. 1, which clarify
surveillance requirement applicability and allow a maximum period of 24
hours to complete a surveillance requirement upon discovery that the
surveillance has been missed.
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
[[Page 28616]]
involve a significant increase in the probability of occurrence or
consequence of an accident previously evaluated. The proposed
changes only affect administrative requirements regarding the
applicability and performance of surveillances. This change
clarifies surveillance requirement applicability and allows a
maximum 24 hour delay period for the performance of a surveillance
when it is discovered that the surveillance has not been performed
within the required frequency, consistent with the STS. There is
minimal safety significance associated with a delay of 24 hours in
completing the required surveillance, particularly due to the fact
that the most probable result of any particular surveillance
performed is the successful verification of conformance with the
requirements.
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 previously evaluated. The proposed changes
only affect administrative requirements regarding the applicability
of surveillance requirements and the performance of surveillances to
allow a maximum 24 hour delay period when it is discovered that a
surveillance has been missed. No changes to plant equipment or
operation are affected.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety since the change contained in the proposed amendment does not
change any existing safety margins.
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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island
Nuclear Station, Unit No. 2 (TMI-2), Dauphin County, Pennsylvania
Date of amendment request: February 16, 1995
Description of amendment request: The proposed amendment would
revise TMI-2 Operating License No. DPR-73 by modifying sections 4.02,
4.04, and 4.1.1.3 of the unit technical specifications. The revisions
to sections 4.02 and 4.04 would add flexibility to the scheduling of
surveillance activities and would allow for a 24 hour period to perform
missed surveillances before declaration of a limiting condition for
operation, respectively. These changes would make the TMI-2 technical
specifications consistent with the Standard Technical Specifications
for B&W Plants (NUREG-1430). The revision to section 4.1.1.3 would
allow extension of the time interval for surveillance of the
containment airlock doors from quarterly to annually. The proposed
changes to the TMI-2 technical specifications section 4.1.1.3 would
allow a decrease in worker exposure to radiation while maintaining an
adequate level of environmental protection at the facility.
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:
10 CFR 50.92 provides the criteria which the Commission uses to
perform a no significant hazards consideration. 10 CFR 50.92 states
that an amendment to a facility license involves no significant
hazards 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.
The proposed changes to the technical specifications sections
4.02 and 4.04 are administrative and do not involve any physical
changes to the facility. No changes are made to operating limits or
parameters, nor to any surveillance activities. The changes to
section 4.1.1.3 extends the interval between surveillance of the
containment airlocks; it does not change the operability
requirements, test methodology or acceptance criteria. Based on
this, GPU Nuclear has concluded that the proposed changes to
sections 4.02 and 4.04 do not:
1. Involve a significant increase in the probability of
occurrence or the consequences of an accident previously evaluated.
The changes do not modify any operating parameters or the release of
radioactive materials. The clarification of maximum time extensions
for surveillance is consistent with the NRC's Standard Technical
Specifications for Babcock and Wilcox Plants (NUREG-1430).
2. Create the possibility of a new or different kind of accident
since these change are administrative and no plant configuration or
operational changes are involved.
3. Involve a change in the margin of safety. These changes are
administrative in nature, compatible with standard technical
specifications, and do not affect any safety settings or operational
limits.
GPU Nuclear has also concluded that the proposed changes to
section 4.1.1.3 do not:
1. Involve a significant increase in the probability of
occurrence of or consequences of an accident previously evaluated.
The change to this section does not change operating parameters,
equipment operability requirements, surveillance test methodology,
or acceptance criteria.
2. Create the possibility of a new or different kind of accident
since the change does not affect plant equipment, plant
configuration, or plant operating parameters.
3. Involve a change in the margin of safety since the change
does not affect any operational limits.
Based on the above analysis the licensee concluded that the
proposed changes involve no significant safety hazards
considerations as defined by 10 CFR 50.92.
The NRC staff has reviewed the analysis of the licensee 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: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: Seymour H. Weiss
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 1, 1996
Description of amendment request: The proposed amendments would
change the Technical Specifications to implement 10 CFR Part 50,
Appendix J, Option B, by referring to Regulatory Guide 1.163,
``Performance-Based Containment Leakage-Test Program.''
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:
[South Texas Project] STP has evaluated the proposed Technical
Specification Amendment and determined that it does not represent a
significant hazards consideration. Based on the criteria for
defining a significant hazards consideration established in 10 CFR
50.92, operation of STP in accordance with the proposed amendment
will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
10 CFR [Part] 50, Appendix J has been amended to include
provisions regarding
[[Page 28617]]
performance based leakage testing requirements (Option B). Option B
allows plants with satisfactory Integrated Leak Rate Testing (ILRT)
performance history to extend the Type A testing interval from three
tests in ten years to one test in ten years. For Type B and Type C
tests, Option B allows extended testing interval[s] based on the
leak rate test history of each component. To be consistent with the
requirements of 10 CFR [Part] 50, Appendix J, Option B, STP proposes
to include appropriate changes to the Technical Specifications that
incorporate the necessary revisions associated with 10 CFR [Part]
50, Appendix J, Option B.
The proposed amendment represents the conversion of current
Technical Specification requirements to maintain consistency with
those requirements specified by 10 CFR [Part] 50, Appendix J, Option
B. The proposed changes are consistent with the current safety
analyses. Implementation of these changes will provide continued
assurance that specified parameters associated with containment
integrity will remain within acceptance limits, and will not
significantly increase the probability or consequences of a
previously evaluated accident.
Some proposed changes represent minor relaxations in current
Technical Specification requirements, but are based on the
requirements specified by Option B of 10 CFR [Part] 50, Appendix J.
Changes are consistent with the current safety analyses and
determined to represent sufficient requirements for the assurance
and reliability of equipment assumed to operate in the safety
analyses, and provide continued assurance that specified parameters
associated with containment integrity remain within their acceptance
limits. These changes will not significantly increase the
probability or consequences of a previously evaluated accident.
The systems affecting containment integrity related to this
proposed amendment request are not assumed in any safety analyses to
initiate any accident sequence. The probability of any accident
previously evaluated is not increased by this proposed amendment.
The proposed changes to Technical Specification LCOs or SRs maintain
an equivalent level of reliability and availability for all affected
systems. The proposed amendment does not increase the consequences
of any accident previously evaluated.
There is no change to the consequences of an accident previously
evaluated because maintaining leakage within the analyzed limit
assumed for any associated accident analyses does not adversely
affect either the on-site or off-site dose consequences resulting
from an accident. There is no adverse impact on the probability of
accident initiators. There is no significant increase in the
probability of any previously analyzed accident. A plant specific
risk-based analysis of Appendix J performed for STP indicates the
containment penetration leakage dose rate contribution to the total
dose rate in person-rem is insignificant.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
10 CFR [Part] 50, Appendix J, Option B specifies, in part, that
a Type A test which measures both the containment system overall
integrated leakage rate at containment pressure and system
alignments assumed during a large break LOCA [loss-of-coolant
accident], and demonstrates the capability of primary containment to
withstand an internal pressure load, may be conducted at an interval
based on the performance of the overall containment system. The
acceptable leakage rates are specified in the plant's Technical
Specifications. For Type B and Type C tests, intervals are proposed
based on the performance history of each component. Acceptance
criteria for each component is based upon demonstration that the sum
leakage rates at design basis pressure conditions for applicable
penetrations, is within the limit specified in the Technical
Specifications.
The proposed amendment represents the conversion of current
Technical Specification requirements to maintain consistency with
those requirements specified in 10 CFR [Part] 50, Appendix J, Option
B. The proposed changes are consistent with the current safety
analyses. Some minor relaxations in current Technical Specification
requirements, associated with containment integrity are based on
generic guidance provided in Option B, NEI 94-01 and ANSI/ANS 56.8,
1994. These changes do not involve revisions to the design of the
station. Some of the changes may involve revision in the testing of
components; however, these are in accordance with the STP current
safety analyses and provide for appropriate testing or surveillance
that are consistent with 10 CFR [Part] 50, Appendix J, Option B. The
proposed changes will not introduce new failure mechanisms beyond
those already considered in the current safety analyses.
The proposed amendment has been reviewed for acceptability
considering similarity of system or component design affecting
containment integrity. No new modes of operation are introduced by
the proposed changes. Surveillance requirements are changed to
reflect corresponding changes associated with Option B of 10 CFR
[Part] 50, Appendix J and improvements in technique or interval of
leak rate testing performance. The proposed changes maintain, at
minimum, the present level of operability of any system that affects
containment integrity. The proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
The associated systems that affect leak rate integrity related
to the proposed amendment, are not assumed in any safety analysis to
initiate any accident sequence. The proposed surveillance
requirements for any affected systems are consistent with the
current requirements specified within the Technical Specifications
and are consistent with the requirements of Option B of 10 CFR
[Part] 50, Appendix J. The proposed surveillance requirements
maintain an equivalent level of reliability and availability of all
affected systems and therefore, does not increase the consequences
of any previously evaluated accident.
3) Involve a significant reduction in the margin of safety
because:
The provisions specified in Option B of 10 CFR [Part] 50
Appendix J allow changes to Type A, Type B, and Type C test
intervals based upon the performance of past leak rate tests. The
effect of extending containment leakage rate testing intervals has a
corresponding increase in the likelihood of containment leakage. The
degree to which intervals can be extended is a direct function of
the potential effect to existing safety margins and the public
health and safety that can occur due to an increased likelihood of
containment leakage. 10 CFR [Part] 50 Appendix J, Option B allows
longer intervals between leakage tests based on performance trends
but does not increase the leakage acceptance criteria. La [maximum
allowable leakage rate] is still limited to 0.3 wt%/day. By
referencing the Containment Leakage Rate Testing Program in LCO
3.6.1.2 ACTION, the point at which ACTION is required is increased
from .75 La to 1.0 La. This makes the specification consistent with
the intent of having margin between an AS-LEFT leakage of less than
or equal to .75 La and maintaining operability with less than or
equal to 1.0 La.
Changing Appendix J test intervals from those currently provided
in the Technical Specification to those provided in 10 CFR [Part]
50, Appendix J, Option B, slightly increases the risk associated
with Type A, Type B, and Type C specified accident sequences.
Historical data suggests that increasing the Type C test interval
can slightly increase the associated risk; however, this is
compensated by the corresponding risk reduction benefits associated
with reduction in component cycling, stress, and wear associated
with increased test intervals. When considering the total integrated
risk which includes all analyzed accident sequences, the risk
associated with increasing test intervals is negligible. A plant
specific risk-based analysis of Appendix J performed for STP
indicates the containment penetration leakage dose rate contribution
to total dose rate in person-rem is insignificant.
STP proposes to revise the Technical Specifications to be
consistent with those provisions specified in Option B of 10 CFR,
Appendix J. The proposed changes are consistent with the STP current
safety analyses. These proposed changes do not involve revisions to
the design of the station. The proposed changes will maintain the
same level of reliability of equipment associated with containment
integrity assumed to operate in the safety analysis, and provide
continued assurance that specified parameters affecting plant leak
rate integrity will remain within acceptance limits. The proposed
changes provide continued assurance of leakage integrity of
containment without adversely affecting the public health and safety
and will not significantly reduce existing safety margins. Plant
specific risk-based analysis indicates sufficient technical
justification exists to further extend the limits beyond those
allowed by Option B.
The proposed amendment to the Technical Specifications
implements present requirements, or the requirements in accordance
with the guidelines set forth in Option B of 10 CFR [Part] 50,
Appendix J. NUREG-1493, ``Performance-Based Containment Leak-Test
Program,'' served as the technical basis for Option B. STP
[[Page 28618]]
performed a plant specific risk-based analysis of containment
penetration leakage dose utilizing the same methodology used in
NUREG-1493. The analysis indicates the containment penetration
leakage dose rate contribution to the total dose rate in person-rem
is insignificant. This plant specific analysis serves to validate
the applicability of the proposed changes for STP. The proposed
changes have been approved by the NRC, are applicable to STP,
maintain necessary levels of system or component reliability
affecting containment integrity, and do not involve a significant
reduction in the margin of safety.
The performance-based approach to leakage rate testing concludes
the impact on public health and safety due to revised testing
intervals is negligible. The proposed amendment will not reduce
availability of systems associated with containment integrity when
required to mitigate accident conditions; therefore, the proposed
changes do not involve a significant reduction in the margin of
safety.
Guidance has been provided in ``Final Procedures and Standards
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,
for the application of standards to license change requests for
determination of the existence of significant hazards
considerations. This document provides examples of amendments which
are and are not considered likely to involve significant hazards
considerations.
This proposed amendment does not involve a significant
relaxation of the criteria used to establish safety limits, a
significant relaxation of the bases for limiting safety system
settings or a significant relaxation of the bases for LCOs.
Therefore, based on the guidance provided in the Federal Register
and criteria established in 10 CFR 50.92(c), the proposed change
does not constitute a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
NRC Project Director: William D. Beckner
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 amendment requests: December 14, 1995
Description of amendment requests: The proposed amendments would
revise the Administrative Control (Chapter 6) Section and other
affected Sections of the Prairie Island Technical Specifications to
generally conform with NUREG-1431, Standard Technical Specifications,
Westinghouse Plants, Revision 1, dated April 7, 1995.
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 will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Operation of the Prairie Island plant in accordance with the
proposed changes does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
None of the proposed changes involve a physical modification to the
plant, a new mode of operation or a change to the Updated Safety
Analysis Report transient analyses. These proposed amendments
generally conform to the guidance of NUREG-1431, Revision 1, Section
5.0 which was previously reviewed, accepted and issued by the NRC.
Some Section 5.0 Specifications in NUREG-1431 were not
incorporated in this License Amendment Request. These Specifications
were not proposed because they 1) specify requirements not currently
in the Prairie Island Technical Specifications or otherwise
committed to, 2) are addressed elsewhere in the current Technical
Specifications, or 3) the current Technical Specifications level of
commitment is maintained. In all these instances, the NRC has
previously reviewed and approved the proposed level of commitment
through the issuance of the current Prairie Island Technical
specifications.
The proposed changes, in themselves, do not reduce the level of
qualification or training such that personnel requirements would be
decreased.
In total these changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed changes, in themselves, do not introduce a new
mode of plant operation, surveillance requirement or involve a
physical modification to the plant. These proposed amendments
generally conform to the guidance of NUREG-1431, Revision 1, Section
5.0 which was previously reviewed, accepted and issued by the NRC.
Some Section 5.0 Specifications in NUREG-1431 were not
incorporated in this License Amendment Request. These Specifications
were not proposed because they 1) specify requirements not currently
in the Prairie Island Technical Specifications or otherwise
committed to, or 2) are addressed elsewhere in the current Technical
Specifications. Other features are not fully implemented but rather,
the current Technical Specification level of commitment is
maintained. In all these instances, the NRC has previously reviewed
and approved the proposed level of commitment through the issuance
of the current Prairie Island Technical Specifications.
In general, the proposed changes are administrative in nature.
The changes propose to revise, delete or relocate Specifications
within the Technical Specifications or from the Technical
Specifications to the Updated Safety Analysis Report, plant
procedures or the Operational Quality Assurance Plan through which
adequate control is maintained. The proposed changes do not alter
the design, function, or operation of any plant components and
therefore, no new accident scenarios are created.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be created
[by] these amendments.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed changes do not involve a significant reduction in a
margin of safety because the Current Technical Specifications
requirements for safe operation of the Prairie Island plant are
maintained or increased. The proposed changes are administrative in
nature and do not involve a physical modification to the plant, a
new mode of operation or a change to the Updated Safety Analysis
Report transient analyses. The proposed changes do not alter the
scope of equipment currently required to be operable or subject to
surveillance testing nor does the proposed change affect any
instrument setpoints or equipment safety functions.
Therefore, a significant reduction in the margin of safety would
not be involved with these amendments.
Based on the evaluation describe above, and pursuant to 10 CFR
Part 50, Section 50.91, Northern States Power Company has determined
that operation [of] the Prairie Nuclear Generating Plant in
accordance with the proposed license amendment request does not
involve any significant hazards considerations as defined by Nuclear
Regulatory Commission regulations in 10 CFR Part 50, Section 50.92.
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 requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
[[Page 28619]]
NRC Project Director: Mark Reinhart (Acting Director)
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 15, 1996
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Nuclear Power Plant, Unit Nos. 1 and 2 to revise Technical
Specification 3.5.2, ``ECCS Subsystems - Tavg Greater Than or Equal to
350 deg.F,'' to change the allowed outage time for any one safety
injection pump from 72 hours to 7 days. The specific TS change proposes
to add a new footnote that increases the allowed outage time (AOT) for
one safety injection (SI) pump from 72 hours to 7 days for performance
of non-routine, emergent maintenance and requires review by the Plant
Staff Review Committee (PSRC), and requires Plant Manager approval
prior to exceeding 72 hours.
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 allowed outage time (AOT) extension does not change
the operating practices of Diablo Canyon Power Plant (DCPP).
Although the proposed change increases the allowed time in which the
safety injection (SI) system may be out of service for maintenance
or testing, this extended AOT will only be used in emergent
circumstances.
Increasing the AOT for the SI pumps does not involve physical
alteration of any plant equipment and does not affect analysis
assumptions regarding functioning of required equipment designed to
mitigate the consequences of accidents. Further, the severity of
postulated accidents and resulting radiological effluent releases
will not be affected by the increased AOT.
Finally, the probabilistic risk assessment determined that the
increase in the core damage probability is not considered
significant.
Therefore, the proposed change does 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 proposed increase to the SI pump AOTs does not change the
method by which DCPP operates. Further, the proposed change would
not result in any physical alteration to any plant system, and there
would not be a change in the method by which any safety related
system performs its function.
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.
There is no safety analysis impact since the extension of the SI
pump AOT interval will have no effect on any safety limit,
protection system setpoint, or limiting condition of operation.
There is no hardware change that would impact existing safety
analysis acceptance criteria.
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
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
]Date of application request: April 17, 1996
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 3/4.3 to support a future
modification to replace existing digital portions of the main steam and
feedwater isolation system (MSFIS) with digital processor equipment and
would authorize revision of the FSAR to include a description of the
MSFIS modification. The MSFIS modification is a change to the facility,
as described in the safety analysis report, that involves an unreviewed
safety question. The modification involves an unreviewed safety
questions because: (1) the MSFIS design will use software which could
result in a common mode failure, (2) the original NRC review of the
MSFIS did not evaluate 2 out of 3 coincidence circuitry, which could
introduce new system failure modes, and (3) the MSFIS modification
utilizes manual handswitches that could introduce new system failure
modes. The NRC will review the modification in accordance with 10 CFR
50.59(a)(2) in conjunction with the review of the proposed TS
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, 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 addition of the MSFIS actuation logic and relays to the TS
has no adverse impact on the probability of occurrences or the
consequences of an accident. The proposed amendment does not change
or alter the design assumptions for the systems or components used
to mitigate the consequences of an accident and the methodologies
used in the accident analysis remain unchanged. The operating limits
will not be changed.
No design basis accidents will be affected by this design change
since the logic which currently exists will continue to be
performed. Thus, the radiological consequences will not change.
The system response time is enveloped by the current 5 second
valve stroke time. The MSFIS response time will be less than 500
msec.
A common mode software failure could exist if both separation
groups have their PLCs [programmable logic controllers] (3 per train
- six total) malfunction at the same time. However, a diverse means
of isolating the feedwater lines exists given the ability of the
Main Feed Control Valves to close on a Feedwater Isolation Signal.
The MSIVs [main steam isolation valves] do not have a diverse means
of isolating their respective steam lines if a common mode software
failure occurs. As a result, this modification provides a means to
manually fast close the valves at the MSFIS cabinets. The operators
will be alerted of the failure conditions of any PLC logic channel
via MCB [main control board] annunciators and indicators. This
failure mode has a low probability of occurrence based upon the
inherent quality of the design provided by the V&V [verification &
validation] process. Therefore, the accident consequences are not
increased for this failure mode.
The test panel in the MSFIS cabinets has been laid out to
provide the same functions as the existing test panel, except that
PLC status indication and coincidence logic test functions are
provided. The Emergency Override Panel, located below the Test
Panel, provides the operator with the ability to bypass the FWIS
[feedwater isolation signal] signal and manually fast close each
MSIV as required by the Emergency Operating Procedures. The MSIV
manual FC [fast close] switch operation is necessary for a diverse
means of operation for software common mode failures. The FWIS
bypass switch will allow main feedwater flow to be re-established to
each Steam Generator.
[[Page 28620]]
The replacement system is functionally the same as the current
system since it performs the same logic, receives the same inputs,
and produces the same outputs. However, the system is more reliable
and possesses triple redundant logic. Therefore, the probability of
malfunction will not be increased.
The electrical load of the A-B PLC equipment and existing 48 VDC
[volts direct current] actuation relays is less than that of the
existing equipment so the system will not require any additional
cooling over the existing equipment. Proper grounding is provided
for the PLC 5 VDC and actuation relay 48 VDC power supplies, which
are electrically isolated from each other.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The addition of the MSFIS actuation logic and relays to the TS
will not create a new type of accident or malfunction than any
previously evaluated in the Safety Analysis Report. The safety
functions of the system are not changed in any manner, nor is the
reliability of any structure, system or component reduced. All
design and performance criteria continue to be met. Since the safety
functions and reliability are not adversely affected, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The operator's ability to adequately respond to an accident is
not hindered by the man-machine interface added as a result of this
modification since the operator interface is similar to the current
system and the MCB controls will not change. The operators will be
alerted to system malfunctions through annunciation. The current
system has a status output for each MSIV and FIV [feedwater
isolation valve] valve on the Engineered Safety Feature Status
Panel, which will be maintained. In addition, an isolated plant
annunciator interface will provide a MSFIS Channel Failure plant
annunciator window for both trains. Training will be provided to the
technicians, engineers, and operators on the new features of the
system prior to installation. Therefore, this modification does not
increase the consequential effects due to the man-machine interface.
The system is compatible with the normal and accident
environments and will be seismically qualified in accordance with
the SNUPPS [standardized nuclear unit power plant system] seismic
spectra profile. The equipment will be qualified for Electromagnetic
Interference concerns in accordance with EPRI [Electric Power
Research Institute] document TR-102323-EPRI Guideline and will meet
the EPRI EMI [electromagnetic interference] limiting practices.
The system has the same failure mode upon loss of power as the
current system and behaves similarly upon power restoration. A loss
of power will not result in a MSFIS actuation.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The addition of the MSFIS actuation logic and relays to the TS
will not affect or change a safety limit or affect plant operations.
This change will not reduce the margin of safety assumed in the
accident analysis nor reduce any margin of safety as defined in the
basis for any TS.
The system response time for any given valve will not exceed the
required valve stroke time. Since the MSFIS does not contain any
analog channels, no channel trip accuracies are impacted.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: October 25, 1995
Description of amendment request: The proposed changes would
provide an allowed outage time of 14 days for the pressurizer power-
operated relief valve (PORV) nitrogen accumulators, as well as provide
separate action statements for the PORV depending on the reason for the
PORV inoperability during plant operation in power Modes 1, 2, or 3.
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.
Specifically, operation of North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The PORVs are assumed to mitigate the consequences of a steam
generator tube rupture as described in the North Anna UFSAR [Updated
Final Safety Analysis Report] as well as to limit the undesired
opening of the pressurizer safety valves for a primary overpressure
event. The proposed action statements ensure that the steam
generator tube rupture accident analysis requirements are met. The
proposed Technical Specification changes require the backup nitrogen
supply be available for the PORVs to be consideredoperable and add
action statements and surveillance requirements for the nitrogen
supply commensurate with its significance. The proposed action
statements enhance the availability of the automatic actuation of
the PORVs by not requiring the block valves to be closed when the
backup nitrogen supplies are inoperable. The proposed surveillance
requirements enhance the reliability of the backup nitrogen supply
to the PORVs by verifying that there is sufficient nitrogen pressure
in the accumulators for the PORVs to perform their design function.
The proposed Technical Specification changes do not change any
accident analyses, therefore, the probability of any accident and
its resulting consequences are not increased.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed Technical Specification changes do not involve any
physical modification to the plant or result in a change in a method
of operation. The backup nitrogen supply continues to be required
for PORV operability. The proposed Technical Specification changes
provide operational flexibility and ensure the availability of the
PORVs using the normal supply of instrument air while the backup
nitrogen supply is being restored. This also prevents undesirable
challenges to the pressurizer safety valves. The new surveillance
requirements verify that there is sufficient nitrogen pressure in
the accumulators for the PORVs to perform their design functions.
3. Involve a significant reduction in a margin of safety.
The proposed Technical Specification changes do not affect any
safety limits or limiting safety system settings. The availability
of the PORVs will be maintained as required in Generic Letter 90-06.
The proposed Technical Specifications will continue to ensure that
the PORVs will be capable of performing their intended functions.
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: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: Eugene V. Imbro
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Date of amendment request: April 24, 1996
[[Page 28621]]
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) Section 15.7, ``Radiological
Effluent Technical Specifications (RETS).'' Portions of the RETS would
be moved to licensee-controlled documents consistent with Nuclear
Regulatory Commission guidance on TS improvements. Changes to other
sections of the TSs are also proposed consistent with the removal of
portions of the RETS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment simplifies the RETS and implements the
recommendations of GL 89-01 and of GL 95-10. The proposed change
relocates the operational requirements of RETS but keeps the
programmatic controls for these requirements in the Technical
Specifications. Therefore, the proposed changes are administrative
in nature and do not affect plant operations. Hence, the proposed
amendment does not involve a significant increase in the probability
or consequences of an accident previously evaluated because no
safety-related equipment, safety function, or plant operation will
be altered as a result of this proposed change. Also, the changes
are unrelated to the initiation and mitigation of accidents and
equipment malfunctions addressed in the Final Safety Analysis
Report.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
As stated above, the proposed action is the relocation of the
RETS procedural details to various manuals while retaining the
administrative controls in RETS. The relocation is consistent with
the intent of the guidance of GL 89-01 and of GL 95-10. It is
administrative and has no impact on plant operation or safety. No
safety-related equipment, safety function, or plant operation will
be altered as a result of this proposed change. No changes to plant
components or structures are introduced which could create new
accidents or malfunctions not previously evaluated.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated because no new accident scenario is created and no
previously evaluated accident scenario is changed by the relocation
of the procedural details of RETS from one controlled document to
another.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
The proposed change does not include a change to any plant
structure, system, component, or operation. The proposed changes do
not alter the basic regulatory requirements and do not affect any
safety analyses. The proposed change is administrative. The
procedural details of the current RETS are relocated while the
programmatic controls consistent with regulatory requirements,
including controls on revisions to the manuals receiving the RETS
procedural details, the Environmental Manual (EM), Radiological
Effluent Control Program Manual (RECM), Offsite Dose Calculation
Manual (ODCM), and Process Control Program (PCP), remain in RETS.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Date of amendment request: April 29, 1996
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) Section 15.3.14, ``Fire Protection
System,'' and Section 15.4.15, ``Fire Protection System.'' These
specifications would be relocated to other licensee-controlled
documents in accordance with Nuclear Regulatory Commission generic
guidance. Additional administrative changes consistent with the
relocation are also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of this facility under the proposed Technical
Specifications change will not increase the probability or
consequences of an accident previously evaluated.
This change request proposes to remove certain fire protection
program requirements from the Point Beach Technical Specifications
and incorporate them into the Final Safety Analysis Report (FSAR)
and the Fire Protection Evaluation Report (FPER). No requirements
are eliminated, modified, or de-emphasized by this change. The
proposed amendment ensures that any future changes to the fire
protection program will be subject to an appropriate evaluation in
accordance with NRC regulations to ensure that there are no
unreviewed safety questions.
Therefore, these proposed changes are administrative in nature.
There are no proposed changes to the physical plant or the processes
which ensure the plant's capability to mitigate fires and achieve
safe shutdown. Therefore, there is no potential effect on the
probability or consequences of previously evaluated accidents.
2. Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
New or different accidents can only be created by new or
different accident initiators or sequences. Because there are no
proposed changes to the physical plant or the processes which ensure
the plant's fire protection capability, new or different kinds of
accident initiators will not be introduced by this change. The
proposed changes are administrative in nature.
3. Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety.
The margins of safety for Point Beach are based on the design
and operation of the reactor and containment and the safety systems
that provide their protection. Because there are no proposed changes
to the physical plant or the processes which ensure the plant's fire
protection capability, there will be no effect on the reactor,
reactor containment, or the safety systems which provide their
protection. Therefore, the proposed changes will not create a
reduction in a margin of safety. The proposed changes are
administrative in nature.
Additionally, the proposed revision to Point Beach's operating
license will not allow Wisconsin Electric to make changes to the
approved fire protection program without prior approval of the
Nuclear Regulatory Commission should these proposed changes
adversely affect the ability to achieve and maintain safe shutdown
in the event of a fire. In accordance with NRC Generic Letter 86-10,
any proposed change to the approved fire protection program requires
the performance of a 10 CFR 50.59 evaluation and a fire hazards
analysis. Should these evaluations indicate that the ability to
reach and maintain safe shutdown has been adversely affected, prior
NRC review and approval will be obtained prior to effecting the
changes. Thus, a significant reduction in a margin of safety cannot
occur.
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: Joseph P. Mann Library, 1516
[[Page 28622]]
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: May 16, 1996. This supersedes the
October 24, 1995, request published in the Federal Register on November
27, 1995 (60 FR 58409).
Description of amendment request: This license amendment request
proposes to revise Surveillance Requirement 4.7.6.e.4 to reflect a
proposed design change to the output rating, from 15kW to 5kW, of the
charcoal filter adsorber unit heater in the pressurization system
portion of the control room emergency ventilation system (CREVS).
Surveillance Requirements 4.7.6.c.2, 4.7.6.d, and 4.9.13.b and c, are
also being revised to reflect a proposed change to the acceptance
criteria for the testing of carbon samples from the CREVS charcoal
adsorbers and the auxiliary/fuel building emergency exhaust system
charcoal adsorbers. Surveillance Requirement 4.7.7.a for the auxiliary
building portion of the auxiliary/fuel building emergency exhaust
system is also affected by this proposed change. However, since
Surveillance Requirement 4.7.7.a refers to Surveillance Requirements
4.9.13.b and c, no changes to 4.7.7.a are required.
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 design function of the filter adsorber unit heater in the
pressurization system portion of CREVS is to reduce the relative
humidity of the air entering the charcoal filter beds to 70%
relative humidity. Although the original design specified a heater
with a rating of 15 kW, review of the design basis calculation for
this system indicates that only about 3.13 kW is actually required
(including applicable margins to allow for voltage variations). The
proposed change to the CREVS heaters output rating from 15 kW to 5
kW will not affect the method of operation of the system, and the
new heater capacity will still exceed filter operational
requirements and safety margin. Neither the heater change nor the
charcoal testing protocol changes will affect system operation or
performance, nor do they affect the probability of any event
initiators. These changes do not affect any Engineered Safety
Features actuation setpoints or accident mitigation capabilities.
Therefore, the proposed changes will not significantly increase the
consequences of an accident or malfunction of equipment important to
safety previously evaluated in the USAR [Updated Safety Analysis
Report].
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The requested change to the CREVS heaters' output rating and the
changes to the charcoal sample testing protocol will not affect the
method of operation of the systems, and the new heater capacity will
still exceed filter operational requirements and safety margin by a
significant amount. The proposed changes only affect the heater size
in the system and the testing criteria for the charcoal samples. No
new or different accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of these changes. Therefore, the possibility of a new or
different kind of accident other than those already evaluated will
not be created by this change.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The requested change to the CREVS heaters' output rating will
reduce the heater output of the system, but the new heater output
will still exceed filter operational requirements and safety margin
by a significant amount. In addition, the reduction in heat load
output from the heater will increase the design margin between the
cooling capacity of the system air conditioning units and the
building heat load. The new charcoal adsorber sample laboratory
testing protocol is more stringent than the current testing practice
and more accurately demonstrates the required performance of the
adsorbers following a design basis LOCA [loss-of-coolant accident].
Therefore, these changes will not reduce the margin of safety of the
HVAC [heating, ventilation, and air conditioning] systems'
operation.
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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
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.
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, 1996, as
supplemented March 4, April 3 and April 10, 1996.
Brief description of amendments: The amendments revise the
Technical Specification so that the containment integrated leak rate
Type A testing will now be performed consistent with the revised 10 CFR
Part 50, Appendix J, Option B, by referring to Regulatory Guide 1.163,
``Performance-Based Containment Leak-Test Program.'' No
[[Page 28623]]
changes to implement Option B for the Type B and Type C tests were
requested by the licensee at this time.
Date of issuance: May 13, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 144 and 138
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 21, 1996 (61 FR
3498); and April 10, 1996 (61 FR 15988) The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
May 13, 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
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: March 5, 1996
Brief description of amendments: These amendments delete the
requirement to perform a pressurizer heater surveillance test and
change the requirement for containment visual inspection to prevent
sump clogging. These changes are in accordance with selected line items
from NRC Generic Letter 93-05, ``Line-Item Technical Specification
Improvements to Reduce Surveillance Requirements for Testing During
Power Operation.''
Date of issuance: May 13, 1996
Effective date: May 13, 1996
Amendment Nos. 184 and 178Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61
FR15989) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 13, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498, South Texas Project, Unit 1, Matagorda County,
Texas
Date of amendment request: January 22, 1996, as supplemented by
letter dated April 18, 1996.
Brief description of amendment: The amendment modified the steam
generator tube plugging criteria in Technical Specification 3/4.4.5,
Steam Generators, and the associated Bases, to allow the implementation
of alternate steam generator tube plugging criteria for the tube-to-
tubesheet joints (known in the industry as F*) for Unit 1.
Date of issuance: May 14, 1996Effective date: May 14, 1996
Amendment No.: 82
Facility Operating License No. NPF-76: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7553) The additional information contained in the supplemental
letter dated April 18, 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 amendments is contained in a Safety
Evaluation dated May 14, 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
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of application for amendment: February 9, 1996, as
supplementedMarch 15, 1996, and April 22, 1996.
Brief description of amendment: The amendment revised the
Administrative Controls Section 5.6.6 of the Ginna Technical
Specifications to incorporate a reference to the methodology for
determining pressure/temperature and low-temperature overpressure
protection limits.
Date of issuance: May 23, 1996
Effective date: May 23, 1996
Amendment No.: 64
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7557) The March 15, 1996, and April 22, 1996, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated May 23, 1996.No significant hazards consideration comments
received: No
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of application for amendment: February 9, 1996
Brief description of amendment: This amendment changes the
setpoints for the steam generator water level-high feedwater isolation
function.Date of issuance: May 20, 1996
Effective date: May 20, 1996
Amendment No.: 63
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7558) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 20, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Saxton Nuclear Experimental Corporation (SNEC), Docket No. 50-146,
Saxton Nuclear Reactor Facility (SNEF)
Date of application for amendment: November 21, 1995, as
supplemented on March 13, 1996.
Brief description of amendment: The amendment adds GPU Nuclear
Corporation as a licensee for the SNEF along with SNEC and transfers
all management-related responsibilities for the SNEF from SNEC to GPU
Nuclear Corporation.
Date of issuance: May 10, 1996
Effective date: May 10, 1996
Amendment No.: 13Amended Facility License No. DPR-4: Amendment
changed the Technical Specifications.
Date of initial notice in Federal Register: January 31, 1996 (61 FR
3502). The Commission also published a notice of consideration of
transfer of control of license pursuant to 10 CFR 50.80 on March 19,
1996 (61 FR 11231). The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated May 10, 1996.o
significant hazards consideration comments received: No
Local Public Document Room Location: Saxton Community Library, 911
Church Street, Saxton, Pennsylvania 16678
[[Page 28624]]
South Carolina Electric & Gas Company, South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: December 8, 1995
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to: 1) add a new surveillance requirement to
4.1.2.2, 2) delete 3.1.2.3 and 3.1.2.4, revise 3.4.9.3 to assure that
only one charging pump is capable of injecting water into the primary
coolant whenthe reactor is in a shutdown mode, 4) add a new
surveillance requirement to 4.4.9.3, 5) revise the Emergency Core
Cooling Water System pump testing acceptance criteria, and 6) revise
the BASES supporting the above changes.
Date of issuance: May 10, 1996
Effective date: 30 days after issuance
Amendment No.: 134
Facility Operating License No. NPF-12: Amendment revises the TS.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1635) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 10, 1996.No significant hazards
consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama.
Date of amendments request: December 19, 1995, as supplemented by
letters dated January 5, 1996 and May 3, 1996.
Brief description of amendments: The amendments replace the
requirements associated with the control room emergency ventilation
system contained in Technical Specification Section 3/4.7.7 with
requirements related to the operation of the control room emergency
filtration/pressurization system and the control room air conditioning
system. In addition, a one-time extension to the allowable outage time
for the control room recirculation filtration system is included to
facilitate implementation of design modifications.
Date of issuance: May 21, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 119 and 111
Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1637) The January 5, 1996 and May 3, 1996 letters provided clarifying
information that did not change the scope of the December 19, 1995,
application and 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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: April 23, 1996
Brief description of amendment: The amendment would allow steam
generator tubes to remain in service with bands of axial degradation in
the tube sheet region, for the remainder of Cycle 11, provided
sufficient undegraded tubing remains to satisfy the L*-type
criteria restrictions established by the licensee.
Date of issuance: May 20, 1996
Effective date: May 20, 1996
Amendment No.: 110
Facility Operating License No. NPF-8. The amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes (61 FR 19092). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by May 30, 1996, but indicated that if the Commission makes a
final no significant hazards consideration determination, any such
hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated May
20, 1996.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
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
[[Page 28625]]
opportunity for public comment. If comments have been requested, it is
so stated. In either event, the State has been consulted by telephone
whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By July 5, 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).
[[Page 28626]]
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units Nos. 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: May 15, 1996
Brief description of amendments: The amendment revised Surveillance
Requirement (SR) 4.5.2.d.2 in Technical Specification 3/4 5.2 to state
that the trisodium phosphate (TSP) contained in the storage baskets in
containment is in the form of anhydrous TSP, rather than dodecahydrate
TSP, as currently specified.
Date of issuance: May 15, 1996
Effective date: May 15, 1996
Amendment Nos.: Unit 1 - 107; Unit 2 - 99; Unit 3 - 79
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendment revised the Technical Specifications.Public comments
requested as to proposed no significant hazards consideration: No.The
Commission's related evaluation of the amendments, finding of emergency
circumstances, and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated May 15, 1996.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: William H. Bateman
Dated at Rockville, Maryland, this 29th day of May 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation
[Doc. 96-13878 Filed 6-4-96; 8:45 am]
BILLING CODE 7590-01-9
