[Federal Register Volume 60, Number 234 (Wednesday, December 6, 1995)]
[Notices]
[Pages 62485-62503]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-31206]
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UNITED STATES 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 November 10, 1995, through November 24,
1995. The last biweekly notice was published on November 27, 1995 (60
FR 58395).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
[[Page 62486]]
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 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 January 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.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
[[Page 62487]]
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.
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 amendments request: May 2, 1995
Description of amendments request: The proposed change revises the
large- break loss-of-coolant accident (LOCA) dose consequences. The
large-break LOCA dose calculation is being changed to include an
additional release path through allowable steam generator tube leakage
to the atmospheric dump valves (ADVs) or turbine bypass valves (TBVs).
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 probability or consequences of an accident previously
evaluated are not significantly increased by this change to the
large break LOCA dose consequences. This change has no effect on the
LOCA safety analysis for emergency core cooling system performance,
which demonstrates conformance to the acceptance criteria of 10 CFR
50.46, as described in the PVNGS Updated Final safety Analysis
Section 6.3.3. This change has no effect on structures, systems or
components prior to a LOCA or any other accident. The new
radiological consequences of the revised large break LOCA dose
calculation are below 10 CFR 100 limits for the exclusion area
boundary (EAB) and low population zone (LPZ), and the 10 CFR 50,
Appendix A, GDC 19 limits for the control room, as shown in Table 1-
1, Column C. The NRC has previously approved changes to the PVNGS
LOCA dose consequences with the acceptance criteria that the doses
are still within the guidelines set forth in 10 CFR 100 and GDC 19.
This acceptance criteria is described in the Safety Evaluation
related to amendment Nos. 64, 50, and 37 to PVNGS Units 1, 2, and 3
respectively, dated September 8, 1992.
The LOCA dose calculation is being changed to include an
additional release path through allowable steam generator tube
leakage to the ADVs or TBVs. This change is necessary to reflect a
revised calculation assumption that, following a large break LOCA,
the secondary system pressure would fall below reactor coolant
system pressure and containment pressure when operators cooldown the
steam generators by using ADVs or the TBVs (in accordance with the
safety analysis and EOPs [emergency operating procedures]). It is
desirable to use the ADVs or TBVs to vent secondary system steam and
thus reduce heat input to the reactor coolant system following a
large break LOCA. No other LOCA analysis assumptions are being
changed, and no changes are being made to structures, systems,
components or procedures.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This change has no impact on any structures, systems,
components, or procedures. The only impact is the revised
radiological consequences of a large break LOCA to include an
additional release path, as discussed in the response to Standard 1
above. 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.
This change to the large break LOCA dose consequences does not
involve a significant reduction in a margin of safety. The new
radiological consequences of the revised large break LOCA dose
calculation are below 10 CFR 100 limits for the EAB and LPZ, and the
10 CFR 50, Appendix A, GDC 19 limits for the control room, as
described in the response to Standard 1 above. The NRC has
previously approved changes to the PVNGS LOCA dose consequences with
the acceptance criteria that the doses are still within the
guidelines set forth in 10 CFR 100 and GDC 19. This acceptance
criteria is described in the Safety Evaluation related to amendment
Nos. 64, 50, and 37 to PVNGS Units 1, 2, and 3 respectively, dated
September 8, 1992. No equipment qualification is affected by the new
assumption of a release path through the secondary system following
a large break LOCA, and no post LOCA radiation zones will be
changed. This change has no impact on any structures, systems,
components, or procedures.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
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
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: November 22, 1995
Description of amendment request: The current Technical
Specifications (TS) Section 3.3.4.2 describes the limiting condition
during which components in the Service Water (SW) system may be
inoperable. The TS Section 3.3.4.2 states, in part, ``During power
operation, the requirements of 3.3.4.1 may be modified to allow any one
of the following components to be inoperable provided the remaining
systems are in continuous operation.'' The proposed change will delete
the qualifying statement,''... provided the remaining systems are in
continuous operation,'' from TS Section 3.3.4.2. Currently, this
statement requires the ``remaining systems to be in continuous
operation'' while allowing one SW loop header, or one SW pump, or one
SW booster pump to be inoperable for a period of 24 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 change would remove the requirement for the
remaining SW system components to be in continuous operation while
one TS-required component is inoperable. Rather, the remaining
components would remain operable, and no change would be made in
normal system operation. The SW system provides an accident
mitigation function and is not involved in accident initiation
sequences. Therefore, the proposed change would not involve a
significant increase in the probability of an accident previously
evaluated.
The capacity of the SW system is such that its accident
mitigation function can be performed by operation of a maximum of
two SW pumps, one SW booster pumps, and one SW header. While a TS-
required component is inoperable, sufficient accident mitigation
capability is provided by the remaining operable components, rather
than requiring the remaining systems to be in continuous operation.
Therefore, the proposed change would not cause a significant
increase in the 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 change would remove the requirement for the
remaining SW system
[[Page 62488]]
components to be in continuous operation while one TS-required
component is inoperable. Rather, the remaining components would
remain operable. The proposed change would not change the normal
operation of the system, nor would any physical modifications result
from the change. The function and capability of the SW systems would
remain unchanged. Therefore, the proposed change would 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 the margin of safety.
The proposed change would remove the requirement for the
remaining SW system components to be in continuous operation while
one allowed TS-required component is inoperable. Rather, the
remaining TS-required components would remain operable. Adequate
assurance of operability is maintained by performance of regular
surveillance testing. Maintaining operable status rather than
placing equipment in continuous operation does not result in a
change in the ability of the SW system to perform its intended
function, since the system provides an automatic response to
accident conditions, and the system possesses adequate capacity to
perform its normal operating function with one allowed TS-required
component inoperable. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: October 24, 1995
Description of amendment request: The proposed amendment will
increase the trip setpoints and allowable values for the low power
block (P-7).
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 10CFR50.92, CYAPCO has reviewed the proposed
change and has concluded that it does not involve a significant
hazards consideration (SHC). The basis for this conclusion is that
the three criteria of 10CFR50.92(c) are not compromised. The
proposed change does not involve an SHC because the change would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change will relax the power level values for the P-
7 interlock by 2 percent. This change affects both the P-7 and P-7N
interlocks. The P-7 interlock affects reactor trips on 1) low flow
in more than one reactor coolant loop, 2) reactor coolant pump bus
under voltage, 3) more than one reactor coolant pump breaker open,
4) main steam line isolation valve closure, 5) turbine trip, and 6)
variable low pressure. The P-7 interlock automatically blocks these
reactor trips on decreasing power and automatically unblocks these
reactor trips on increasing power. The P-7N interlock affects the
reactor trip on wide range, neutron flux, high startup rate. P-7N
automatically enables this reactor trip on decreasing power level
and automatically blocks this reactor trip on increasing power
level. The Applicable Modes requirement and Action Statements for
the P-7 interlock and the reactor trips associated with both P-7 and
P-7N in the Instrumentation Channel and Surveillance Requirements of
Technical Specification 3/4.3.1 are being changed by 2 percent to be
consistent with the change to P-7. The interlock setpoint cannot
cause an accident. Also, the proposed 2 percent increase in the
power level still results in a power level well below the power
level at which the P-7 interlocked reactor trips are required for
accident mitigation, as well as maintaining the high startup rate
trip enabled at a higher power level. This proposed power level is
consistent with the technical specification requirement prior to the
conversion to standard format technical specifications and is also
consistent with the Standard Westinghouse technical specification
value. Therefore, the proposed change can neither increase the
consequences of the design basis accident nor the probability of
occurrence of the design basis accidents.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change only modifies the power level for the P-7
and P-7N interlocks. The proposed setpoint is a power level at which
stable plant conditions are easier to maintain while transferring
the power supply for the reactor coolant pumps between offsite power
and the main generator. The setpoint is also well below the power
level for which the reactor protection afforded by the trips that
are bypassed by P-7 is needed. This cannot create the possibility of
a new or different kind of accident from any previously analyzed.
3. Involve a significant reduction in a margin of safety.
The proposed change maintains the power level for the P-7
interlock below the power level for which the reactor trips that are
blocked by the P-7 interlock are required. It also raises the power
level to a value at which it is easier to maintain stable plant
conditions. This will reduce the likelihood of an automatic reactor
trip during the transferring of power for the reactor coolant pumps
between offsite power and the main generator. The proposed change
will result in the high startup rate reactor trip being enabled at a
higher power level. This is conservative since it expands the range
of coverage for the trip. Therefore, the proposed change does not
impact 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: November 1, 1995
Description of amendment request: The proposed amendment will
modifiy Surveillance Requirement 4.6.3.2, ``Containment Isolation
Valves,'' (CIVs) to change the surveillance interval from at least once
per 18 months to at least once per refueling interval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
CYAPCO has reviewed the proposed change in accordance with
10CFR50.92 and concluded that the change does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed change does not involve an SHC because the
change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to Surveillance Requirement 4.6.3.2 of the
Haddam Neck Plant Technical Specifications extends the frequency for
verifying that each CIV actuates to its required position in
response to a safety injection actuation test signal. The proposal
would extend the frequency from at least once per 18 months to at
least once per
[[Page 62489]]
refueling interval (24 months + 25% as allowed by Technical
Specification 4.0.2).
The proposed change to Surveillance Requirement 4.6.3.2 does not
alter the intent or method by which the surveillance is conducted,
does not involve any physical changes to the plant, does not alter
the way any structure, system, or component functions, and does not
modify the manner in which the plant is operated.
Additional assurance of CIV operability is provided by
Surveillance Requirement 4.6.3.3. Surveillance Requirement 4.6.3.3
requires the confirmation of the mechanical operability of the CIVs
by the inservice inspection program. The proposed change does not
modify these requirements.
Equipment performance over the last four operating cycles was
evaluated to determine the impact of extending the frequency of
Surveillance Requirement 4.6.3.2. This evaluation included a review
of surveillance results, preventive maintenance records, and
corrective maintenance records. It has been concluded that the CIVs
are highly reliable, and that there is no indication that the
proposed extension could cause deterioration in valve condition or
performance.
As such, the proposed change to the frequency of Surveillance
Requirement 4.6.3.2 will not degrade the ability of the CIVs to
perform their safety function.
Based on the above, the proposed change to Surveillance
Requirement 4.6.3.2 of the Haddam Neck Plant Technical
Specifications does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change to Surveillance Requirement 4.6.3.2 of the
Haddam Neck Plant Technical Specifications extends the frequency for
verifying that each CIV actuates to its required position in
response to a safety injection actuation test signal. The proposal
would extend the frequency from at least once per 18 months to at
least once per refueling interval (24 months + 25% as allowed by
Technical Specification 4.0.2).
The proposed change does not alter the intent or method by which
the surveillance is conducted, does not involve any physical changes
to the plant, does not alter the way any structure, system, or
component functions, and does not modify the manner in which the
plant is operated. As such, the proposed change in the frequency of
Surveillance Requirement 4.6.3.2 will not degrade the ability of the
CIVs to perform their safety function.
Based on the above, the proposed change to Surveillance
Requirement 4.6.3.2 of the Haddam Neck Plant Technical
Specifications will not create the possibility of a new or different
kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to Surveillance Requirement 4.6.3.2 of the
Haddam Neck Plant Technical Specifications extends the frequency for
verifying that each CIV actuates to its required position in
response to a safety injection actuation test signal. The proposal
would extend the frequency from at least once per 18 months to at
least once per refueling interval (24 months + 25% as allowed by
Technical Specification Section 4.0.2).
The proposed change does not alter the intent or method by which
the surveillance is conducted, does not involve any physical changes
to the plant, does not alter the way any structure, system, or
component functions, and does not modify the manner in which the
plant is operated. As such, the proposed change in the frequency of
Surveillance Requirement 4.6.3.2 will not degrade the ability of the
CIVs to perform their safety function.
Additional assurance of the operability of the CIVs is provided
by Surveillance Requirement 4.6.3.3.
Equipment performance over the last four operating cycles was
evaluated to determine the impact of extending the frequency of
Surveillance Requirement 4.6.3.2. This evaluation included a review
of surveillance results, preventive maintenance records, and
corrective maintenance records. It has been concluded that the CIVs
are highly reliable, and that there is no indication that the
proposed extension could cause deterioration in valve condition or
performance.
Based on the above, the proposed change to Surveillance
Requirement 4.6.3.2 of the Haddam Neck Plant Technical
Specifications 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: 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. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan
Date of amendment request: November 8, 1995, as supplemented
November 17, 1995
Description of amendment request: The proposed amendment would
remove the prescriptive Type A containment leakage test rate frequency
of 40 plus or minus 10 months and add a reference to perform
containment leakage rate tests in accordance with the criteria
specified in Appendix J of 10 CFR Part 50 as amended by approved
exemptions. In addition, the proposed amendment would revise the test
pressure for Type B and C testing to correct a typographical error.
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:
Leakage test rate frequency
1) The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This change is administrative in nature and does not impact
plant systems, structures or components. The proposed change will
allow the facility's technical specifications to be revised to allow
containment sphere leakage testing in accordance with Appendix J to
10 CFR Part 50 as modified by approved exemptions.
2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This change is administrative in nature and does not impact
plant syst ems, structures or components. The proposed change will
allow the facility's technical specifications to be revised to allow
containment sphere leakage testing in accordance with Appendix J to
10 CFR Part 50 as modified by approved exemptions.
3)The proposed change does not involve a significant reduction
in a margin of safety.
This change is administrative in nature and does not impact
plant systems, structures or components. The underlying purpose of
Appendix J is still achieved. Appendix J states that the leakage
test requirements provide for periodic verification testing of the
leak tightness integrity of the primary reactor containment. The
appendix further states that the purpose of the tests is to assure
that leakage through the primary containment shall not exceed the
allowable leakage rate values as specified in the technical
specifications or associated bases. As stated previously, for Big
Rock Point and a large percentage of other plants, the Appendix J
Type B and C testing programs provide the most significant and
meaningful assessment of containment leak tightness. The testing
history and structural capability of the containment establish that
there is significant assurance that the extended interval between
Type A tests will not adversely impact the integrity of the
containment.
Test pressure revision
As stated in the technical specification change request, this
revision is being performed to be consistent with accident pressure,
Pa, used for Big Rock Point. 20 psig is a typographical error.
23 psig has always been used for these tests.
The proposed change does not:
1) involve a significant increase in the probability or
consequences of an accident previously evaluated.
This change is administrative in nature and does not impact
plant systems, structures or components. The proposed change will
allow the facility's technical specifications to be revised to
reflect current containment sphere leakage testing in accordance
with Appendix J to 10 CFR Part 50.
[[Page 62490]]
2) create the possibility of a new or different kind of accident
from any accident previously evaluated.
This change is administrative in nature and does not impact
plant systems, structures or components. The proposed change will
allow the facility's technical specifications to be revised to
reflect current containment sphere leakage testing in accordance
with Appendix J to 10 CFR Part 50.
3) involve a significant reduction in a margin of safety.
This change is administrative in nature and does not impact
plant systems, structures or components.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: Brian E. Holian, Acting
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of amendment request: October 16, 1995
Description of amendment request: Appendix J of 10 CFR Part 50,
``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors,'' has recently been revised to include Option B. This option
allows the implementation of a performance based Type B and C testing
program. The proposed change will add a footnote to Technical
Specification (TS) 4.6.1.2.d stating that the Type B and C tests
scheduled for Unit 1 refueling outage Cycle 6 (1R6) will be conducted
in accordance with Option B and using the guidance of Regulatory Guide
1.163, Revision 0. This option is being incorporated into the
licensee's request to implement the improved TS. However, the improved
TS are not scheduled to become effective until after the Unit 1
refueling outage 1R6.
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 change does not involve a significant increase in
probability or consequences of an accident previously evaluated. The
proposed change does not involve a change to structures, systems, or
components which would affect the probability or consequences of an
accident previously evaluated in the Vogtle Electric Generating
Plant (VEGP) Final Safety Analysis Report (FSAR). The proposed
change only provides a mechanism within the Technical Specifications
for implementing a performance-based method of determining the
frequency for leak rate testing which has been approved by the NRC
via a revision to 10 CFR 50, Appendix J.
2. The proposed change will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The amendment will not change the design, configuration, or method
of plant operation. It only allows for the implementation of Option
B of 10 CFR 50, Appendix J for Unit 1 refueling outage 1R6 without
violating the plant Technical Specifications.
3. Operation of VEGP, Unit 1, in accordance with the proposed
change will not involve a significant reduction in the margin of
safety. The proposed change does not affect a safety limit, an LCO
[limiting condition for operation], or the way plant equipment is
operated. The NRC is aware that changes similar to this proposed
change are required in order to implement Option B of 10 CFR 50,
Appendix J. In fact, the staff indicates in Paragraph V.B. of
Appendix J that Option B or parts thereof may be adopted by a
licensee 30 days after the rule becomes effective by submitting
notification of its implementing plan and a request for revision to
Technical Specifications. Since the NRC has approved the provision
for performance-based testing and must approve this Technical
Specification[] change before the performance-based Option B can be
implemented, the margin of safety will not be significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308
NRC Project Director: Herbert N. Berkow
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 30, 1995 (noticed in the Federal
Register July 5, 1995, (60 FR 35080) as supplemented by letter dated
November 20, 1995
Description of amendment request: The proposed amendment would
revise the Technical Specifications as follows:
1. The Surveillance Frequency for the drywell bypass test is
changed from 18 months to 10 years with an increased testing frequency
required if performance degrades.
2. The following changes are requested for the drywell air lock
testing: (a) the leakage rate surveillance is moved from the air lock
Limiting Condition for Operation (LCO) to the drywell LCO, (b) the
requirement for the air lock to meet a specific overall leakage limit
is deleted, (c) the Note that an inoperable air lock door does not
invalidate the previous air lock leakage test is deleted, (d) the Note
which required that the air lock leakage test at 3 psid be preceded by
pressurizing the air lock to 19.2 psid is moved to the bases, and (e)
the Surveillance Frequency for the air lock leakage test and interlock
test is changed from 18 months to 24 months.
3. The Actions Notes in the drywell air lock LCO and the drywell
isolation valve LCO that identifies that the Actions required by the
drywell LCO must be taken when the drywell bypass leakage limit is not
met is deleted.
4. The requirement for the drywell air lock seal leakage rate to
meet a specific leakage limit is deleted.
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 for River Bend Station (RBS) and Grand Gulf Nuclear
Station (GGNS), which is presented below:
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The requested changes are either administrative changes which
clarify the format of the requirement or change the requirement to
match the design bases of the plant, a change which relocates the
requirement to the Technical Specification Bases, or a change in
surveillance interval. Each of these types of change are discussed
below:
1. The administrative changes clarify the format of the
requirement or change therequirement to match the design bases of
the plant. Clarifying administrative format of
[[Page 62491]]
the Technical Specifications does not result in any changes to the
Technical Specification requirements and, as a result, does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. Also, changing the requirements of
the Technical Specifications to more closely match the design bases
of the plant will continue to assure that the plant will respond as
assumed in the accident analyses and, as a result, does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes relocate information to the Technical
Specification Bases. In the Technical Specifications Bases the
relocated information will be maintained in accordance with 10 CFR
50.59 and subject to the change control provisions in Chapter 5 of
Technical Specifications. Since any changes to the Technical
Specifications Bases will be evaluated per the requirements of 10
CFR 50.59, no increase (significant or insignificant) in the
probability or consequences of an accident previously evaluated will
be allowed. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
3. The proposed changes in frequency for the drywell bypass
leakage and drywell air lock surveillances will continue to ensure
that no paths exist through passive drywell boundary components that
would permit gross leakage from the drywell to the primary
containment air space and result in bypassing the primary
containment pressure-suppression feature beyond the design basis
limit. The Mark III primary containment system satisfies General
Design Criterion 16 of Appendix A to 10 CFR Part 50. Maximum drywell
bypass leakage was determined previously by reviewing the full range
of postulated primary system break sizes. The limiting case was a
primary system small break loss of coolant accident (LOCA) and
yielded a design allowable drywell bypass leakage rate limit of
approximately 35,000 scfm for GGNS and 46,000 scfm (the Technical
Specification limit is based on a lower limit of 40,110 scfm) for
RBS. The Technical Specifications acceptable limit for the bypass
leakage following a surveillance is less than 10% of this design
basis value. The most recent bypass leakage value was approximately
2.5% for GGNS and .91% for RBS of the design allowable leakage rate
limit for the limiting event. EOI is committed to maintaining
programmatic and oversight controls that ensure that drywell bypass
leakage remains a small fraction of the design allowable leakage
limit.
The drywell is typically exposed to essentially 0 psig during
normal plant operation and 3 psig during drywell bypass leak rate
testing. These pressures are considerably lower than the structural
integrity test pressure and are less likely to initiate a crack or
cause an existing crack to grow. Visual inspections of the
accessible drywell surfaces that have been performed since the
structural integrity tests have not revealed the presence of
additional cracking or other abnormalities. Therefore, additional
cracking of the drywell structure is not expected due to testing or
operation and, similar to the justification for the ten year 10 CFR
50 Appendix J Type A test interval, it is not considered credible
for the passive drywell structure to begin to leak sufficiently to
impact the design drywell bypass leakage limit.
The primary containment's ability to perform its safety function
is fairly insensitive to the amount of drywell leakage, thereby
providing a margin to loss of the drywell safety function that is
not normally available for safety systems. This insensitivity is
demonstrated by the extremely high limiting event design basis
allowable leakage for the drywell (e.g., 35,000 scfm for GGNS and
46,000 scfm for RBS). The limiting leakage is almost an order of
magnitude higher for other events. Additionally, an even higher
allowable leakage can be realistically accommodated by the primary
containment due to the margins in the containment design. Because of
the margins available, it will take valves in multiple penetration
flow paths leaking excessively to cause the primary containment to
fail as a result of overpressurization, the probability that drywell
isolation valve leakage will result in primary containment failure
due to excessive drywell leakage is not considered significant and
this drywell/primary containment failure mode is not considered
credible.
The proposed Technical Specification changes have no significant
impact on the GGNS Individual Plant Examination (IPE) or the RBS IPE
conducted per NRC Generic Letter 88-20. The IPEs considered
overpressurization failure of primary containment as part of the
primary containment performance assessment. Due to the magnitude of
acceptable drywell leakage and the extremely low probabilities of
achieving such leakage, primary containment failure due to
preexisting excessive drywell leakage was considered a non
significant contributor to primary containment failure. Primary
containment overpressurization failure can occur with or without
preexisting excessive drywell leakage in a severe accident. This is
due to physical phenomena associated with potentially extreme
environmental conditions inside primary containment following a
severe accident. However, the calculated frequency of such extreme
conditions is very small. The proposed changes do not impact the IPE
evaluated phenomena causing primary containment overpressurization
failure nor significantly increase the probability that the drywell
has preexisting excessive leakage and therefore would not contribute
to these accident scenarios.
For the reasons discussed above, the proposed changes do not
have any significant risk impact to accidents previously evaluated
and do not significantly increase the consequences of an accident
previously evaluated. Additionally, drywell leakage is not the
initiator of any accident evaluated; therefore, changes in the
frequency of the surveillance for drywell leakage does not increase
the probability of any accident evaluated.
Therefore, the proposed changes do 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 requested changes are either administrative changes which
clarify the format of the requirement or change the requirement to
match the design bases of the plant, a change which relocates the
requirement to the Technical Specification Bases, or a change in
surveillance interval. Each of these types of change are discussed
below:
1. The administrative changes in the Technical Specification
requirements do not involve a physical alteration of the plant (no
new or different type of equipment will be installed) nor does it
change the methods governing normal plant operation. Thus, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
2. The proposed relocation of requirements does not involve a
physical alteration of the plant (no new or different type of
equipment will be installed) nor does it change the methods
governing normal plant operation. The proposed change will not
impose or eliminate any requirements. Adequate control of the
information will be maintained in the Technical Specification Bases.
Thus, the change proposed does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. The proposed change modifies the surveillance frequency for
drywell bypass leakage and drywell air lock surveillances. The
changes only impact the test frequency and do not result in any
change in the response of the equipment to an accident. The changes
do not alter equipment design or capabilities. The changes do not
present any new or additional failure mechanisms. The drywell is
passive in nature and the surveillance will continue to verify that
its integrity has not deteriorated. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
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 requested changes are either administrative changes which
clarify the format of the requirement or change the requirement to
match the design bases of the plant, a change which relocates the
requirement to the Technical Specification Bases, or a change in
surveillance interval. Each of these types of changes are discussed
below:
1. The administrative changes in the Technical Specification
requirements do not involve a physical alteration of the plant (no
new or different type of equipment will be installed) nor does it
change the methods governing normal plant operation. Thus, this
change does not cause a significant reduction in the margin of
safety.
[[Page 62492]]
2. The relocation of requirements will not reduce a margin of
safety because it has no impact on any safety analysis assumptions.
In addition, the requirements to be transferred from the Technical
Specifications to the Technical Specifications Bases are the same as
the existing Technical Specifications. Since any future changes to
these requirements in the Technical Specifications Bases will be
evaluated per the requirements of 10 CFR 50.59, no reduction
(significant or insignificant) in a margin of safety will be
allowed.
3. The proposed change modifies the surveillance frequency for
drywell bypass leakage and associated air lock surveillances.
Reliability of drywell integrity is evidenced by the measured
leakage rate during past drywell bypass leakage surveillances.
Appropriate design basis assumptions will be upheld, even when
combined with the complementary bypass leakage surveillances as
proposed. Drywell integrity will continue to be tested by means of
the proposed periodic drywell bypass leakage test, performance of
the drywell air lock door latching and interlock mechanism
surveillance, and performance of additional surveillances including
excercising of drywell isolation valves. The combination of these
surveillances will provide adequate assurance that drywell bypass
leakage will not exceed the design basis limit. Margins of safety
would not be reduced unless leakage rates exceeded the design
allowable drywell bypass leakage limit. Therefore, the proposed
change does not cause a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 26, 1995
Description of amendment request: The proposed amendment would
revise the technical specifications for sixteen editorial changes and
would delete the requirement for a program to prevent and detect
Asiatic Clams (Corbicula) in the service water system (SWS). The
editorial changes covers such things as removing systems or components
that do not exist in the River Bend Station, correcting typographical
errors, correcting to be consistent with the writers guide for Improved
Technical Specifications, adding descriptions for systems to make them
clear, and wording changes to be consistent with approved facility
operations. The Corbicula program is no longer needed because the
facility has been modified and SWS no longer takes water from the
Mississippi River; source of the larvae and infestation.
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:
EDITORIAL CHANGES
The purposed changes involves reformatting, renumbering and
rewording of the existing Technical Specifications. The
reformatting, renumbering and rewording process involves no
technical changes to existing Technical Specifications. As such,
these changes are administrative in nature and do not impact
initiators of analyzed events or assumed mitigation of accident or
transient events. Therefore, these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes do 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 not impose or eliminate any new or different
requirements. Thus, these changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes will not reduce a margin of safety because
they have no impact on any safety analysis assumptions. These
changes are administrative in nature. As such, no question of safety
is involved, and the changes do not involve a significant reduction
in a margin of safety.
CORBICULA PROGRAM
The proposed change deletes the program associated with the
prevention and detection of Asiatic Clams (Corbicula) based upon
improvements to the non-safety related Normal Service Water System
(SWS). The source of makeup water to the SWS is no longer the
Mississippi River, which is the source of Asiatic Clams.
Demineralized water or well water is used eliminating the source of
asiatic clams. To prevent biofouling SWS is treated with chlorine/
bromine. This program is not considered as an initiator for any
previously evaluated accident. Therefore, the proposed change will
not increase the probability or consequences of any accident
previously evaluated.
The proposed change introduces no new mode of plant operation
and it does not involve a physical modification to the plant. The
possibility of the SES becoming contaminated by any other means is
highly unlikely since it is a ``closed-loop'' system. Therefore it
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Prevention of Asiatic Clam infestation in the SWS and associated
safety-related equipment is ensured by the ``closed-loop'' design of
the SWS. Post Refuel Outage (RF-4) inspections of the safety-related
heat exchangers that interface with the ``closed-loop'' SWS have
shown no evidence of clam infestations. Therefore, the change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: November 20, 1995
Description of amendment request: The proposed amendment would
revise the technical specifications to eliminate the response time
testing requirements for selected Reactor Protection System
Instrumentation.
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 purpose of the proposed Technical Specification (TS) change
is to eliminate response time testing requirements for selected
components in the Reactor Protection System (RPS). The Boiling Water
reactors Owners' Group (BWROG) has completed an evaluating which
demonstrates that response time testing is redundant to the other
TS-required testing. These other tests, in conjunction with actions
taken in response to NRC Bulletin 90-01, ``Loss of Fill-Oil in
Transmitters Manufactured by Rosemount,'' and Supplement 1, are
sufficient to identify failure modes or degradation in instrument
response times and ensure operation of the associated systems within
acceptable limits. There are no known failure modes that can be
detected by response time testing that cannot also be detected by
the other TS-required testing. This evaluation was
[[Page 62493]]
documented in NEDO-32291, ``System Analyses for Elimination of Selected
Response Time Testing Requirements,'' January 1994. Entergy
Operations, Inc. (EOI) has confirmed the applicability of this
evaluation to River Bend Station (RBS). In addition EOI will
complete the actions identified in the NRC staff's safety evaluation
of NEDO-32291.
Because of the continued application of other existing TS-
required tests such as channel calibration, channel checks, channel
functional tests, and logic system functional tests, the response
time of these systems will be maintained within the acceptance
limits assumed in plant safety analyses and required for successful
mitigation of an initiating event. The proposed changes do not
affect the capability of the associated systems to perform their
intended function within their required response time, nor do the
proposed changes themselves affect the operation of any equipment.
As a result, EOI has concluded that the proposed changes do not
involve a significant increase in the probability or the
consequences of an accident previously evaluated.
The proposed changes only apply to the testing requirements for
the components identified above and do not result in any physical
change to these or other components or their operation. As a result,
no new failure modes are introduced. Therefore, the proposed changes
do not create the possibility of a new or different kind of accident
from any accidents previously evaluated.
The current TS-required response times are based on the maximum
allowable values as assumed in the plant safety analyses. These
analyses conservatively establish the margin of safety. As described
above, the proposed changes do not affect the capability of the
associated systems to perform their intended function within the
allowed response time used as the basis for the plant safety
analyses. The potential failure modes for the components within the
scope of this request were evaluated for impact on instrument
response time. This evaluation confirmed that, with the exception of
loss of fill-oil of Rosemount transmitters, the remaining TS-
required testing is sufficient to identify failure modes or
degradation in instrument response times and ensure operation of the
instrument within the scope of this request is within acceptable
limits. The actions taken in response to NRC Bulletin 90-09 and
Supplement 1 are adequate to identify loss of fill-oil failures of
Rosemount transmitters. As a result, it has been concluded that
plant and systems response to an initiating event will remain in
compliance with the assumptions of the safety analysis.
Further, although not explicitly evaluated, the proposed changes
will provide an improvement to plant safety and operation by
reducing the time safety systems are unavailable, reducing the
potential for safety system actuations, reducing plant shutdown
risk, limiting radiation exposure to plant personnel, and
eliminating the diversion of key personnel resources to conduct
unnecessary testing. Therefore, EOI has concluded that this request
will result in an overall increase in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: September 22, 1995
Description of amendment request: The proposed amendment would
modify a requirement of the Seabrook Station, Unit No. 1 Technical
Specifications. Specifically, the proposed amendment would change the
ACTION referenced in Table 3.3-3, Engineered Safety Features Actuation
System Instrumentation, for Functional Unit 8.b, Automatic Switchover
to Containment Sump/RWST Level Low-Low. The ACTION requirement would be
changed to ACTION 15 from ACTION 18. ACTION 15 requires an inoperable
channel to be placed in bypass (with no time limit specified) while
ACTION 18 requires an inoperable channel to be placed in the tripped
condition within 6 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. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated (10
CFR 50.92(c)(1)) because the proposed change would result in an
inoperable Functional Unit 8.b. protective channel being placed in
the bypassed condition vice tripped condition. Functional Unit 8.b.
is not involved in any accident initiation sequence; therefore, the
probability of a previously-analyzed accident is not increased.
Placing an inoperable Functional Unit 8.b. in bypass vice trip
reduces the probability of premature opening of the containment
building sump isolation valves thereby reducing the potential for
increasing the consequences of a previously-analyzed accident. Thus,
the consequences of a previously-analyzed accident is not increased.
B. The change does not create the possibility of a new or
different kind of accident from any accident previously evaluated
(10 CFR 50.92(c)(2)) because the change does not reduce the minimum
required number of channels of instrumentation to be operable. The
change does not alter the function of or affect the failure modes of
Functional Unit 8.b. instrumentation channels. The proposed change
does not otherwise affect the manner by which the facility is
operated, and it does not involve any changes to equipment or
features which affect the operational characteristics of the
facility.
C. The change does not involve a significant reduction in a
margin of safety (10 CFR 50.92(c)(3)) because the change does not
reduce the minimum required number of channels of instrumentation to
be operable, and it does not involve any changes to equipment or
features which affect the operational characteristics of the
facility. Therefore, the protection previously provided remains
unchanged.
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: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London,
Connecticut
Date of amendment request: May 26, 1995, supplemented and revised
October 20, 1995.
Description of amendment request: The proposed changes would modify
TS 3.8.1.1., ``Electrical Power Systems, A.C. Sources, Operating,'' TS
3.8.1.2, ``Electrical Power Systems, Shutdown,'' TS 3.8.2.2,
``Electrical Power Systems, A.C. Distribution - Shutdown,'' and TS
3.8.2.4, ``Electrical Power Systems, D.C. Distribution - Shutdown,'' to
provide operational flexibility as well as consistency between action
statements and to eliminate certain surveillance requirements that are
not applicable in Modes 5 or 6.
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 (SHC), which is presented below:
[[Page 62494]]
In accordance with 10 CFR 50.92, NNECO has reviewed the proposed
changes and has concluded that they do not involve an SHC. The basis
for this conclusion is that the three criteria of 10 CFR 50.92(c)
are not compromised. The proposed changes do not involve an SHC
because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to Surveillance Requirement 4.8.1.1.1 is
being made because presently, the surveillance requirement for
demonstrating offsite sources are operable states that ``two''
independent circuits are required. The surveillance requirement is
referenced for both operating and shutdown modes. While it is
accurate for operating modes, it is inconsistent with the limiting
condition for operation for shutdown. The proposed change is safe
because it renders the surveillance requirement consistent with the
applicable limiting condition for operation (i.e., operating or
shutdown) and eliminates a potential source of confusion.
The change to Surveillance Requirement 4.8.1.2 and Technical
Specification 3.8.2.2 merely clarifies the diesel generator
surveillance and operability requirements for Modes 5 and 6 and
renders action statements for related technical specification
sections consistent with and appropriate for operational Modes 5 and
6.
Regarding diesel generator surveillance requirements, automatic
A.C. power for LNP events in Modes 5 and 6 is not required. This is
validated by the fact that the undervoltage sensors are only
required to be operable in Modes 1, 2 and 3 to meet technical
specifications. Because the undervoltage sensors provide the logic
that results in actuation of the sequencer, it follows that the
sequencer need not be operable in Modes 5 and 6. Accordingly, the
sequencer is not required to support operability of the available
diesel generator in Modes 5 and 6. Further, because SIAS is blocked
in Modes 5 and 6, automatic start of the diesel generator upon
receipt of a SIAS is similarly not required to support operability
of the diesel generator in Modes 5 and 6.
Additionally, operation of the diesel generator in parallel with
the system during Modes 5 and 6 is not required to perform its
intended safety function. In fact, such operation may compromise
both sources as the result of a single event.
Since automatic A.C. power is not credited in the mitigation of
Mode 5 and 6 events and accidents, such as fuel handling accidents,
there is no increase in the probability or consequences of
previously evaluated accidents.
The action statement in Technical Specification 3.8.2.2 has been
revised to cite actions that are more appropriate for Modes 5 and 6
for Millstone Unit No. 2. This is due to the ability to maintain the
plant in a safe condition without needing to automatically load the
diesel generator through the sequencers in Modes 5 and 6. In
addition, the proposed change is consistent with the CE Owner's
Group Standard Technical Specification and with other Millstone Unit
No. 2 action statements. Consequently, there is no increase in the
probability or consequences of previously evaluated accidents.
The change to TS 3.8.2.4 merely renders the action statement
consistent with, and appropriate for, operational Modes 5 and 6.
Since D.C. power is not credited in the mitigation of Mode 5 and
6 events and accidents, such as fuel handling accidents, there is no
increase in the probability or consequences of previously evaluated
accidents.
The action statement in TS 3.8.2.4 has been revised to cite
actions that are more appropriate for Modes 5 and 6 for Millstone
Unit No. 2. This is due to the ability to maintain the plant in a
safe condition without D.C. power distribution available in Modes 5
and 6. In addition, the proposed change is consistent with the CE
Owner's Group Standard Technical
Specifications (NUREG-1432) and with other Millstone Unit No. 2
action statements. Consequently, there is no increase in the
probability or consequences of previously evaluated accidents.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed changes do not alter or affect the design,
function, failure mode, or operation of the plant. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to the technical specifications provides
greater consistency between the action statements and clarifies
which surveillance requirements are required in Modes 5 and 6. Since
the diesel generators are not required to be loaded automatically in
Modes 5 and 6, and since it is part of our shutdown risk management
program to assure that adequate cooling is able to be provided, and
since the diesel will still be verified to start and achieve rated
speed, the proposed changes to the technical specifications do not
reduce the margin of safety.
The proposed change to the TS provides greater consistency among
action statements during Modes 5 and 6. Since the D.C. distribution
system is not credited in the mitigation of Mode 5 and 6 events and
accidents, and since it is part of our shutdown risk management
program to assure that adequate fuel cooling is able to be provided,
the proposed change to the TS does not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
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
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: June 27, 1995, as supplemented July 21,
1995
Description of amendment request: The amendment revises the
Technical Specifications (TS) to relocate TS requirements for the
containment purge exhaust and supply valves, and to remove a duplicate
testing requirement for the safety injection input from engineered
safety features from the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
... The proposed changes do not involve an SHC [significant
hazards consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The first proposed change relocates the operability and
surveillance requirements for the containment high range radiation
monitors from Technical Specification Section 3.3.3 to Technical
Specification Section 3.3.2. The proposed changes are administrative
in nature. The proposed changes do not alter the way any structure,
system, or component functions and do not modify the manner in which
the plant is operated and do not involve any physical changes to the
plant.
The second proposed modification will delete the testing
requirement for functional unit 16, ``Safety Injection Input from
ESF,'' of Table 4.3-1 because the logic circuitry that processes
the safety injection signals and produces a reactor trip is
tested under functional unit 19 ``Automatic Trip and Interlock
Logic,'' and the testing is performed on a more frequent basis
(i.e., on a monthly staggered bases versus on an 18-month
frequency). In addition, the same logic testing is accomplished with
an 18-month TADOT of functional unit 1.a of Table 4.3-2 and with a
monthly staggered actuation logic testing of functional unit 16 of
Table 4.3-2. This testing ensures that operability of the logic
under functional unit 16 of Table 4.3-1 is verified. The other tests
will continue to verify the operability of the reactor trip system
and that a reactor trip will be initiated when required.
Therefore, there is no change in the potential for an increase
in the consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
[[Page 62495]]
The proposed changes do not affect the operation or response of
any plant equipment or introduce any new failure mechanisms. The
proposed elimination of the testing requirement line item does not
affect the test results since the logic circuitry that processes the
safety injection signal and produces a reactor trip will be tested
and is tested under functional unit 19 of Table 4.3-1. As such, the
changes do not create the possibility of a new or different kind of
accident previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed changes do not have any adverse impact on the
protective boundaries nor do they affect the consequences of any
accident analyzed. The operability and surveillance requirements,
although relocated to other technical specifications, will still
ensure that the system (the radiation monitors) is tested and within
limits. The proposed elimination of the testing equipment will not
change the performance or operating conditions of the safety
systems. The operable reactor trip system instrumentation ensures
that the assumptions in the Bases of the Technical Specifications
are not affected and ensures that the margin of safety is not
reduced. Therefore, the proposed changes do not reduce the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: November 14, 1994
Description of amendment requests: The proposed amendment would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Nuclear Power Plant, Unit Nos. 1 and 2, for the slave relay test
frequency from quarterly (Q) to refueling (R). The request would also
remove table notation 4 from Table 4.3-2. The associated Bases would
also be appropriately revised.
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 results of WCAPs 14117 and 13878 demonstrate that slave
relays are highly reliable. The WCAPs also provide guidance to
assure that slave relays remain highly reliable. The aging
assessment concludes that the age/temperature-related degradation of
all ND relays, and NE relays produced after May 1990, is
sufficiently slow such that a refueling frequency surveillance
interval will not significantly increase the probability of slave
relay failures. Finally, the evaluation of the interposing slave
relays in the emergency diesel generator start circuitry, control
room ventilation and auxiliary building ventilation realignments,
steam generator blowdown isolation and radwaste isolation systems
has concluded that based on the tests of the interposing relays
performed during other equipment testing, reasonable assurance is
provided that failures will be identified if the associated slave
relays are tested on a refueling frequency.
The removal of table notation 4 from TS Table 4.3-2 is an
administrative change that eliminates unnecessary redundancy from
the TS and does not affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not alter the performance of the ESFAS
mitigation systems assumed in the plant safety analysis. Changing
the interval for periodically verifying ESFAS slave relays (assuring
equipment operability) will not create any new accident initiators
or scenarios.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated for DCPP.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the total ESFAS response
assumed in the safety analysis since the reliability of the slave
relays will not be significantly affected by the increased
surveillance frequency.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman
South Carolina Electric & Gas Company, South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: August 18, 1995, as supplemented on
November 1, 1995
Description of amendment request: The proposed amendment would
revise the Operating License and Technical Specifications to allow for
a power uprate to 2900 MWt. The current maximum power level is 2775
MWt.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The probability or consequences of an accident previously
evaluated is not significantly increased.
Implementation of uprate power operation does not contribute to
any accident evaluated in the FSAR [Final Safety Analysis Report].
The NSSS [Nuclear Steam Supply System] Components (RV [reactor
vessel], RCPs [reactor coolant pumps], CRDMs [control rod drive
mechanisms], SGs [steam generators], and piping) are compatible with
the revised operating conditions. These components have been
reanalyzed and the results show that ASME [American Society of
Mechanical Engineers] Code requirements remain satisfied and are
within the current Licensing Basis.
Interfacing Systems which are important to safety are not
adversely impacted and will continue to perform their design
function. Overall secondary plant performance is not significantly
altered by the proposed changes.
The revision to the Pressure Temperature Limits will not
adversely impact the RCS [reactor coolant system] Pressure Boundary.
The length of time these curves will be applicable, due to increased
neutron fluence, is being reduced. Before the 13 Effective Full
Power Years have elapsed, new curves will be generated to reflect
the analysis of the specimen capsule and will be derived utilizing
NRC approved methodology.
Therefore, since the Reactor Coolant pressure boundary integrity
and system functions are not adversely impacted, the probability of
occurrence of an accident evaluated in the VCSNS [Virgil C. Summer
Nuclear Station] FSAR will be no greater than the original design
basis of the plant.
An extensive analysis has been performed to evaluate the
consequences of the following accident types currently evaluated in
the VCSNS FSAR:
[[Page 62496]]
- Non-LOCA [loss-of-coolant accident] Events
- Large Break and Small Break LOCA
- Steam Generator Tube Rupture
With the [delta]75 SGs and revised operating conditions, the
calculated results (i.e., DNBR [departure from nucleate boiling
ratio], Primary and Secondary System Pressure, Peak Clad
Temperature, Metal Water Reaction, Challenge to Long Term Cooling,
Environmental Conditions Inside and Outside containment, etc.) for
the accidents are similar to those currently reported in the VCSNS
FSAR and remain within applicable Regulatory Acceptance Criteria.
Select results (i.e., Containment Pressure during a Steam Line
Break, Minimum DNBR for Rod Withdrawal from Subcritical, etc.) are
slightly more limiting than those currently reported in the FSAR due
to the use of the assumed operating conditions with the [delta]75
SGs and in some cases, use of an uprated core power of 2900 MWt.
However, in all cases, the calculated results do not challenge the
integrity of the primary/secondary/ containment pressure boundary
and remain within the regulatory acceptance criteria applied to
VCSNS's current licensing basis.
Given that calculated radiological consequences are not
significantly higher than current FSAR results and remain well
within 10 CFR 100 limits, it is concluded that the consequences of
an accident previously evaluated in the FSAR are not significantly
increased.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Uprate power operation will not introduce any new accident
initiator mechanisms. Structural integrity of the RCS is maintained
during all plant conditions through compliance with the ASME code
and 10 CFR 50 Appendix G requirements. Design requirements of
auxiliary systems are met with the RSGs [replacement steam
generators] and uprate power operation. No new failure modes or
limiting single failures have been identified. Since the safety and
design requirements continue to be met and the integrity of the
reactor coolant system pressure boundary is not challenged, no new
accident scenarios have been created. Therefore, the types of
accidents defined in the FSAR continue to represent the credible
spectrum of events to be analyzed which determine safe plant
operation.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
Although uprate power operation will require changes to the
VCSNS Technical Specifications, the proposed changes are supported
by extensive LOCA, NON-LOCA and SGTR [steam generator tube rupture]
analyses. These analyses show acceptable consequences with margin to
the applicable regulatory limits. All equipment required to function
during accident conditions has been shown to remain qualified and
thus will perform their design function, and all components remain
in compliance with the codes and standards in effect when VCSNS was
originally licensed (with the exception of the replacement steam
generators which use the 1986 ASME Code Section III Edition).
Low Temperature Overpressure transients which could challenge
RCS structural integrity are not impacted by the revision to the
Pressure Temperature Limitations Curves. The curves are not directly
impacted, the changes do not reduce any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: Frederick J. Hebdon
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: August 29, 1995
Description of amendment request: The proposed amendment would
revise the Technical Specifications for allowable values and trip
setpoints for selected plant process instrumentation. The new allowable
values/setpoints are in accordance with the instrument setpoint
methodology accepted by the NRC staff in a letter dated July 18, 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 change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed revised Trip Setpoints and Allowable Values are
more conservative than those currently approved in the Technical
Specifications. Therefore, any proposed system or component
actuations will occur earlier, resulting in a more conservative
plant response. Thus, 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 change to the Technical Specifications does not
introduce any new components nor does it modify the design of any
existing components. Other than making Trip Setpoints and Allowable
Values of existing instrumentation more conservative, the change
does not affect the design or function of any plant system,
structure, or component, nor does it change the way plant systems
are operated. Thus, the possibility of a new or different kind of
accident previously evaluated is not created.
3. The proposed change does not result in a significant
reduction in the margin of safety.
Since the proposed revised Trip Setpoints and Allowable Values
are more conservative than the existing values, the margin of safety
would be increased by issuance of the changes. Thus, the proposed
change does not result in a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: November 2, 1995
Description of amendment request: The proposed amendment would
revise the Technical Specifications to allow 120 volt AC buses EV-1-A
and EV-1-B to be energized from either their normal inverter power
supply or from their alternate power supply.
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:
These buses are not used as the initiator of any analyzed
accidents. Therefore, the probability of any previously evaluated
accident has not increased. If an accident were to occur while the
buses are supplied from the alternate power supply, there would
[[Page 62497]]
be no change in the analyzed accident scenario since even in the event
of a loss of offsite power event, the safety functions would be
completed. Thus, the consequences of any previously evaluated
accident have not increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated:
The proposed change introduces no new mode of plant operation
and it does not involve physical modification to the plant.
Therefore, it 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:
This change does not involve a significant reduction in a margin
of safety since the proposed change maintains a safety related,
diesel-backed power supply to these buses whether the power is
supplied from the inverters or from the alternate power supply. If a
loss of offsite power event were to occur while the buses were
supplied from the alternate power source, the safety functions being
performed by components supplied from these buses would occur. Thus,
there has been no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: November 2, 1995
Description of amendment request: The proposed amendment to the
Perry Nuclear Power Plant Technical Specifications revises those
specifications associated with handling irradiated fuel in Primary
Containment and the Fuel Handling Building, and selected specifications
associated with CORE ALTERATIONS. Specifically, analysis identifies
that only recently irradiated fuel contains sufficient fission
products to require OPERABILITY of accident mitigation features to meet
the accident analysis assumptions. Analyses also show that accident
mitigation features such as building INTEGRITY and engineered safety
feature (ESF) ventilation systems are not required for CORE ALTERATION
events.
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 requirements are imposed during specific
activities which can be postulated to result in significant
radioactive releases. The proposed APPLICABILITY requirements are
consistent with either the original design basis analyses or with
revised analyses performed to support this proposed amendment.
Because the equipment controlled by the revised Specifications is
not considered an initiator to any previously analyzed accident,
inoperability of the equipment cannot increase the probability of
any previously evaluated accident.
Consistent with the original design basis analysis, the reanalysis
concludes that radiological consequences of the fuel handling accident
are well within the 10 CFR 100.11 limits, as defined by acceptance
criteria in Standard Review Plan Section 15.7.4. The reanalysis has
previously been submitted to the Nuclear Regulatory Commission for
review, and NRC confirmatory calculations reached consistent results
(reference NRC Safety Evaluation for License Amendment No. 35). The
results of the CORE ALTERATION events other than the fuel handling
accident remain unchanged from the original design basis, which showed
that these events do not result in fuel cladding integrity damage or
radioactive releases. Therefore, the proposed changes do not
significantly increase the consequences of any previously evaluated
accident.
Based on the above, the proposed changes do not significantly
increase the probability or 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 requirements are imposed when specific activities
represent situations where significant radioactive releases can be
postulated. The proposed APPLICABILITY requirements are consistent
with design basis analyses. The proposed changes do not introduce
any new modes of plant operation and do not involve physical
modifications to the plant. Therefore, the proposed changes do not
create the possibility of a new or different kind of accidident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change imposes controls to ensure that during
performance of activities which represent situations where
radioactive releases are postulated, the radiological consequences
are at or below the established licensing limit. Safety margins and
analytical conservatisms have been evaluated and are well
understood. Substantial conservatism is retained to ensure that the
analysis adequately bounds all postulated event scenarios. The
current margin of safety is retained.
Specifically, the margin of safety for the fuel handling
accident is the difference between the 10 CFR 100 limits and the
licensing limit defined by the Standard Review Plan (NUREG 0800),
Section 15.7.4. The licensing limit is defined by the Standard
Review Plan as being well within the 10 CFR 100 limits, with
``well within'' defined as 25% of the 10 CFR 100 limits for the fuel
handling accident. Excess margin is the difference between the
postulated doses and the corresponding licensing limit. In the NRCs
initial licensing review of the Perry Nuclear Power Plant (NUREG-
0887, Section 15.3.3), the NRC accepted the design and analyses
based on the results of the analyses being well within the guideline
values of 10 CFR 100.
The proposed APPLICABILITY requirements continue to ensure that
the whole-body and thyroid doses at the exclusion area and low
population zone boundaries as well as control room doses are at or
below the corresponding licensing limit. The margin of safety is
unchanged; therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
The margin of safety for the CORE ALTERATION events other than
the fuel handling accident discussed above also remains the same as
in the original design basis analyses, since the proposed changes do
not impact on the Technical Specification requirements for systems
needed to prevent or mitigate such CORE ALTERATION events.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
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
[[Page 62498]]
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.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of application for amendments: October 6, 1995, and
supplemented November 20, 1995
Brief description of amendments: The amendments revise the
Technical Specifications by incorporating a new acceptance criterion
for steam generator tubes with degradation in the tubesheet roll
expansion region.
Date of issuance: November 21, 1995
Effective date: November 21, 1995
Amendment Nos.: 172 and 159
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 16, 1995 (60 FR
53648) The supplemental letter provided clarifying information that did
not affect the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 21, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of application for amendment: August 10, 1995
Brief description of amendment: The amendment revises the Haddam
Neck Technical Specification Section 3/4.4.3, ``Pressurizer,'' to add a
footnote to allow the pressurizer level to be controlled, outside of
the programmed level, between 25 to 50 percent, plus or minus 5 percent
in Mode 3 when the reactor coolant system is borated to the required
Mode 5 concentrations.
Date of Issuance: November 14, 1995
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 186
Facility Operating License No. DPR-61. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52928) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated November 14, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of application for amendments: September 13, 1995, as
supplemented October 16,1995
Brief description of amendments: These amendments revise the
Administrative Controls section of the BVPS-1 and BVPS-2 TSs to make
them consistent with the requirements of the Offsite Dose Calculation
Manual (ODCM). The ODCM was recently updated to reflect the radioactive
liquid and gaseous effluent release limits and the liquid holdup tank
activity limit of BVPS-1 License Amendment No. 188 and BVPS-2 License
Amendment No. 70 which were issued June 12, 1995.
Date of issuance: November 21, 1995
Effective date: As of the date of issuance, to be implemented
within 10 days.
Amendment Nos.: 194 and 77
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 22, 1995 (60
FR 49292) The October 16, 1995, letter did not change the initial
proposed no significant hazards consideration determination or expand
the amendment request beyond the scope of the September 22, 1995,
Federal Register notice. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated November 21, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 14, 1994, as supplemented by
letters dated July 25, August 15, and August 29, 1995
Brief description of amendment: The amendment changes the Appendix
A Technical Specifications (TSs) to make them consistent with the
revised 10 CFR Part 20, Standards for Protection Against Radiation.
Date of issuance: November 17, 1995
Effective date: November 17, 1995
Amendment No.: 116
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14888) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 17, 1995. The July 25,
August 15, and August 29, 1995 letters provided clarifying information
that did not change the initial propose no significance hazards
consideration determination.
No significant hazards consideration comments received: No
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of application for amendments: May 12, 1995, as supplemented
by letters dated July 6 and October 2, 1995.
[[Page 62499]]
Brief description of amendments: The amendments revise Technical
Specification Surveillance Requirement 4.6.1.2 to add the provision
that 10 CFR Part 50, Appendix J, applies, except as modified by NRC-
approved exemptions.
Date of issuance: November 17, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 91 and 69
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35078) The July 6 and October 2, 1995, letters provided clarifying
information that did not change the scope of the May 12, 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 November 17, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: July 28, 1995, as supplemented
September 12, October 18, and October 31, 1995.
Brief description of amendment: In order to support a full-core
offload as a normal end-of-cycle event, the amendment adds License
Condition 2.C(6) and will require that: (1) the reactor be subcritical
for at least 100 hours prior to the start of reactor refueling
operations, (2) the spent fuel pool bulk temperature be maintained less
than or equal to 140F, and (3) two trains of shutdown cooling be
operable during reactor refueling operations.
Date of issuance: November 9, 1995
Effective date: As of the date of issuance.
Amendment No.: 89
Facility Operating License No. DPR-21. Amendment revised the
license.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45180) The September 12, October 18, and October 31, 1995, submittals
provided additional information that did not change the initial
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendment and Final No
Significant Hazards Consideration Determination are contained in a
Safety Evaluation dated November 9, 1995.
No significant hazards consideration comments received: No public
comments received. A request for a hearing was received from We the
People, the Seacoast Anti-Pollution League, the New England Coalition
on Nuclear Pollution, and Donald Del Core of Uncasville, Connecticut.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: October 6, 1995, supplemented
October 23, November 2, and November 15, 1995.
Brief description of amendment: The amendment adds footnotes to
Action Statement (AS) 3.8.1.1.a of the Technical Specification (TS) and
its bases to allow a one-time extension of the allowed outage time
(AOT) for an inoperable offsite power source from the current 72 hours
to 7 days.
Date of issuance: November 22, 1995
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 192
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 17, 1995 (60 FR
53812). The October 23, November 2, and November 15, 1995, letters
provided clarifying information and slight modifications to the
original request that were not outside the scope of the original notice
and 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 November 22, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360
Northern States Power Company, Docket No. 50-282, Prairie Island
Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota
Date of application for amendment: January 10, 1995, as
supplemented August 9 and September 20, 1995.
Brief description of amendment: The amendments revise the Prairie
Island event monitoring instrumentation Technical Specifications and
associated Bases to conform to Standard Technical Specifications for
post-accident monitoring.
Date of issuance: November 9, 1995
Effective date: November 9, 1995, with full implementation within
30 days.
Amendment Nos.: 121/114
Facility Operating License No. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8753) The August 9 and September 20, 1995, letters provided updated
Technical Specification pages and clarifying information in response to
discussions with the staff during various teleconferences conducted
during the review process. This information was within the scope of the
original application and did not change the staff's initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated November 9, 1995.
No Significant hazards consideration comments received: No
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: December 2, 1994, as
supplemented May 12, 1995.
Brief description of amendments: These amendments relocate the fire
protection requirements from the Technical Specifications to the
Updated Final Safety Analysis Report in accordance with the guidance in
Generic Letter (GL) 86-10, ``Implementation of Fire Protection
Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements
from Technical Specifications.''
Date of issuance: November 20, 1995 Effective date: As of date of
issuance, both units, to be implemented within 30 days.
Amendment Nos.: 104 and 68
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications and the License.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20524) The supplemental letter provided clarifying information and did
not
[[Page 62500]]
change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 20, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: September 14, 1995 and
supplemented by letter dated October 27, 1995
Brief description of amendments: These amendments revise the
technical specifications by deleting Reactor Enclosure and Refueling
Area Secondary Containment Isolation Valve Tables 3.6.5.2.1-1 and
3.6.5.2.2-1, and references to them, in accordance with Generic Letter
91-08, ``Removal of Component lists from Technical Specifications.''
The TS have been modified to state requirements in general terms that
include the components listed in the tables removed from the TS.
Date of issuance: November 20, 1995
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment Nos.: November 20, 1995
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52934) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 20, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464
Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear
Plant, Unit 3, Limestone County, Alabama
Date of application for amendments: October 4, 1995 (TS 368)
Brief description of amendment: The amendment delete requirements
for daily checks for certain instruments that do not have indications,
and provides editorial changes.
Date of issuance: November 13, 1995
Effective Date: November 13, 1995
Amendment No.: 202
Facility Operating License No. DPR-68: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52935) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 13, 1995.
No significant hazards consideration comments received: None
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 35611
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 7, 1995 (TS 95-03)
Brief description of amendments: The amendments address operation
with a rod urgent failure condition, including limited operation with
one control or shutdown bank inserted up to 18 steps below its
insertion point. In addition, the surveillance interval for rod
movement verifications has been increased from 31 to 92 days.
Date of issuance: November 21, 1995
Effective date: November 21, 1995
Amendment Nos.: 215 and 205
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45186) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 21, 1995.
No significant hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: April 28, 1995
Brief description of amendment: The amendment removes the license
conditions for the Transamerica Delaval, Inc. emergency diesel
generators specified by paragraph 2.C.(9) and defined in Attachment 2
to the Operating License.
Date of issuance: November 16, 1995
Effective date: November 16, 1995
Amendment No.: 74
Facility Operating License No. NPF-58: This amendment revises the
license.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29889) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 16, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: June 23, 1995, and facsimile
transmission dated October 31, 1995
Brief description of amendment: This amendment relocates TS 3/
4.3.3.3, ``Seismic Instrumentation;'' TS 3/4.3.3.4, ``Meteorological
Instrumentation;'' and TS 3/4.4.11, ``Reactor Coolant System Vents;''
and the Bases for each of the three sections from the TS to the Updated
Safety Analysis Report, and eliminates the special reporting
requirements for inoperable seismic and meteorological monitoring
instrumentation from TS 6.9.2.
Date of issuance: November 14, 1995 Effective date: November 14,
1995, and shall be implemented not later than 90 days after issuance.
Amendment No.: 201
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39455) The October 31, 1995, facsimile transmission was clarifying in
nature and did not affect the initial no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated November 14, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: June 7, 1995
Brief description of amendment: This amendment revises Technical
Specification 3/4.9.4, Refueling Operations - Containment Penetrations;
[[Page 62501]]
Bases 3/4.9.4, Containment Penetrations; and Limiting Condition for
Operation (LCO) 3.9.4.b to allow both doors of the containment
personnel airlock to be open during core alterations or movement of
irradiated fuel within the containment, provided that certain specified
conditions are meet. Additional changes revise or clarify TS LCO
3.9.4.c, TS Action 3.9.4.a, and TS Surveillance Requirement 4.9.4, and
modify the associated Bases.
Date of issuance: November 17, 1995
Effective date: November 17, 1995
Amendment No.: 202
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39454) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 17, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: December 6, 1994
Brief description of amendments: These changes revise Technical
Specifications to allow appropriate remedial action for high
particulate levels in the diesel generator fuel oil inventory and other
out-of-limit properties in new diesel generator fuel oil that has been
added to the existing diesel generator fuel oil storage inventory.
Date of issuance: November 17, 1995
Effective date: November 17, 1995
Amendment Nos.: Unit 1 - Amendment No. 43; Unit 2 - Amendment No.
29
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6311) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 17, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
Union Electric Company, Docket No. 50-483, Callaway Plant, Callaway
County, Missouri
Date of amendment request: January 13, 1995
Brief description of amendment: The amendment revises Technical
Specifications (TS) 3.3.1 and 3.3.2 to relocate Tables 3.3-2 and 3.3-5,
which provide the response time limits for the reactor trip system and
the engineered safety features actuation system instruments, from the
TS to the updated Final Safety Analysis Report (FSAR). The amendment
also relocates the Bases discussion for TS 3.3.1 and TS 3.3.2 to
Section 16.3 of the updated FSAR.
Date of issuance: November 22, 1995
Effective date: November 22, 1995, to be implemented within 30 days
of issuance.
Amendment No.: 104
Facility Operating License No. NPF-30. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8741) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 22, 1995.
No significant hazards consideration comments received: No
Local Public Document Room locations: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: June 6, 1995
Brief description of amendment: The amendment modifies the Index of
the WNP-2 Technical Specifications by deleting reference to the Bases
pages.
Date of issuance: November 24, 1995
Effective date: November 24, 1995
Amendment No.: 143
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37102) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 24, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: September 13, 1995, and October
19, 1995, as supplemented by letter dated October 25, 1995
Brief description of amendments: These amendments revise Technical
Specification (TS) Section 15.1, ``Definitions,'' TS Section 15.3.1.G,
``Operational Limitations'' (and basis), and TS Figure 15.2.1-2,
``Reactor Core Safety Limits, Point Beach Unit 2.'' The changes reduce
the reactor coolant system raw measured total flow rate limit and
reflect new reactor core safety limits for Unit 2.
Date of issuance: November 17, 1995
Effective date: November 17, 1995
Amendment Nos.: 165 and 169
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications. Public comments requested as to
proposed no significant hazards consideration: Yes (60 FR 54527 dated
October 24, 1995). That notice provided an opportunity to submit
comments on the Commission's proposed no significant hazards
consideration determination. No comments have been received. The notice
also provided for an opportunity to request a hearing by November 24,
1995, 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 is contained in a
Safety Evaluation dated November 17, 1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: September 14, 1995
Brief description of amendment: The amendment revised Technical
Specification 3/4.5.5 to increase the allowed outage time for
adjustment of boron concentration for the refueling water storage tank
from 1 hour to 8 hours.
Date of issuance: November 13, 1995
[[Page 62502]]
Effective date: November 13, 1995, to be implemented within 30 days
of issuance.
Amendment No.: 91
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 11, 1995 (60 FR
52936) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 13, 1995.
No significant hazards consideration comments received: No
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (exigent public announcement or emergency
circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By January 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.
[[Page 62503]]
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, 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, 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).
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: November 9, 1995, as
supplemented by letters dated November 13, 1995, and November 16, 1995
Brief description of amendments: These amendments revise Technical
Specification (TS) Section 15.4.2, ``In-Service Inspection of Safety
Class Components,'' to incorporate a new steam generator tube
acceptance criterion for the Unit 2 steam generators. This criterion
allows tubes that are degraded or defective in a location (within the
tubesheet) that does not affect the structural integrity of the tube to
remain in service. The applicable basis is also changed.
Date of issuance: November 22, 1995
Effective date: November 22, 1995
Amendment Nos.: 166 and 170
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
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 November 22,
1995.
No significant hazards consideration comments received: No
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Ernest L. Blake, Jr., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: Gail H. Marcus
Dated at Rockville, Maryland, this 29th day of November 1995.
For the Nuclear Regulatory Commission
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects - III/IV, Office of
Nuclear Reactor Regulation
[Doc. 95-29540 Filed 12-5-95; 8:45 am]
BILLING CODE 7590-01-F
