[Federal Register Volume 65, Number 8 (Wednesday, January 12, 2000)]
[Notices]
[Pages 1918-1934]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-611]
-----------------------------------------------------------------------
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 December 13, 1999, through December 31,
1999. The last biweekly notice was published on December 29, 1999 (64
FR 73083).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to
[[Page 1919]]
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 February 11, 2000, 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 electronically from
the ADAMS Public Library component on the NRC Web site, http://
www.nrc.gov (the Electronic Reading Room). If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800)-342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: October 25, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specification allowable values for the reactor
protection system electric power monitoring assembly overvoltage and
undervoltage trip setpoints.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or
[[Page 1920]]
consequences of any accident previously evaluated.
The proposed Technical Specification (TS) change revises the
Reactor Protection System (RPS) Electric Power Monitoring Assembly
overvoltage and undervoltage Allowable Values. The new Allowable
Values and setpoints will continue to provide adequate margin to the
normal operating voltage range for the RPS and MSIV [main steam
isolation valve] solenoids, thus minimizing the potential for
inadvertent trips. The proposed change does not have a detrimental
impact on the condition or performance of any plant structure,
system, or component that may initiate an analyzed event. The
proposed change does not physically impact the plant nor does it
impact any design or functional requirements of the associated
system. That is, the proposed change does not degrade the
performance or increase the challenges of any safety systems assumed
to function in the accident analysis. Further, the proposed change
does not impact the Surveillance Requirements themselves nor the way
in which the Surveillances are performed. Consequently, the
probability of an accident previously evaluated is not significantly
increased.
Additionally, the proposed change does not effect the affect the
availability of equipment or systems required for mitigating the
consequences of an accident. The revision of the overvoltage and
undervoltage setpoints will ensure that the associated trip
functions continue to protect the RPS scram solenoids and main steam
isolation valve (MSIV) solenoids so that these devices will perform
their intended safety function. Thus, the affected equipment is
still required to be maintained Operable and capable of performing
the accident mitigation functions assumed in the accident analysis.
As a result, the consequences of any accident previously evaluated
are not significantly affected.
Therefore, based on the above, this change does not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed TS change revises the Reactor Protection System
(RPS) Electric Power Monitoring Assembly overvoltage and
undervoltage Allowable Values. The proposed change does not involve
a physical alteration of the plant (no new or different type of
equipment will be installed) or a change in the methods governing
normal plant operation. The revised setpoints will continue to
ensure that the RPS bus would be disconnected from its power supply
under specified conditions that could damage the RPS bus powered
equipment. Thus, this change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The proposed TS change revises the Reactor Protection System
(RPS) Electric Power Monitoring Assembly overvoltage and
undervoltage Allowable Values. The proposed change provides
necessary conservatism in the Allowable Values in the RPS
Surveillance Requirement to ensure that the equipment used to meet
the Limiting Condition for Operation (i.e., each of the two electric
power monitoring assemblies) can continue to perform its required
functions. At the same time, the revised setpoint/Allowable Values
continue to provide adequate margin to the expected operating
voltage range to prevent inadvertent or unnecessary tripping of the
electric power monitoring assemblies (thus preventing unnecessary or
excessive transfer to the alternate power source). The affected
equipment will thus continue to be tested (calibrated and
functionally tested) in a manner that gives confidence that the
equipment can perform its assumed safety function. Therefore, this
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.
Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius
LLP, 1800 M Street, NW, Washington, DC 20036.
NRC Section Chief: Anthony J. Mendiola.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: October 25, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specification definitions for channel
calibrations, channel functional tests, and logic system functional
tests in accordance with Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-205, Revision 3,
``Revision of Channel Calibration, Channel Functional Test, and Related
Definitions.''
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 any accident previously
evaluated.
The proposed change clarifies the Technical Specification
requirements for performance of channel calibrations, channel
functional test, and logic system functional tests. Specifically,
the proposed change incorporates the NRC-approved Technical
Specification Task Force (TSTF) Standard Technical Specification
Change Traveler, TSTF-205, Revision 3, ``Revision of Channel
Calibration, Channel Functional Test, and Related Definitions.'' The
change approved per this TSTF is not expected to adversely affect
the performance and effectiveness of required testing as testing
appropriate to the associated Surveillance Requirements will
continue to be performed. The proposed change does not have a
detrimental impact on the condition or performance of any plant
structure, system, or component that initiates an analyzed event.
Consequently, the probability of an accident previously evaluated is
not significantly increased. The equipment being tested is still
required to be operable and capable of performing the accident
mitigation functions assumed in the accident analysis. As a result,
the consequences of any accident previously evaluated are not
significantly affected. Therefore, this change does not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The scope of the proposed change is limited to the clarification
of existing test requirements. As such, the proposed change does not
involve a physical alteration of the plant (no new or different type
of equipment will be installed) or a change in 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.
3. The proposed change will not involve a significant reduction
in the margin of safety.
As noted above, the proposed change clarifies requirements for
the performance of channel calibrations, channel functional tests,
and logic system functional tests. Specifically, the proposed change
incorporates the NRC-approved Technical Specification Task Force
(TSTF) Standard Technical Specification Change Traveler, TSTF-205,
Revision 3, ``Revision of Channel Calibration, Channel Functional
Test, and Related Definitions.'' No changes or setpoints to plant
process limits are involved. The surveillance requirements as
revised will continue to ensure that affected equipment is tested in
a manner that gives confidence that the equipment can perform its
appropriate safety function. Therefore, this 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.
Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius
LLP, 1800 M Street, NW, Washington, DC 20036.
NRC Section Chief: Anthony J. Mendiola.
[[Page 1921]]
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: December 16, 1999.
Description of amendment request: The proposed amendment would
allow a one-time extension of some Technical Specification surveillance
intervals to support elimination of a planned spring 2000 mid-cycle
outage (PO-8). For the applicable surveillances, the licensee proposes
to extend their current surveillance intervals to November 30, 2000,
the scheduled startup date from refueling outage 7 (RF-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:
1. The proposed changes do not involve a significant increase in
the probability or consequences of any accident previously
evaluated.
The proposed Technical Specification (TS) changes involve a one-
time only change in the surveillance test intervals of selected
Surveillance Requirements (SRs). The proposed TS changes do not
impact the TS surveillance performance requirements themselves nor
the way in which the surveillances are performed. The proposed TS
changes do not physically involve any changes to the plant, nor do
they impact any design or functional requirements of the associated
systems. Thus, the proposed TS changes do not increase the
challenges of any safety systems assumed to function in the accident
analysis.
In addition, the proposed TS changes do not significantly affect
the availability of equipment or systems required to mitigate the
consequences of an accident because (1) extension of the test
intervals to the extent requested is not expected to have a
significant impact on availability (i.e., no extended test interval
would exceed 30 months), and (2) other or more frequent testing
performed for the affected systems or components, as well as for
redundant systems or components, supports continued availability of
the affected function. The equipment subject to testing per the
affected SRs is still required to be operable and capable of
performing any accident mitigation functions assumed in the accident
analysis. Furthermore, a historical review of surveillance test
results identified no failures that would invalidate these
conclusions.
Based on the above, the proposed TS changes do not significantly
increase the probability or consequences of an accident previously
evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed TS changes involve a one-time only change in the
surveillance testing intervals of selected SRs. Such changes do not
introduce any failure mechanisms of a different type than those
previously evaluated since there are no physical changes being made
to the facility. In addition, the surveillance test requirements
themselves, and the way surveillance tests are performed, will
remain unchanged. Therefore, the proposed TS changes do not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The one-time extended surveillance frequencies do not result in
a significant reduction in the margin of safety. Although the
proposed TS changes will result in an increase in the interval
between surveillance tests, the impact, if any, on system
availability is small. This is because, as noted previously,
extension of the test intervals to the limited extent proposed would
not be expected to have a significant impact on availability. Other
or more frequent testing performed for the affected systems or
components, as well as the testing performed for redundant systems
or components, supports continued availability of the affected
functions.
In addition, the proposed changes do not involve any physical
changes to the affected systems or components, nor do they involve
any changes to setpoints, operating limits, or safety limits.
Based on the above, the assumptions in the licensing basis are
not impacted, and the proposed TS changes do not significantly
reduce 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.
Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius
LLP, 1800 M Street, NW, Washington, DC 20036.
NRC Section Chief: Anthony J. Mendiola.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2 (ANO-1&2), Pope County, Arkansas
Date of amendment request: September 17, 1999.
Description of amendment request: The proposed change to the
Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2), Technical
Specifications would lower the maximum limit for contents of the
gaseous radioactive system from 300,000 curies (Ci) to 78,782 Ci and
82,400 Ci for ANO-1 and ANO-2, respectively. This limit would ensure
that, upon an uncontrolled release of the tank's contents over a 2-hour
period, the resulting total whole body exposure to a member of the
public at the nearest exclusion area boundary would not exceed 0.5
roentgen equivalent man (rem).
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:
An evaluation of the proposed change has been performed in
accordance with 10CFR50.91(a)(1) regarding no significant hazards
considerations using the standards in 10CFR50.92(c). A discussion of
these standards as they relate to this amendment request follows:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated
The proposed change to lower the current technical specification
(TS) gas storage tank activity limits does not require new hardware
or physical equipment modifications to the plant design. By lowering
the setpoint, the resultant exposure at the exclusion area boundary
upon an inadvertent release of a gas storage tank's content will be
limited to 0.5 rem. Therefore the consequences of such an
uncontrolled release of activity are effectively reduced.
Additionally, no new accident is introduced by the proposed
reduction in activity limits associated with the gas storage tanks.
Therefore, reducing the gas storage tank limits from 300,000
Curies (Ci) to 78,782 Ci and 82,400 Ci (ANO-1 and ANO-2,
respectively) does not involve a significant increase in the
probability or consequences of any accident previously evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident From any Previously Evaluated
The proposed change affects the consequences of an event
associated with the loss of gas storage tank radioactive contents on
either ANO-1 or ANO-2. Since this event has been previously
evaluated, no new or different accident can be associated with the
proposed change. Decreasing the present TS activity limits results
in an exposure at the exclusion area boundary to be limited to 0.5
rem in the event of an inadvertent release of a gas storage tank's
content.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety
The proposed change conservatively lowers the existing TS GRW
[Gaseous Radwaste] System gas storage tank activity limits from
300,000 Ci to 78,782 Ci and 82,400 Ci (ANO-1 and ANO-2
respectively). In doing so, the resultant exposure to a member of
the public at the exclusion area boundary during an inadvertent
release of gas storage tank contents over a two-hour period is
reduced to 0.5 rem or less. The proposed change, therefore, retains
the margin of safety for both ANO-1 and ANO-2.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Therefore, based on the reasoning presented above and the
previous discussion
[[Page 1922]]
of the amendment request, Entergy Operations, Inc. has determined
that the requested change does not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: December 16, 1999.
Description of amendment request: The proposed license amendment
request would revise Fuel Handling Accident (FHA) dose calculations for
3 scenarios documented in the River Bend Station, Unit 1 (RBS), Updated
Safety Analysis Report (USAR). The first is a FHA in the fuel building,
assumed to occur 24 hours post-shutdown. A second FHA analysis was
prepared to support Amendment 35 to RBS Technical Specifications (TS)
which assumed a FHA occurs in the primary containment 80 hours post-
shutdown during Local Leakage Rate Testing (LLRT). A third analysis was
prepared in support of Amendment 85 to the RBS TS which assumed the
containment is open at 11 days.
These analyses are being updated to account for several changes.
The primary reason for the revisions, as stated by the licensee, was to
update the analyses to reflect current RBS operating strategies and
make the analyses consistent with each other. Specifically, Cases 1 and
2 of the three analyses assumed a Radial Peaking Factor (RPF) of 1.5
consistent with Regulatory Guide (RG) 1.25. However, current core
design strategies could lead to an RPF as high as 1.65. In addition, to
account for the potential impact of extended burnup fuel in future
operating cycles, an increased iodine-131 gap fraction of 0.12 was more
conservatively assumed in lieu of the 0.10 recommended by RG 1.25. The
revised analysis also includes a change to the control room atmospheric
dispersion factors (/Q) for the Main Control Room (MCR)
ventilation system. Credit is taken for Standard Review Plan (SRP)
Section 6.4 guidance for manual dual control room air intakes in that
the /Q's are divided by 4. The revised FHA analyses also
credit this action at a 20 minute delay to be consistent with the Loss
of Coolant Accident (LOCA) analysis.
Furthermore, an error was discovered in one of the FHA
calculations. The release rate assumed in the analysis did not ensure
that the RG 1.25 assumption of a 2-hour release was preserved. The
error is the result of an inherent bias in the secondary mixing effects
in the dose calculation. The results continue to be bounded by the
guidance contained in SRP 15.7.4 and RG 1.25.
Reanalysis showed that the release rate error, compounded with the
other changes discussed above, resulted in calculated doses greater
than those currently found in the RBS USAR. In addition, some of the
doses were also greater than those presented in the Amendment 85
submittal. However, the licensee has stated that the results of the
revised analyses remain ``well within'' 10 CFR 100, the guidance
contained in SRP 15.7.4, and RG 1.25. Since the analyses results are
above those reported in the RBS USAR, the criterion of 10 CFR
50.59(a)(2)(i) is, therefore, satisfied. Accordingly, the licensee has
concluded that these changes involve an unreviewed safety question.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated.
The analyses changes described by this proposed change to the
USAR are not initiators to events, and, therefore, do not involve
the probability of an accident. The changes to the FHA calculations
for radiological doses following a FHA reflect the current operating
strategies and make the analyses consistent. These changes included:
Accounting for the impact of extended burnup fuel,
Addressing a change to the control room atmospheric
dispersion factors assumed in the analysis, and
Revising the Radial Peaking Factor (RPF) used in the
analysis. Current core design strategies could lead to a RPF higher
[than] that assumed in Regulatory Guide 1.25.
The TRANSACT code is used for offsite dose and control room dose
calculations. The TRANSACT code is derived from the TACT V code
documented in
NUREG/CR-5109. RBS has benchmarked the TRANSACT code as
discussed in the request dated August 17, 1995, (RBG-41728) which
resulted in the NRC granting Amendment 85.
The revisions to the FHA are used to establish operational
conditions where specific activities represent situations where
significant radioactive releases can be postulated. These
operational conditions include:
Initial fuel movement in the Fuel Building 24 hours
after shutdown,
Fuel movement in Primary Containment after 80 hours
with leakrate testing being conducted, and
Fuel movement in Primary Containment with the Primary
Containment open.
Because the analyses affected by the changes are not considered
an initiator to any previously analyzed accident, these changes
cannot increase the probability of any previously evaluated
accident. Therefore, this change does not increase the probability
of occurrence of an accident evaluated previously in the safety
analysis report (SAR).
This proposed change to the USAR does increase the consequences
of an accident, but the increase is within all regulatory limits and
guidance. While the calculated off-site and control room doses of a
FHA did increase, the dose consequences remain below the regulatory
limits of 10 CFR 100 and 10 CFR 50, Appendix A, GDC [General Design
Criterion]-19 as approved per NUREG-0989, and the guidance contained
in SRP 15.7.4 of less than 25% of the 10 CFR 100 limits. The cause
of these events remains the failure of the fuel assembly lifting
mechanism. These analyses demonstrated that for the worst case
bundle drop, the regulatory dose guidelines of SRP 15.7.4 continue
to be satisfied for the required decay periods.
This change accounts for the potential effects of current fuel
design and operating strategies including increased burnup of fuel,
increased iodine-131 fraction released, Main Control Room
ventilation system operation, and release rate timing assumptions.
Reanalysis of the off-site dose calculation demonstrates that the
revised doses are increased but remain less than the regulatory
limits of 10 CFR 100 and within the guidance of SRP 15.7.4.
Therefore, this change does not significantly increase the
consequences of an accident previously evaluated in the SAR.
The proposed changes, in conjunction with existing
administrative controls, bound the conditions of the current design
basis fuel handling accident analysis. The analysis also concludes
the limiting offsite radiological consequences are well within the
acceptance criteria of NUREG[-]0800, Section 15.7.4 and 10 CFR 50,
Appendix A, GDC[-]19. The analysis is also conducted in a
conservative manner containing margins in the calculation of
mechanical analysis, iodine inventory, and iodine decontamination
factor. Each of these conservatisms will further decrease the
consequences. Therefore, the proposed changes do not significantly
increase the probability or consequences of any previously evaluated
accident.
2. The proposed changes would not create the possibility of a
new or different kind of accident from any previous[ly] analyzed.
This change does not involve initiators to any events in the
SAR, nor does the activity create the possibility for any new
accidents. Rather, this change is a result of the
[[Page 1923]]
evaluation of the most limiting FHA, which can occur at River Bend.
The proposed changes to the dose analyses are consistent with
previous limits, only revising previous evaluations to account for
current operating strategies and assumptions. These changes
included:
Accounting for the impact of extended burnup fuel,
addressing a change to the control room atmospheric
dispersion factors assumed in the analysis, and
Revising the Radial Peaking Factor (RPF) used in the
analysis. Current core design strategies could lead to a RPF higher
[than] that assumed in Regulatory Guide 1.25.
The radiological consequences remain within accepted limits of
10 CFR 100 and guidance of the Standard Review Plan (NUREG-0800)
Section 15.7.4. Therefore, these changes are consistent with the
design basis analysis. 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 accident from any
previous[ly] analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The dose consequences are calculated in accordance with
regulatory guidance found in Regulatory Guide 1.25 and the SRP
[S]ection 15.7.4. The RBS analyses conservatively assumed that
failures are consistent with those in the standard General Electric
GESTAR II. These analyses result in a bounding number of fuel
failures. The RBS analyses are also consistent with those approved
by the NRC [Nuclear Regulatory Commission] in support of Technical
Specification Amendments 35 and 85 to the River Bend Station license
(NPF-47). The radiological dose consequences resulting from these
failures are therefore analyzed using accepted methods and criteria.
In addition, the analyses contain known conservatisms and margins to
ensure the results will remain bounding.
The revised limits are used to establish operational conditions
where specific activities represent situations where significant
radioactive releases can be postulated. These operational conditions
are consistent with the design basis analysis and are established
such that the radiological consequences are at or below the current
regulatory limits and guidance. Safety margins and analytical
conservatisms have been evaluated and are well understood.
Conservative methods of analysis are maintained through the use of
accepted methodology and benchmarking the proposed methods to
previous analysis. Margins are retained to ensure that the analysis
adequately bounds all postulated event scenarios. The proposed
change only eliminates some excess conservatism from the analysis.
In addition, EOI [Entergy Operations, Inc.] has implemented
NUMARC [Nuclear Management and Resources Council (now NEI)] 91-06
guidelines for shutdown operations at RBS. Shutdown Operations
Protection Plan and Primary-Secondary Containment Integrity
procedures presently include guidance for closure of the containment
hatch and other significant openings in containment, in addition to
the requirements contained in the license and design basis. This
additional protection will enhance the ability to limit offsite
effects.
Acceptance limits for the fuel handling accident are provided in
10 CFR 100 with additional guidance provided in NUREG[-]
0800, Section 15.7.4. The proposed changes 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 below the
corresponding regulatory limits. These margins are unchanged,
therefore, the proposed changes do not involve a significant reduction
in a margin of safety.
The commission has provided guidance concerning the application
of the standards of 10 CFR 50.92 by providing certain examples (51
FR 7751, March 6, 1986) of amendments that are not considered likely
to involve a significant hazards consideration. This proposed
amendment is very similar to example (vi):
(vi) A change which either may result in some increase to the
probability or consequences of a previously-analyzed accident or may
reduce in some way a safety margin, but where the results of the
change are clearly within all acceptable criteria with respect to
the system or component specified in the Standard Review Plan: for
example, a change resulting from the application of a small
refinement of a previously used calculational model or design
method.
As we have shown in the preceding discussion, this refinement to
the FHA dose calculation results in a small increase to the
consequences of a previously analyzed accident, but the results of
the change remain clearly within the guidelines of 10 CFR 100,
Appendix A, GDC[-]19, and the guidance of SRP [S]ection 15.7.4,
without reducing 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.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: November 17, 1999 (L-99-241)
Description of amendment request: The proposed amendment would
revise the St. Lucie Unit 1 and Unit 2 Technical Specifications (TS) to
require laboratory testing of activated charcoal samples for applicable
engineered safety feature ventilation systems using the American
Society for Testing and Materials (ASTM) D3803-1989 protocol. The
affected TS are Units 1 and 2 shield building ventilation system, TS
4.6.6.1; Unit 1 emergency core cooling system area ventilation system,
TS 4.7.8.1; Unit 1 control room emergency ventilation system, TS
4.7.7.1; Unit 2 control room emergency air cleanup system, TS 4.7.7;
and Unit 1 fuel pool ventilation system--fuel storage, TS 4.9.12.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment does not involve a significant increase
in the probability or consequences of any accident previously
evaluated. The new charcoal testing protocol is performed offsite on
samples extracted from the safety-related ventilation systems.
Therefore, there is no impact on any accident initiator and
therefore, no changes on the probability. The proposed testing
protocol is more conservative than previous tests; therefore, the
efficiency of charcoal for the affected safety-related systems would
not be overestimated. With the new testing protocol, more
conservative testing results are expected since the temperature at
which testing is performed is lower and the charcoal retention
capability is more consistent with actual accident conditions. The
proposed change thus ensures that the charcoal in service will
comply with the penetration requirements to meet the design basis
accident conditions.
Therefore, operation of the facility in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed new charcoal testing protocol only affects
surveillance testing requirements for ventilation systems. The
functions of these systems remain unchanged and unaffected. No new
system interactions have been introduced by the proposed amendment,
which would create a new or different type of accident than
previously analyzed. No physical changes are being made to any
structure, system or component. The operation of the facility has
not been altered by the proposed amendment. The systems involved are
not considered to initiate any accidents as previously evaluated.
[[Page 1924]]
The proposed amendment will not change the physical plant or the
modes of operation defined in the facility license. The changes do
not involve the addition of new equipment or the modification of
existing equipment, nor do they alter the design of St. Lucie plant
systems. Therefore, operation of the facility in accordance with the
proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment does not involve a reduction in the
margin of safety. The margin of safety of the Technical
Specifications, its bases, the Final Safety Analysis Report, the
Safety Evaluation Report or in any other design document has not
been affected by the proposed amendment. The change provided in this
proposed amendment is related to introducing an improved testing
protocol for the activated charcoal in safety related ventilation
systems. The change consists of testing the charcoal with a new
testing protocol and with lower test temperatures to resemble more
closely accident conditions and to eliminate potential
overestimation of charcoal efficiency.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420
NRC Section Chief: Richard P. Correia
GPU Nuclear, Inc., et al., Three Mile Island Nuclear Station, Unit 2
(TMI-2), Docket No. 50-320, Dauphin County, Pennsylvania
Date of amendment request: November 5, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) Sections 6.8.1.4, 6.5.4.6, 6.13,
6.14, and 6.8.3. Specifically, Sections 6.13, 6.14 and 6.8.3 would be
revised to eliminate the requirement to notify the Nuclear Regulatory
Commission (NRC) of exceeding environmental limits and changes to
environmental permits such as National Pollution Discharge Elimination
System (NPDES). The requirements contained in the individual
environmental permits and program regulations administered by the U.S.
Environment Protection Agency (EPA), Pennsylvania Department of
Environmental Protection (PADEP), and other regulatory agencies with
program jurisdiction for reporting are included in plant procedures and
data base tracking systems. Sections 6.8.1.4 and 6.5.4.6 are changes to
the amendment that are administrative in nature and reflect a
streamlining of the GPU Nuclear, Inc. management structure.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analyses of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes to the TMI-2 [Three Mile Island, Unit 2]
Technical Specifications do not involve a significant increase in
the probability of occurrence or consequences of an accident or
malfunction of equipment important to safety previously analyzed in
the safety analysis report. The changes have no impact on plant
operations or the release of radioactive materials.
2. The proposed changes to the TMI-2 Technical Specifications
will not create the possibility for an accident or malfunction of a
different type than any previously evaluated in the safety analysis
report because no plant configuration or operational changes are
involved.
3. The changes will not involve a significant reduction in the
margin of safety as defined in the basis for any technical
specification for TMI-2 because no change to operational limits will
be made.
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.
Attorney for Licensee: Ernest L. Blake, Jr Esq., Shaw, Pittman,
Potts & Trowbridge, 2300 N. Street, N.W., Washington, DC 20037.
NRC Section Chief: Mike Masnik.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: November 10, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.7.c, to commit to the American
Society for Testing and Materials (ASTM) D3803-1989 test protocol for
the ventilation filter testing program. The proposed changes are
consistent with Attachment 2, Sample Technical Specifications, in
Generic Letter 99-02. Because the current TS penetration limits do not
reflect a safety factor in excess of that assumed in the dose
calculations of the accident analysis, the TS change request would also
revise the allowable penetration values to correspond to a safety
factor of 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below for the administrative changes:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The ESF [engineered safety feature] ventilation systems are not
initiators of any accident previously evaluated and the change in
testing protocol to ASTM D3803-1989 as requested by the NRC will be
more accurate and realistic and provide greater assurance of
consistency. The acceptance criteria will be more conservative than
those currently used in TS 5.5.7.c.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
No new types of accidents are being introduced because no
modifications or changes in operations are being proposed for the
ESF [engineered safety feature] ventilation systems. The proposed
changes to TS 5.5.7.c impact acceptance criteria and test protocols
only.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
The margin of safety is not reduced. The proposed change in ESF
ventilation testing protocol includes a safety factor of two (2) for
the penetration limit in excess of that assumed in the dose
calculations of the DAEC [Duane Arnold Energy Center] accident
analysis.
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.
Attorney for licensee: Al Gutterman, Morgan, Lewis & Bockius, 1800
M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Claudia M. Craig.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: November 22, 1999.
Description of amendment request: The proposed amendment would
adopt selected NRC-approved generic changes to the Improved Technical
Specifications (ITS) NUREGs. The 16 changes come from the Technical
Specification Task Force (TSTF) process
[[Page 1925]]
developed by the industry and the NRC. Three of these changes are
Bases-only changes but are included for completeness relative to the
TSTF process.
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 for the administrative changes:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change involves reformatting, renumbering, and
rewording the existing Technical Specifications. The reformatting,
renumbering, and rewording process involves no technical changes to
the existing Technical Specifications. As such, this change is
administrative in nature and does not affect initiators of analyzed
events or assumed mitigation of accident or transient events.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in methods governing normal plant operation. The proposed
change will not impose any new or eliminate any old requirements.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change will not reduce a margin of safety because
it has no effect on any safety analyses assumptions. This change is
administrative in nature. Therefore, the change does not involve a
significant reduction in a margin of safety.
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 for more restrictive changes:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change provides more stringent requirements for
operation of the facility. These more stringent requirements do not
result in operation that will increase the probability of initiating
an analyzed event and do not alter assumptions relative to
mitigation of an accident or transient event. The more restrictive
requirements continue to ensure process variables, structures,
systems, and components are maintained consistent with the safety
analyses and licensing basis. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in methods governing normal plant operation. The proposed
change does impose different requirements. However, these changes
are consistent with the assumptions in the safety analyses and
licensing basis. Thus, this change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The imposition of more restrictive requirements either has no
effect on or increases the margin of plant safety. As provided in
the justification, each change in this category is, by definition,
providing additional restrictions to enhance plant safety. The
change maintains requirements within the safety analyses and
licensing basis. Therefore, the change does not involve a
significant reduction in a margin of safety.
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 for less restrictive changes--removed detail:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relocates certain details from the Technical
Specifications to other documents under regulatory control. The
Bases, UFSAR [updated final safety analysis report], and Technical
Requirements Manual will be maintained in accordance with 10 CFR
50.59. In addition to 10 CFR 50.59 provisions, the Technical
Specification Bases are subject to the change control provisions in
the Administrative Controls Chapter of the Technical Specification.
The UFSAR is subject to the change control provisions of 10 CFR
50.71(e). Other documents are subject to controls imposed by
Technical Specifications or regulations. Since any changes to these
documents will be evaluated, no significant increase 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.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change will not impose or eliminate any requirements and
adequate control of the information will be maintained. Thus, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change will not reduce a margin of safety because
it has no effect on any safety analyses assumptions. In addition,
the details to be moved from the Technical Specifications to other
documents are the same as the existing Technical Specifications.
Since any future changes to these details will be evaluated, no
significant reduction in a margin of safety will be allowed. A
significant reduction in the margin of safety is not associated with
the elimination of the 10 CFR 50.92 requirement for NRC review and
approval of future changes to the relocated details. The proposed
change is consistent with the BWR [Boiling Water Reactor]/4 Standard
Technical Specifications, NUREG-1433, issued by the NRC Staff,
revising the Technical Specifications to reflect the approved level
of detail, which indicates that there is no significant reduction in
the margin of safety.
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 for less restrictive changes--category 3, relaxation of
completion time:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relaxes the Completion Time for a Required
Action. Required Actions and their associated Completion Times are
not initiating conditions for any accident previously evaluated and
the accident analyses do not assume that required equipment is out
of service prior to the analyzed event. Consequently, the relaxed
Completion Time does not significantly increase the probability of
any accident previously evaluated. The consequences of an analyzed
accident during the relaxed Completion Time are the same as the
consequences during the existing Completion Time. As a result, the
consequences of any accident previously evaluated are not
significantly increased. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
Required Actions and associated Completion Times have been evaluated
to ensure that no new accident initiators are introduced. Thus, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The relaxed Completion Time for a Required Action does not
involve a significant reduction in the margin of safety. As provided
in the justification, the change has been evaluated to ensure that
the allowed Completion Time is consistent with the safe operation
under the specified Condition, considering the operability status of
the
[[Page 1926]]
redundant systems of required features, the capacity and capability
of remaining features, a reasonable time for repairs or replacement
of required features, and the low probability of a DBA [design basis
accident] occurring during the repair period. Therefore, this change
does not involve a significant reduction in a margin of safety.
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 for less restrictive changes--category 4, relaxation of
required action.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relaxes Required Actions. Required Actions
and their associated Completion Times are not initiating conditions
for any accident previously evaluated and the accident analyses do
not assume that required equipment is out of service prior to the
analyzed event. Consequently, the relaxed Required Actions do not
significantly increase the probability of any accident previously
evaluated. The Required Actions in the change have been developed to
provide assurance that appropriate remedial actions are taken in
response to the degraded condition, considering the operability
status of the redundant systems of required features, and the
capacity and capability of remaining features while minimizing the
risk associated with continued operation. As a result, the
consequences of any accident previously evaluated are not
significantly increased. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
Required Actions and associated Completion Times in the change have
been evaluated to ensure that no new accident initiators are
introduced. Thus, this change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The relaxed Required Actions do not involve a significant
reduction in the margin of safety. As provided in the justification,
the change has been evaluated to minimize the risk of continued
operation under the specified Condition, considering the operability
status of the redundant systems of required features, the capacity
and capability of remaining features, a reasonable time for repairs
or replacement of required features, and the low probability of a
DBA [design basis accident] occurring during the repair period.
Therefore, this change does not involve a significant reduction in a
margin of safety.
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 for less restrictive changes--category 6, relaxation of
surveillance requirement acceptance criteria:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relaxes the acceptance criteria of
Surveillance Requirements. Surveillances are not initiators to any
accident previously evaluated. Consequently, the probability of an
accident previously evaluated is not significantly increased. The
equipment being tested is still required to be Operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Therefore, this
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in 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.
3. Does this change involve a significant reduction in a margin
of safety?
The relaxed acceptance criteria for Surveillance Requirements do
not result in a significant reduction in the margin of safety. As
provided in the justification, the relaxed Surveillance Requirement
acceptance criteria have been evaluated to ensure that they are
sufficient to verify that the equipment used to meet the LCO
[limiting condition for operation] can perform its required
functions. Thus, appropriate equipment continues to be tested in a
manner that gives confidence that the equipment can perform its
assumed safety function. Therefore, this 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.
Attorney for licensee: Al Gutterman, Morgan, Lewis & Bockius, 1800
M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Claudia M. Craig.
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: September 3, 1998, as supplemented by
letters dated January 22, February 5, March 17, and November 24, 1999.
The September 3, 1998, amendment application was previously noticed in
the Federal Register on December 16, 1998 (63 FR 69345).
Description of amendment requests: The amendment would revise
Section 5.6.6, ``Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE
LIMITS REPORT (PTLR),'' of the improved Technical Specifications (TSs),
that were issued in Amendment Nos. 135 and 135 on May 28, 1999. The
amendment would add the phrase ``and LTOP'' (low-temperature
overpressure protection) to the first sentence of item 5.6.6.b that
identifies the limits that can be determined by the licensee in the
PTLR, and (2) replace the current list of documents listed in item
5.6.6.b by the NRC letter that would approve this amendment and
Westinghouse WCAP-14040-NP-A, ``Methodology Used to Develop Cold
Overpressure Mitigation System Setpoints and RCS Heatup and Cooldown
Limit Curves,'' dated January 1996. WCAP-14040-NP-A is the NRC-approved
topical report which provides a methodology for developing the LTOP
setpoints and RCS heatup and cooldown limit curves for Westinghouse
plants, such as Diablo Canyon Nuclear Power Plant, Units 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to Figures 3.4-2 and 3.4-3 of Technical
Specification (TS) 3.4.9.1 and the associated Bases adjust the
reactor coolant system (RCS) heatup and cooldown pressure/
temperature (P/T) limits to permit operation through 16 effective
full power years (EFPY). The 16 EFPY P/T limits are more restrictive
than the current limits; this accounts for an expected incremental
increase in reactor vessel embrittlement, and assures the reactors
will continue to be operated within acceptable stresses and at
temperatures for which the reactor vessel metal exhibits ductile
properties. The P/T limits developed for 16 EFPY were determined in
accordance with 10 CFR 50, Appendix G, and maintain the same margins
of safety as the current limits. The proposed changes will not
impact the probability of overpressurization or brittle fracture of
the vessel, and therefore will not impact the consequences of an
accident.
The present low temperature overpressure protection (LTOP)
pressure and enable temperature setpoints were reviewed and found to
be acceptable and conservative for
[[Page 1927]]
use through 16 EFPY, based on use of ASME [American Society of
Mechanical Engineers] Code Case N-514, which provides acceptable
margins to the prevention of vessel overpressurization and brittle
fracture. Therefore, there is no change to the consequences of
accidents previously analyzed. Since no changes are proposed in the
actual LTOP setpoints, nor any physical alteration of the LTOP
system, nor a change to the method by which the LTOP system performs
its function, there would be no change to the probability of an
accident previously evaluated. The proposed change to the Bases
incorporates use of ASME Code Case N-514, which will benefit DCPP
[Diablo Canyon Power Plant] by not resulting in a reduced RCS P/T
window and reduced power-operated relief valve (PORV) pressure
setpoint for LTOP. This maintains the current level of operator
flexibility during heatup and cooldown, and prevents an increase in
the probability of an accident associated with an inadvertent PORV
actuation.
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 to TS 3.4.9.1, ``Reactor Coolant System--
Pressure/Temperature Limits,'' do not involve any physical
alteration to any plant system or change the method by which any
safety-related system performs its function. The changes to TS
3.4.9.1 account for the effects of an incremental increase in
reactor vessel embrittlement and are requested in order to restrict
future reactor operation to within acceptable stress levels and
temperature regimes in accordance with 10 CFR 50, Appendix G,
requirements. These changes are needed to maintain the current P/T
limit margins of safety as defined by 10 CFR 50, Appendix G, and
ASME XI, Appendix G, for operation through 16 EFPY. The possibility
of a new kind of accident such as catastrophic failure of the
reactor vessel is prevented by maintaining acceptable margins of
safety.
The present LTOP pressure setpoint was reviewed and found to be
acceptable and conservative for the extension of the P/T curves to
16 EFPY.
Additionally, the proposed changes will not affect the ability
of the LTOP system to provide pressure relief at low temperatures,
thereby maintaining the LTOP design basis.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes to TS 3.4.9.1 adjust the RCS heatup and
cooldown P/T limits to permit operation through 16 EFPY. The P/T
limits have been determined in accordance with 10 CFR 50, Appendix
G, and include the safety margins with regard to brittle fracture
required by the ASME Section XI, Appendix G, which maintain the same
margins of safety as the current limits.
The LTOP setpoints were reevaluated using the requirements of
ASME Code Case N-514. This code case was developed to provide the
necessary margins of safety for the prevention of reactor vessel
overpressurization and brittle fracture. The LTOP evaluation results
conclude the current LTOP setpoints are conservative for operation
through 16 EFPY. In addition, avoiding an unnecessary reduction in
the LTOP, the PORV pressure setpoint prevents an increase in the
likelihood of an inadvertent PORV actuation
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.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of amendment request: December 1, 1999.
Description of amendment request: The proposed amendment would
revise the Humboldt Bay Power Plant (HBPP) Unit 3 Technical
Specifications (TS) related to fire protection, administrative
controls, and quality assurance audits. The fire protection
requirements would be relocated verbatim from the TS to the HBPP
Defueled Safety Analysis Report (DSAR). The administrative controls
requirements would be revised to (1) refer to the DSAR for a
description of the plant organization, (2) modify information
pertaining to plant staff titles and qualifications to reflect the
current organization, and (3) replace a reference to the Final Hazards
Summary Report with a reference to the DSAR. Quality assurance audit
requirements would be relocated from the TS to the Quality Assurance
Plan.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analyses of the issue of no significant hazards
consideration, which are presented below:
For the proposed changes to the fire protection requirement, the
licensee's analysis states:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The FPP [Fire Protection Program] and FPS [Fire Protection
System] are not being changed. Operability requirements and
procedural controls of the FPP and FPS are not being changed. The
proposed changes involve only where the FPP and FPS description is
located and how changes can be made. Consequently, the changes will
not affect the probability or consequences of an accident occurring.
Future changes to the FPP and FPS as described in the Defueled
Safety Analysis Report would be made in accordance with 10 CFR
50.59. This ensures that adequate controls will remain in place so
that the public health and safety will be protected.
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 FPP and FPS are not being changed. Operability requirements
and procedural controls of the FPP and FPS are not being changed.
The proposed changes involve only where the FPP and FPS description
is located and how changes can be made. Consequently, the changes
will not affect the probability or consequences of an accident
occurring.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The FPP and FPS are not being changed. Operability requirements
and procedural controls of the FPP and FPS are not being changed.
The proposed changes involve only where the FPP and FPS description
is located and how changes can be made. Consequently, the changes
will not affect the probability or consequences of an accident
occurring.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
For the proposed changes to the administrative controls
requirements, the licensee's analysis states:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The organization title and responsibility changes update the
Technical Specification (TS) to reflect the current organization and
have no impact on the function or operability of plant systems,
structures, or components, or the ability of the plant to safely
maintain SAFSTOR status. Consequently, the changes will not affect
the probability or consequences of an accident occurring.
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
[[Page 1928]]
accident from any accident previously evaluated.
The organization title and responsibility changes update the TS
to reflect the current organization and have no impact on the
function or operability of plant systems, structures, or components,
or the ability of the plant to safely maintain SAFSTOR status.
Consequently, the changes will not affect the probability or
consequences of an accident occurring.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The organization title and responsibility changes update the TS
to reflect the current organization and have no impact on the
function or operability of plant systems, structures, or components,
or the ability of the plant to safely maintain SAFSTOR status.
Consequently, the changes will not affect the probability or
consequences of an accident occurring.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
For the proposed changes to the quality assurance audit
requirements, the licensee's analysis states:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes simplify the Technical Specifications (TS),
meet regulatory requirements for relocated TS, and implement: (1)
The recommendations of NRC's letter dated October 25, 1993, from
William T. Russell to the chairpersons of the industry owners
groups; (2) the Commission's Final Policy Statement on TS
Improvements; and (3) the current revision of 10 CFR 50.36. Future
changes to these requirements will be controlled by 10 CFR 50.54.
This ensures that adequate controls will remain in place so that the
public health and safety will be protected. The proposed changes are
administrative in nature and do not involve any modifications to any
plant equipment or 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 are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety-related system performs
its function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not alter implementation of the basic
regulatory requirements and do not affect any safety analyses.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Michael Masnik.
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: October 12, 1999 (TS 99-15).
Brief description of amendments: The proposed amendments would
change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79
(Unit 2) by revising the Technical Specification (TS) to provide for
unisolation of containment penetrations under administrative controls.
This revision will add a footnote to Specification 3.9.4.c indicating
this allowance and the necessary Bases addition for this section to
clarify the use of this allowance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed revision will allow the opening of specific
containment penetrations during the movement of irradiated fuel or
core alterations provided administrative controls are implemented.
These controls will establish the proper awareness of the unisolated
penetration condition, designate individuals to isolate the
penetration in the event of an FHA [fuel handling accident], and
[to] ensure the auxiliary building gas treatment system (ABGTS) is
available. The status of containment penetrations does not impact
the generation of an accident nor does the ability to unisolate
penetrations affect this potential. The proposed revision does not
alter any plant equipment or operating practices other than
penetration isolation such that the probability of an accident is
increased.
The administrative controls provide adequate requirements to
provide timely identification and closure of penetrations opened
under this allowance should a fuel handling event occur. Designated
individuals ensure that adequate resources are available to isolate
the penetration such that the offsite dose consequences are not
significantly impacted. The lack of motive force in containment
during fuel movement to expel the radioactive material allows a more
flexible isolation interval. The exception for the containment
ventilation isolation valves is based on being exposed to a motive
force and the flow paths outside the auxiliary building secondary
containment enclosure (ABSCE) is based on being exposed to an
unfiltered atmosphere. Timely isolation of the specified flow paths
is required to ensure that the unlikely transmission of radioactive
material does not occur. Interactions that may occur during the
period of time before isolation will be controlled by operation of
the ABGTS and will not significantly increase the consequences of an
accident as previously evaluated. Completion of penetration
isolation and operation of the ABGTS, as required by the
administrative controls, will maintain the offsite dose consequences
well within the 10 CFR 100 limits.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed allowance to open penetrations in Mode 6 will not
alter plant functions or equipment operating practices other than
penetration isolation. Containment penetration status is not
considered to be the source of an accident. Therefore, since the
plant functions and equipment are not altered and the isolation
status of containment penetrations do not contribute to the
initiation of postulated accidents, the proposed revision will not
create a new or different kind of accident.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The isolation requirements for containment penetrations ensure
that the release of radioactivity is minimized to maintain the 10
CFR 100 limits for offsite dose consequences in the event of an FHA.
The proposed change to allow penetrations to be unisolated does not
significantly affect the expected dose consequence because of the
absence of containment pressurization potential during fuel movement
or core alterations. The most significant offsite dose contributor
to the fuel handling event is the containment purge system that
generates a motive force for the radioactive material. This flow
path is excluded from the proposed allowance because of this motive
force potential along with flow paths outside the ABSCE. Without
this motive force, as is the case with other penetrations during
fuel movement or core alterations, the potential for additional
offsite dose consequence is unlikely. As an additional measure, this
allowance applies to flow paths that can be filtered by the ABGTS.
Therefore, the margin of safety provided by the containment building
penetration requirements is not significantly impacted by the
proposed allowance to open penetrations under administrative
controls. With the timely provision to identify and isolate affected
penetrations and the provision for ABGTS operability, the margin of
safety is maintained without a significant reduction.
[[Page 1929]]
The NRC 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.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
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: November 24, 1999 (TS 99-16).
Brief description of amendments: The proposed amendments would
change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79
(Unit 2) by updating the Technical Specification (TS) surveillance
requirements for penetration efficiency tests of charcoal adsorbers to
comply with American Society for Testing and Materials (ASTM) test
standard ASTM D3803-1989 as directed by NRC Generic Letter (GL) 99-02.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed revision will require laboratory tests of safety-
related charcoal filter adsorbers to tighter specifications. NRC
research indicates that the new test protocols yield more accurate
measures of filter efficiency and better reproducibility of test
results. No physical change is made to the filter by these expanded
timeframes of testing and tighter controls; therefore, no change to
the filter behavior is expected. Current methods for selecting and
obtaining charcoal samples for testing will be retained without
change. The proposed revision does not alter any plant equipment or
operating practices other than filter tests that are conducted away
from the plant site, and as such the probability of an accident is
not increased.
Laboratory test acceptance criteria contain a safety factor to
ensure that the efficiency assumed in the accident analysis is still
valid at the end of the operating cycle. Because ASTM D3803-1989 is
a more accurate and demanding test than older tests, upgrading TSs
to the ASTM D3803-1989 protocol allows use of a safety factor of 2
for determining the acceptance criteria for charcoal filter
efficiency. This safety factor can be used for systems with or
without humidity control because the lack of humidity control is
already accounted for in the test conditions.
Applying the ASTM D3803-1989 test methodology and using the new
safety factor is expected to yield a net improvement in safety. The
ASTM D3803-1989 test protocol is expected to improve the
identification of degraded charcoal filters and lead to their timely
replacement without any adverse effects on filter performance.
Therefore, the change in testing does not significantly increase the
consequences of an accident.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change in laboratory tests performed on charcoal
filters will not alter plant functions or equipment operating
practices other than possibly resulting in more frequent replacement
of charcoal filters. As stated previously, current methods for
selecting and obtaining charcoal samples for testing will be
retained without change. The ASTM D3803-1989 test methodology is not
expected to alter the filters; therefore, it will not adversely
alter the resulting filter performance. Since the plant functions
and equipment are not altered, the proposed revision will not create
a new or different kind of accident.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
Analyses of design-basis accidents assume a particular ESF
[Engineered Safety Feature] charcoal filter adsorption efficiency
when calculating offsite and control room operator doses. Charcoal
filter samples are tested to determine whether the filter adsorber
efficiency is greater than that assumed in the design-basis accident
analysis. The laboratory test acceptance criteria contains a safety
factor to ensure that the efficiency assumed in the accident
analysis is still valid at the end of the operating cycle. Because
ASTM D3803-1989 is a more accurate and demanding test than older
tests, NRC indicated in GL 99-02 that licensees upgrading their TS
to this new protocol will be able to use a safety factor as low as 2
for determining the acceptance criteria for charcoal filter
efficiency. This safety factor can be used for systems with or
without humidity control because the lack of humidity control is
already accounted for in the test conditions. As stated in the GL,
the new test protocol and associated safety factors have been
reviewed and found to not significantly decrease the margin of
safety.
The NRC 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.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: December 16, 1999.
Brief description of amendment: The proposed change would amend
Technical Specification 4.18.5.b to allow tube 110/60 to remain in
service through the current operating cycle (cycle 16) with two axial
indications that have potential through-wall depths greater than the
plugging limit. The axial indications are located in the roll
transition region and are contained within the upper tubesheet.
Date of publication of individual notice in Federal Register:
December 29, 1999 (64 FR 73080).
Expiration date of individual notice: Comments on no significant
hazards considerations by January 12, 2000; requests for hearing by
January 28, 2000. Clarification: The December 29, 1999, notice
indicated that requests for a hearing with respect to issuance of this
amendment must be filed by January 12, 2000. The correct deadline for
this action is January 28, 2000.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: December 16, 1999.
Description of amendment request: The proposed changes would revise
the River Bend Station (RBS) Technical Requirements Manual, Section TR
[[Page 1930]]
3.9.14, and add an exception to the current prohibition for travel of
loads in excess of 1200 pounds over fuel assemblies in the spent fuel
storage pool. The exception would allow the licensee to move the spent
fuel pool (SFP) watertight gates, which separate the SFP from the cask
and lower transfer pools, to perform maintenance and repairs on the
gates and watertight seals. Related sections of the RBS Updated Safety
Analysis Report would also be revised to be consistent with the
exception. The licensee determined that movement of the gate, with its
associated rigging, over spent fuel would involve an unreviewed safety
question in accordance with Title 10 of the Code of Federal
Regulations, Section 50.59.
Date of publication of individual notice in Federal Register:
December 21, 1999 (64 FR 71511).
Expiration date of individual notice: January 20, 2000. Correction:
The December 21, 1999, notice indicated that requests for a hearing
with respect to issuance of this amendment must be filed by January 28,
2000. The correct deadline for this action is January 20, 2000.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: December 16, 1998, as
supplemented July 16, September 29, and December 21, 1999.
Brief description of amendments: The amendments revise Technical
Specifications 3.8.1 and 3.37 to ensure that the appropriate actions
are taken to prevent double sequencing of safety-related loads and that
the setpoint allowable values for the degraded voltage relays reflect
the required function of the relays.
Date of issuance: December 29, 1999.
Effective Date: December 29, 1999, to be implemented within 90
days.
Amendment Nos: Unit 1-123, Unit 2-123, Unit 3-123.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14279) The July 16, September 29, and December 21, 1999, letters
provided additional clarifying information that was written within the
scope of the original application and Federal Register notice and did
not change the staff's initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated December 29, 1999.
No significant hazards consideration comments received: No.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of application for amendments: September 1, 1999.
Brief description of amendments: The amendments revised the
Technical Specifications as follows:
1. Technical Specification 1.1 is changed to replace the definition
of Azimuthal Power Tilt with a new definition.
2. Technical Specification 2.1.1.2 is changed by replacing the peak
linear heat rate safety limit with less than or equal to 22 kW/ft.
3. Technical Specification Surveillance Requirement (SR) 3.3.6.2 is
changed by replacing the degraded voltage function with transient
degraded voltage and steady-state degraded voltage functions.
4. Technical Specification SRs 3.8.1.9 and 3.8.1.15 are changed by
replacing the steady-state voltage range with the range of greater than
or equal to 4060 volts and less than or equal to 4400 volts.
5. Technical Specification 5.6.5.a is changed by adding Technical
Specifications 3.1.4 and 3.3.1 to the list.
6. Technical Specification Figure 2.1.1-1 is changed by removing
the reference to Figure B2.1-1.
7. Various Technical Specifications and Figures 2.1.1-1a are
changed by removing references to Unit 2, Cycle 12, and deleting Figure
2.1.1-1a.
8. Technical Specification 5.6.5.b, Item 41.ii is changed by
correcting CEN-99(B)-P to CEN-119(B)-P.
Date of issuance: December 15, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 232 and 208.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 6, 1999 (64 FR
54372).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated December 15, 1999.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: August 6, 1999, as supplemented
on November 15, 1999.
Brief description of amendments: The amendments revised Technical
[[Page 1931]]
Specification 3/4.4.6, ``Vacuum Relief'' to remove specific operability
requirements related to position indication for the suppression
chamber-drywell vacuum breakers. The amendments also reformat the
action statement for inoperable vacuum breakers, increase the
surveillance interval for verifying that the vacuum breakers are
closed, and delete the requirement to verify that the manual isolation
valves are closed for an inoperable and open vacuum breaker.
Date of issuance: December 21, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 138 and 122.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46428).
The November 15, 1999, submittal provided additional clarifying
information that did not change the staff's initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated December 21, 1999.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: July 16, 1999.
Brief description of amendments: The amendments revise Technical
Specification 4.7.D.6 by replacing the leakage limit of 11.5 standard
cubic feet per hour (scfh) for each main steam isolation valve (MSIV)
with a limit of 46 scfh on the total combined leakage for the MSIVs of
all four main steam lines.
Date of issuance: December 21, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 192 and 188.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46429).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 21, 1999.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: October 15, 1999.
Brief description of amendments: The amendments revise Section
5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of the
Technical Specifications by adding a new paragraph. The existing single
paragraph of Section 5.5.7 requires that inspection of each reactor
coolant pump flywheel be done per the recommendations of Regulatory
Position C.4.b of Regulatory Guide 1.14. The amendments add a new
paragraph which specifies that in lieu of Regulatory Positions C.4.b(1)
and C.4.b(2), alternative inspection techniques may be used. Date of
issuance: December 21, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment Nos.: 182 (Unit 1); 174 (Unit 2).
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 17, 1999 (64
FR 62705).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 21, 1999.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: October 15, 1999.
Brief description of amendments: The amendments revise Section
5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of the
Technical Specifications by adding a new paragraph. The existing single
paragraph of Section 5.5.7 requires that inspection of each reactor
coolant pump flywheel be done per the recommendations of Regulatory
Position C.4.b of Regulatory Guide 1.14. The amendments add a new
paragraph which specifies that in lieu of Regulatory Positions C.4.b(1)
and C.4.b(2), alternative inspection techniques may be used.
Date of issuance: December 21, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment Nos.: 190 (Unit 1); 171 (Unit 2).
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 17, 1999 (64
FR 62706).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 21, 1999.
No significant hazards consideration comments received: No
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: December 18, 1998, as
supplemented September 13, 1999.
Brief description of amendment: This amendment revises the St.
Lucie, Unit 2 (SL-2), Technical Specifications (TS) Index Page III, TS
1.10, Dose Equivalent iodine-131; TS 2.1.1.2, Linear Heat Rate; TS
3.1.1.1/4.1.1.1.1, Shutdown Margin--Tavg Greater than 200
deg.F; TS 3/4.1.1.2, Shutdown Margin--Tavg Less Than or
Equal to 200 deg.F; TS 3.1.2.2, Boration Systems Flow Paths--Operating;
TS 3.1.2.4, Charging Pumps--Operating; TS 3.1.2.6, Boric Acid Makeup
Pumps--Operating; TS 3.1.2.8, Borated Water Sources--Operating; and TS
6.9.1.11, Core Operating Limits Report (COLR). The amendment also
relocates the core operating limits for shutdown margin to the SL-2
COLR. The following Bases have also been changed in connection with
this amendment: TS Bases 2.1.1, Reactor Core; Bases Figure B2.1-1,
Axial Power Distributions for Thermal Margin Safety Limits; TS Bases
2.2.1, Reactor Trip Setpoints (Variable Power Level-High); TS Bases 3/
4.1.1.1 and 3/4.1.1.2, Shutdown Margin; and TS Bases 3/4.1.2, Boration
Systems.
Date of Issuance: December 21, 1999.
Effective Date: As of date of issuance, to be implemented prior to
fuel reload for Cycle 12.
Amendment No.: 105.
Facility Operating License No. NPF-16: Amendment revised the TS.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6697). The supplemental letter dated September 13, 1999, provided
additional information that did not expand the scope of the amendment
request as noticed or change the original proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1999.
No significant hazards consideration comments received: No.
[[Page 1932]]
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida
Date of application for amendments: April 26, 1999.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) for Turkey Point Units 3 and 4 to correct
the Technical Specification Index and to remove inconsistencies, and
make administrative changes. A portion of the request, related to the
proposed deletion of dates for the approved security plans, was denied.
Date of issuance: December 20, 1999.
Effective date: December 20, 1999.
Amendment Nos.: 203 and 197.
Facility Operating License Nos. DPR-31 and DPR-41: Amendments
revised the TS.
Date of initial notice in Federal Register: June 2, 1999 (64 FR
29711).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 20, 1999.
No significant hazards consideration comments received: No.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit 1 Dauphin County, Pennsylvania
Date of application for amendment: December 3, 1998, as
supplemented January 11, February 4, March 4, March 10, and March 15,
1999.
Brief description of amendment: This amendment conforms the license
to reflect the transfer of Facility Operating License No. DPR-50 for
the Three Mile Island Nuclear Station, Unit 1, from GPU Nuclear, Inc.,
et al., to AmerGen Energy Company, LLC, as previously approved by Order
dated April 12, 1999.
Date of issuance: December 20, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 218.
Facility Operating License No. DPR-50: Amendment revised the
license and the Technical Specifications.
Date of initial notice in Federal Register: December 21, 1998 (63
FR 70436).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 1999.
Comments received: Yes. See safety evaluation dated April 12, 1999.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of application for amendment: April 12, 1999, as supplemented
October 5 and 8, 1999.
Brief description of amendment: The amendment revises Technical
Specification (TS) Surveillance Requirement (SR) 3.6.1.3.7 to allow a
``representative sample'' of reactor instrumentation line excess flow
check valves (EFCVs) to be tested every 24 months, instead of testing
each EFCV every 24 months.
Date of issuance: December 29, 1999.
Effective date: December 29, 1999.
Amendment No.: 230.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 14, 1999 (64 FR
38028).
The October 5 and 8, 1999, letters provided clarifying information
that was within the scope of the original Federal Register notice 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 December 29, 1999.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of application for amendment: December 3, 1998.
Brief description of amendment: This amendment revised the
Technical Specifications for sealed source leakage testing to
specifically address testing requirements for fission detectors.
Date of issuance: December 20, 1999.
Effective date: December 20, 1999, with full implementation within
45 days.
Amendment No.: 235.
Facility Operating License No. DPR-58: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43773).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 20, 1999.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: September 17, 1999, as
supplemented November 10 and 19, 1999.
Brief description of amendments: The amendments would approve the
licensee's revision of the Updated Final Safety Analysis Report and
Emergency Operating Procedures to use methodology to credit the
negative reactivity provided by insertion of the rod cluster control
assemblies (RCCAs) into the reactor core following any design basis
loss-of-coolant accident, during realignment from a cold leg
recirculation to a hot leg recirculation configuration. This change to
the licensing basis, when evaluated by the licensee in accordance with
10 CFR 59.59, resulted in an unreviewed safety question that requires
prior approval by the NRC staff in accordance with the provisions of 10
CFR 50.90 prior to implementation. The amendments also change the Bases
for Technical Specifications Section 3/4.5.5, Refueling Water Storage
Tank.
Date of issuance: December 28, 1999.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 236 and 218.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 20, 1999 (64 FR
56531).
The licensee's letters of November 10 and 19, 1999, provided
additional information that did not change scope of the application or
the staff's proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 28, 1999.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: November 5, 1999.
Brief description of amendment: The amendments would revise Unit 1
and 2 Technical Specification (TS) 3.5.1, Action ``a'' and ``b,'' to
reflect the monitoring of pressure from the Reactor Coolant System
instead of the pressurizer. The amendment would also revise Unit 1 and
2 TS Surveillance Requirement 4.5.1.c to require verification that
power is removed from each emergency core cooling system accumulator
isolation valve operator instead of verification that each accumulator
isolation valve breaker is physically removed from the circuit.
Furthermore, the amendment would make administrative changes to Unit 1
and 2 TS Bases 3/4.5.1.
Date of issuance: December 23, 1999.
[[Page 1933]]
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 237 and 219.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 23, 1999 (64
FR 65735).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 23, 1999.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of application for amendment: August 17, 1999.
Brief description of amendment: The amendment removes the steam
generator voltage-based repair criteria, F* repair criteria, and
sleeving methodologies from the Unit 1 Technical Specifications and
clarifies the Bases sections accordingly.
Date of issuance: December 22, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 238.
Facility Operating License No. DPR-58: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 6, 1999 (64 FR
54375).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 1999.
No significant hazards consideration comments received: No.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: November 8, 1999.
Brief description of amendment: The amendment changed action
statements, definitions, and footnotes pertaining to the Technical
Specifications for primary containment leakage and primary containment
purge system to allow an alternative approach for isolating a bypass
leakage path and/or purge system line.
Date of issuance: December 16, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 87.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 16, 1999 (64
FR 62228).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 16, 1999.
No significant hazards consideration comments received: No
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: March 31, 1999.
Brief description of amendment: Amendment changes Technical
Specification Table 3.6.1.2-1 by adding two relief valves, and
associated leak rate criteria, to be installed on the drywell equipment
drain line and drywell floor drain line.
Date of issuance: December 16, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 88.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24197).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 16, 1999.
No significant hazards consideration comments received: No
North Atlantic Energy Service Corporation, et al., Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: March 5, 1998.
Description of amendment request: This amendment revises the
Technical Specifications (TSs) by relocating the procedural details of
the Radiological Effluent Technical Specifications (RETS) to the
Offsite Dose Calculation Manual. The TSs were also revised to relocate
procedural details associated with solid radioactive wastes to the
Process Control Program. In addition, the Administrative Controls
section of the TSs was revised to incorporate programmatic controls for
radioactive effluents and environmental monitoring. These changes are
consistent with the guidance provided in Generic Letter 89-01,
``Implementation of Programmatic Controls for Radiological Effluent
Technical Specifications in the Administrative Controls Section of the
Technical Specifications and the Relocation of Procedural Details of
RETS to the Offsite Dose Calculation Manual or to the Process Control
Program.''
Date of issuance: December 15, 1999.
Effective date: As of its date of issuance, and shall be
implemented within 120 days.
Amendment No.: 66.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19972). The Commission received comments which were addressed in the
staff's Safety Evaluation dated December 15, 1999.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 15, 1999.
No significant hazards consideration comments received: Yes.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: October 20, 1999.
Brief description of amendment: The amendment changes from December
31, 1999, to June 30, 2001, the date specified in TS 4.3.1.1.b Note
associated with maintaining spent fuel pool boron concentration >2300
ppm at all times until a permanent resolution to the current
criticality concerns is implemented.
Date of issuance: December 21, 1999.
Effective date: December 21, 1999.
Amendment No.: 75.
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 19, 1999 (64
FR 63345).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1999.
No significant hazards consideration comments received: No.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: September 4, 1998, as
supplemented on February 8, April 16, August 26, September 16, and
November 17, 1999.
Brief description of amendment: The amendment increases the spent
fuel pool storage capacity from 2,870 to 3,353 fuel assemblies.
Date of Issuance: December 21, 1999.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
[[Page 1934]]
Amendment No.: 182
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 1, 1998 (64 FR
52774). The supplemental information did not affect the staff's
proposed no significant hazards consideration determination, and was
within the scope of the original amendment application as published.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated December 21, 1999.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: November 18, 1998, as
supplemented by letter dated October 22, 1999.
Brief description of amendments: The amendments change the North
Anna Power Station Technical Specifications (TS) to increase the
allowable groundwater elevation at the southeast section of the service
water reservoir dike from 277 to 280 feet at the toe and from 280 to
295 feet at the crest. In addition, TS Table 3.7-6 has been reorganized
to clarify zones of interest in the Service Water Reservoir, the
location of piezometer devices, and piezometer device numbers. The
proposal to eliminate device numbers from the TS was denied because the
device number helps to indicate the location of the piezometer within
the zone as well as the piezometer itself. Finally the column heading
for Allowable Drain Flow Rate was clarified to be the total flow rate.
Date of issuance: As of the date of issuance and shall be
implemented within 30 days.
Effective date: December 29, 1999.
Amendment Nos.: 220 and 201.
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69349). The supplemental letter dated October 22, 1999, contained
clarifying information only, and did not change the initial no
significant hazards determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 29, 1999.
No significant hazards consideration comments received: No.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: April 12, 1999.
Brief description of amendments: These amendments update references
in the Technical Specifications to information in the updated Final
Safety Analysis Report (FSAR). The update is necessary to reflect
relocation of the referenced information in the updated FSAR.
Date of issuance: December 23, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-192; Unit 2-197.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24204).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 23, 1999.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland this 5th day of January 2000.
For the Nuclear Regulatory Commission.
Suzanne Black,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 00-611 Filed 1-11-00; 8:45 am]
BILLING CODE 7590-01-P