[Federal Register Volume 63, Number 184 (Wednesday, September 23, 1998)]
[Notices]
[Pages 50932-50948]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-25281]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 28, 1998, through September 11, 1998.
The last biweekly notice was published on September 9, 1998 (63 FR
48256).
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 and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By October 23, 1998, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or
[[Page 50933]]
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: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: June 26, 1998.
Description of amendment request: The proposed Technical
Specification (TS) amendment would amend various TS pages to correct
typographical errors, remove inadvertent replication of information,
and update various Bases sections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed administrative changes involving typographical
errors and updating the Bases reflect plant design, safety limit
settings, and plant system operation previously reviewed and
approved by the NRC. These changes, therefore, do not modify or add
any initiating parameters that would significantly increase the
probability or consequences of any previously analyzed accident.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
These proposed changes do not involve any potential initiating
events that would create a new or different kind of accident.
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 amendment does not involve a significant
reduction in a margin of safety.
These changes reflect information previously reviewed and
approved by the NRC. The proposed changes will make the information
in the Technical Specifications consistent with that already
approved by the NRC. Therefore, it is concluded that the proposed
amendment does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Cecil O. Thomas.
[[Page 50934]]
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: April 25, December 23, 1996, August 8,
September 5, 1997, March 26, July 31, and August 24, 1998. The August
24, 1998, supplement supersedes the previous no significant hazards
consideration determination included in letters dated April 25, 1996,
and March 26, 1998 for the EDG AOT.
Description of amendment request: The proposed Technical
Specification (TS) amendment would extend the Emergency Diesel
Generator (EDG) allowed outage time (AOT) from 72 hours to 14 days. In
support of this change the licensee has proposed various TS changes to
decrease the consequences of the extended AOT.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Operation of Pilgrim Nuclear Power Station in accordance with
the proposed license amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated because of the following:
An Individual Plant Examination (IPE) for Internal Events was
submitted to the NRC in response to Generic Letter 88-20 in
September 1992. The supporting probabilistic safety analysis (PSA)
model was updated as described in BECo letter 95-127, dated December
28, 1995. The updated PSA model was used to quantify the overall
impact of the proposed EDG 14-day AOT on core damage frequency. Part
III of BECo No. 2.96.040 provides the results of a comprehensive
[probabilistic safety assessment] PSA of the impact of the proposed
AOTs for the EDGs and [startup transformer] SUT and [shutdown
transformer] SDT. As shown in Part III, there is no significant
increase in risk due to the proposed change. Thus, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The existing specification 3.9.B.1 is separated into two
segments (a and b) because of the proposed different AOTs for the
SUT and SDT transformers. As a result of the PSA, the AOT for the
SUT (a) is reduced from 7 days to 72 hours, while the AOT for the
SDT (b) remains at 7 days. The reduction of the AOT from 7 days to 3
days is based on the relative risk importance of the SUT support to
the balance of plant systems. Similarly, an additional reduction
from 72 hours to 48 hours is proposed in the AOT for a simultaneous
loss of both the SUT or SDT and an EDG (TS 3.9.B.4) based upon the
SUT's or SDT's contribution to risk and that two power sources have
been removed from the associated bus. The AOT reductions represent a
measurable decrease in risk as assessed in the PSA. Thus, the
probability or consequences of an accident previously evaluated are
not increased.
The current technical specifications allow one EDG to be out of
service for three days based on the availability of the SUT and SDT
and the fact that each EDG carries sufficient engineered safeguards
equipment to cover all design basis accidents. Additionally, the SDT
can provide adequate power for one train of ESF equipment for all
operating, transient, and accident conditions. With one EDG out of
service and a Loss of Offsite Power (LOOP) condition, the capability
to power vital and auxiliary system components remains available via
the other EDG. Increasing the EDG AOT to 14 days provides
flexibility in the maintenance and repair of the EDGs. The EDG
unavailability will be monitored and trended in accordance with the
Maintenance Rule. The PSA analyses supports the change to a 14 day
AOT for the EDGs based on an insignificant increase in overall risk.
Implementation of the proposed change is expected to result in less
than a one percent increase in the baseline core damage frequency
(2.84E-05/yr), which is considered to be insignificant relative to
the underlying uncertainties involved with PSA. An additional
condition is added requiring the SBO-DG to remain operable for
extending the inoperable EDG AOT from 3 days to 14 days, thereby
assuring that one EDG and SBO-DG are available during the extended
EDG AOT. Thus, the 14-day EDG AOT does not involve an increase in
the probability or consequences of an accident previously evaluated.
The proposed addition of the CRMP does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Because the changes are administrative in
nature and deal only with risk assessment, they have no bearing on
accident initiation or mitigation. Therefore, the changes will not
affect the probability or consequences of an accident previously
evaluated.
The proposed change does not affect the design or performance of
the EDGs, and the change will not result in a significant increase
in the consequences or probability of an accident previously
analyzed. These changes do not involve a increase in the probability
or consequences of an accident previously analyzed.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The operation of PNPS in accordance with the proposed license
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated because of the
following:
The proposed amendment will extend the action completion/allowed
outage time for an inoperable EDG from 3 days to 14 days. During
this extension, the [station blackout diesel generator] SBO-DG is
required to be operable and normal breaker configuration is required
to be verified to ensure the SBO-DG is capable of energizing the
safety bus associated with the inoperable EDG. These actions assure
one EDG and SBO-DG are operable during extended EDG AOTs. The EDGs
are designed as backup AC power sources for essential safety systems
in the event of loss of offsite power. The SBO-DG is designed to
cope with a station black out transient. The proposed AOT does not
change the conditions, operating configurations, or minimum amount
of operating equipment assumed in the safety analysis for accident
mitigation. The EDGs, SBO-DG and AC equipment are not accident
initiators. No change is being made in the manner in which the EDGs
provide plant protection. No new modes of plant operation are
involved. An extended AOT for one EDG does not create a new or
different kind of accident [than] previously evaluated. The PSA
results concluded the risk contribution of the EDG AOT extension is
insignificant.
Pilgrim has implemented an EDG reliability program to maintain
reliability of EDGs. The SBO-DG is included in the reliability
program, and the performance of EDGs and SBO-DG are trended for
compliance with Maintenance Rule requirements. Thus, the proposed
change does not introduce any new mode of plant operation or new
accident precursors, involve any physical alterations to plant
configurations, or make changes to system set points that could
initiate a new or different kind of accident. Therefore, operation
in accordance with the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
The AOT for an inoperable SUT is reduced from 7 days to 72 hours
based upon the PSA that was performed to quantitatively assess the
risk impact of the proposed amendment. Additionally, removal of the
SUT from service degrades the reliability of the offsite power
system and renders the balance of plant unavailable upon a plant
shutdown. The proposed reduction in AOT improves overall AC power
source availability because the SUT will potentially be inoperable
for shorter time periods. Therefore, reducing the AOT does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
The proposed addition of the [Configuration Risk Management
Program] CRMP does not create the possibility of a new or different
kind of accident from any accident previously evaluated because the
CRMP will not affect the manner in which [structures, systems, and
components] SSCs are designed, operated, or maintained. The
administrative changes proposed will only require a risk assessment
for specified plant configurations. Any risk assessments performed
as a result of this program will only serve to provide plant
personnel with risk insights associated with particular plant
configurations. Since the changes will not impact SSCs and all
accidents involving SSCs, the proposed change does not create a new
kind of accident from any previously evaluated.
[[Page 50935]]
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The operation of PNPS in accordance with the proposed license
amendment will not involve a significant reduction in a margin of
safety. As shown in Part III [of the application dated April 25,
1996), incorporation of the proposed change involves an
insignificant reduction in the margin of safety (less than a one
percent increase in the baseline core damage frequency (2.84E-05/
yr), which is considered to be insignificant relative to the
underlying uncertainties involved with PSA).
Also, the proposed changes do not significantly reduce the basis
for any technical specification related to the establishment of, or
the maintenance of, a safety margin nor do they require physical
modifications to the plant. An additional condition is added
requiring the SBO-DG to remain operable, in addition to the operable
EDG associated with the redundant train while in the 14-day EDG AOT.
The PSA results showed that the risk contribution of extending the
AOT for an inoperable EDG is insignificant. Also, the reduction in
the AOT for the SUT should improve availability thereby reducing
overall risk with no reduction of the safety margin. Moreover, the
proposed changes affect neither the way in which the EDGs perform
their safety function nor the bases for their LCOs.
The proposed change does not involve a significant reduction in
a margin of safety. The proposed administrative change to include a
risk management program will not impact how plant SSCs are designed,
operated, or maintained. The required risk assessments are intended
to provide insights that influence decisions on the acceptability of
abnormal plant configurations. These insights work in conjunction
with existing inputs into the decision-making process rather than as
the sole basis for making decisions. Therefore, the changes will not
reduce a margin of safety.
As previously stated, implementation of the proposed changes is
expected to result in an insignificant increase in: (1) power
unavailability to the emergency buses (given that a loss of offsite
power has occurred), and (2) core damage frequency. Implementation
of the proposed changes does not 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W.S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Cecil O. Thomas.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: September 1, 1998.
Description of amendment request: The licensee's request proposes
to revise Technical Specification 3/4.9.11 ``Water Level--New and Spent
Fuel Pools.'' As a result of the proposed amendment, the licensee has
also revised the Fuel Handling Building fuel handling accident analysis
and the Containment fuel handling accident analyses.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Revising the required spent fuel pool water level will not
increase the probability of a fuel handling accident. There is no
other physical alteration to any plant system, nor is there a change
in the method in which any safety related system performs its
function. Harris Nuclear Plant (HNP) has revised the fuel handling
accident analyses using the conservative assumptions associated with
this change. The revised fuel handling accident analyses demonstrate
that dose consequences as a result of a fuel handling accident
remain below 25% of the 10 CFR 100 guidelines as described in the
NRC Standard Review Plan.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated
because there is no physical alteration to any plant system, other
than revising spent fuel pool water level, nor is there a change in
the method in which any safety related system performs its function.
HNP has design features to mitigate the consequences of a loss of
spent fuel pool water level which are unaffected by this change.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
Revising the required spent fuel pool water level does not
involve a significant reduction in the margin of safety. There is no
other physical alteration to any plant system, other than revising
spent fuel pool water level, nor is there a change in the method in
which any safety related system performs its function. HNP has
revised the fuel handling accident analyses using the conservative
assumptions associated with this change. The revised fuel handling
accident analyses demonstrate that dose consequences as a result of
a fuel handling accident remain below 25% of the 10 CFR 100
guidelines as described in the NRC Standard Review Plan.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Pao-Tsin Kuo.
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 amendment request: August 31, 1998.
Description of amendment request: The proposed amendment would
change the Quad Cities Technical Specifications (TS) to reflect an
increase in the maximum allowable Main Steam Isolation Valve (MSIV)
leakage from 11.5 standard cubic feet per hour (scfh) to 30 scfh per
valve when tested at 25 psig, in accordance with Surveillance
Requirement 4.7.D.6
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change to Technical Specification Surveillance
Requirement 4.7.D.6 increases the maximum allowable leakage rate for
a single Main Steam Isolation Valve (MSIV) from 11.5 scfh to 30
scfh. This change has no impact on the automatic or manual closure
features of the valve including automatic actuations and response
times. Closure of the MSIVs is a postulated transient considered in
the design basis of the plant. Since the proposed change does not
impact the response characteristics of the MSIVs during a postulated
transient
[[Page 50936]]
condition, the change does not impact the probability of an accident
previously evaluated.
The change in allowable MSIV leakage has been evaluated to
assess the impact on control room operator dose and offsite dose
levels. The radiological assessment was performed with an updated
radiological methodology that included significant enhancements,
such as credit for suppression pool scrubbing, updated iodine dose
conversion factors, and allowance for higher burnup fuel designs.
Using this revised methodology, which is consistent with current
regulatory requirements, the resulting dose levels from a postulated
design basis accident continue to remain below the limits
established in 10 CFR 50, Appendix A, General Design Criteria 19
(GDC-19) and 10 CFR 100. Therefore, the proposed change does not
involve a significant increase in the consequences of an accident
previously evaluated
Therefore this proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The safety function of the MSIVs is to provide a timely steam
line isolation to mitigate the release of radioactive steam and
limit reactor inventory loss under certain accident and transient
conditions. The MSIVs are designed to automatically close whenever
plant conditions warrant a main steam line isolation. The proposed
increase in allowable MSIV leakage does not impact the MSIV's
ability to perform its underlying safety function, nor does the
change involve any physical features of the valves and associated
steam lines to create a new or different type of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed increase in allowable MSIV leakage represents a
nominal increase in the release of radioactivity during a design
basis event. The radiological assessment was performed with an
updated radiological methodology that included significant
enhancements, such as credit for suppression pool scrubbing, updated
iodine dose conversion factors, and allowance for higher burnup fuel
designs. Using this revised methodology, which is consistent with
current regulatory requirements, the resulting dose levels from a
postulated design basis accident continue to remain below the limits
established in 10 CFR 50, Appendix A, GDC-19 and 10 CFR 100.
Therefore, these changes do not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Stuart A. Richards.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of application of amendments: June 2, 1998.
Description of amendment request: The proposed amendment relocates
seismic monitoring equipment requirements from the Technical
Specifications to the Technical Requirements Manual (TRM), a document
which is controlled under 10 CFR 50.59.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
CYAPCO has reviewed the proposed changes to the Technical
Specifications in accordance with 10 CFR 50.92 and concluded that the
changes do not involve a significant hazards consideration (SHC). The
basis for this conclusion is that the three criteria of 10 CFR 50.92(c)
are not compromised. The proposed changes do not involve an SHC because
the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
As a result of the present plant configuration which has the
fuel permanently removed from the reactor, the reactor-related
accidents previously evaluated (i.e., LOCA, MSLB, etc.) are no
longer possible. The accidents previously evaluated that are still
applicable to the plant are fuel handling accidents and gaseous and
liquid radioactive releases.
There is no significant increase in the probability of a fuel
handling accident since refueling operations have ceased. In fact,
there is a decrease in probability of a fuel handling accident since
the need to move/rearrange fuel assemblies is minimal until they are
removed from the spent fuel pool (i.e., for dry cask storage or for
transferring to USDOE possession). In addition, the consequences of
a fuel handling accident are continuing to decrease since the fuel
in the spent fuel pool is continuing to decay.
The radiological consequences of a gaseous or liquid radioactive
release are bounded by the fuel handling accident during defueled
operation and a spent resin fire during the reactor coolant system
decontamination. With the plant defueled and permanently shutdown,
the demands on the radwaste systems are lessened since no new
radioisotopes are being generated by irradiation or fission.
Therefore, there is no increase in the probability or consequences
of a gaseous or liquid radioactive release.
The ability of the plant to withstand a seismic event is not
affected by this proposed change. The seismic instrumentation does
not actuate any protective equipment or serve any direct role in the
mitigation of an accident. The equipment will continue to be
adequately controlled by the Technical Requirements Manual (TRM) to
ensure operability and alert operators to a seismic event, should
one occur, so that appropriate actions can be taken. Therefore,
there is no increase in the consequences of a seismic event.
This material is being transferred to the TRM. This transfer is
in accordance with Generic Letter 95-10, ``Relocation of Selected
Technical Specifications Requirements Related to Instrumentation,''
dated December 15, 1995 and is consistent with the NUREG-1431,
``Standard Technical Specifications, Westinghouse Plants,'' Volume
1, Revision 1, dated April, 1995. The removed material included in
this category is Technical Specification 3/4.3.3.3 and the related
tables.
Based on the above, the proposed changes to the Technical
Specifications do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
There is no change in how spent fuel is stored or moved in the
spent fuel pool. Therefore, the postulated fuel handling accidents
are still bounding and are still considered as credible postulated
accidents.
There is no change in the design and construction of plant
systems, structures and components with respect to the capability to
withstand a seismic event. Therefore, the currently assumed
radioactive releases are still bounding.
This material is being transferred to the TRM. This transfer is
in accordance with Generic Letter 95-10 and is consistent with
NUREG-1431. The removed material included in this category are
Technical Specification 3/4.3.3.3 and the related tables.
Based on the above, the proposed changes to the Technical
Specifications do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The capability of the plant to withstand a seismic event or
other design basis accident is determined by the design and
construction of systems, structures, and components. The
instrumentation is used to alert operators to the seismic event and
evaluate the plant response. The NRC's Final Policy Statement on
Technical Specification Improvements (SECY-93-067) stated that
instrumentation to detect precursors to reactor coolant pressure
boundary leakage, such as seismic instrumentation, is not included
in the first criterion. As discussed above, the seismic
[[Page 50937]]
instrumentation does not serve as a protective design feature or
part of a primary success path for events which challenge fission
product barriers. The NRC staff, in Generic Letter 95-10, has
concluded that the seismic monitoring instrumentation does not
satisfy the 10 CFR 50.36 criteria and need not be included in the
technical specifications.
This material is being transferred to the TRM. This transfer is
in accordance with Generic Letter 95-10 and is consistent with
NUREG-1431. The removed material included in this category are
Technical Specification 3/4.3.3.3.
The proposed changes to the Technical Specifications do not
involve a significant reduction in a margin of safety due to the
fact that the capability of the plant to withstand a seismic event
or other design bases accident is not affected by this proposed
change.
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
amendments request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457.
Attorney for the licensee: Mr. John A. Ritsher, Esquire, Ropes &
Gray, One International Place, Boston, Massachusetts, 02110.
NRC Project Director: Seymour H. Weiss, Director.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: April 9, 1998 (NRC-98-0071).
Description of amendment request: The proposed amendment would
revise the ``**'' footnote to Technical Specification (TS) 3.7.1.2,
``Emergency Equipment Cooling Water System,'' Action ``a'' and add a
``*'' footnote to TS 3.8.1.1, ``A.C. Sources--Operating,'' Action ``c''
to make the actions consistent with TS 3.3.7.5, ``Accident Monitoring
Instrumentation,'' for the case of inoperable primary containment
oxygen monitoring instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change will permit operation with both of the
primary containment oxygen monitoring instrument channels inoperable
for up to 48 hours before requiring entry into a 12 hour shutdown
statement, consistent with Technical Specification 3.3.7.5, but less
restrictive than the requirements in Technical Specification 3.7.1.2
Action a and Technical Specification 3.8.1.1 Action c, which require
entry into the 12 hour shutdown statement immediately if the channel
in the remaining division is inoperable, followed by continued
shutdown to the COLD SHUTDOWN condition. The shutdown action
statement entry conditions for the primary containment oxygen
monitoring instrumentation should be no more restrictive in
Technical Specification 3.7.1.2 or Technical Specification 3.8.1.1,
than they are in Technical Specification 3.3.7.5 for both channels
being inoperable. The primary containment oxygen monitoring
instrumentation provides the same non-critical function regardless
of the reason for the system inoperability. The primary containment
oxygen monitors provide the control room operators with indication
and alarm of the oxygen concentration in the primary containment,
but do not provide any automatic function to mitigate an accident.
Because they perform only a monitoring function, the oxygen monitors
are not associated with the initiation of any previously evaluated
accident; therefore, there is no change in the probability of an
accident previously evaluated.
The indication provided by the primary containment oxygen
monitors is used by the control room operators to ensure that the
oxygen concentration remains within limits and to help make
decisions regarding the use of the Combustible Gas Control System,
if necessary. Alternate methods using grab samples and laboratory
analytical equipment are available for obtaining primary containment
oxygen concentration if no primary containment oxygen monitoring
instrumentation is available. Additionally, the loss of both oxygen
analyzers is not critical for entry into the Emergency Operating
Procedures. Entry conditions for the post accident control of
hydrogen are based upon the primary containment hydrogen monitor
readings, and both channels of primary containment hydrogen
monitoring instrumentation are still required to remain operable in
accordance with Technical Specification 3.3.7.5. Therefore, this
change will not involve a significant increase in the consequences
of a previously evaluated accident.
2. The change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
As discussed above, the primary containment oxygen monitors are
indication and alarm only instruments which provide information to
the control room operators. The proposed change does not introduce a
new mode of plant operation, nor does it involve a physical
modification to the plant. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed change involves the length of time that both
primary containment oxygen monitoring instrument channels may be out
of service. It does not increase the out of service time beyond what
is already allowed by Technical Specification 3.3.7.5 for both
channels being inoperable. The primary containment oxygen monitors
are indication and alarm only instruments which do not affect any
parameters or assumptions used in the calculation of any safety
margin associated with Technical Specification Safety Limits,
Limiting Safety System Settings, Limiting Control Settings or
Limiting Conditions for Operation, or other previously defined
margins for any structure, system, or component. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Cynthia A. Carpenter.
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: August 24, 1998.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TS) to clarify, for St. Lucie
Units 1 and 2, component operations to be verified in response to a
containment sump recirculation signal. For St. Lucie Unit 1, the
proposed amendment would modify the list of equipment that comprises an
operable control room emergency ventilation system to more accurately
reflect installed equipment. For St. Lucie Unit 2, license conditions
related to the movement of spent nuclear fuel between units will be
deleted and modified as appropriate to reflect the completion of the
Unit 1 spent fuel pool re-rack activities.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not
[[Page 50938]]
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed amendments do not involve accident initiators. The
changes to the Unit 1 and Unit 2 Technical Specifications provide
additions and clarification to component lists to ensure that
explicit terms of the affected specifications are consistent with
existing requirements. Other changes to the Unit 2 facility
operating license simply delete superseded license conditions that
have been previously satisfied and are therefore obsolete. The
revisions do not involve any change to the configuration or method
of operation of any plant equipment that is used to mitigate the
consequences of an accident, nor do the changes alter any
assumptions or conditions in the plant safety analyses. Therefore,
operation of either facility in accordance with its proposed
amendment would 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 amendments are administrative in nature and will
not change the physical plant or the modes of operation defined in
the facility operating licenses. The changes do not involve the
addition or modification of equipment nor do they alter the design
or operation of plant systems. Therefore, operation of either
facility in accordance with its 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 changes proposed for the Unit 1 and Unit 2 Technical
Specifications provide additions and clarification to component
lists to ensure that explicit terms of the affected specifications
are consistent with existing requirements. Other changes to the Unit
2 facility operating license simply delete superseded license
conditions that have been previously satisfied and are therefore
obsolete. The revisions do not alter the plant safety analyses or
the basis for any technical specification that is related to the
establishment of, or the maintenance of, a nuclear safety margin.
Therefore, operation of either unit in accordance with its proposed
amendment would not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment requests: September 4, 1998.
Description of amendment requests: The proposed amendments would
modify the surveillance requirements and limiting conditions for
operation of the technical specifications (TS) for the reactor coolant
vent system. Specifically, the proposed amendments would modify the
limiting conditions for operation as specified in TS Section 3.1.A.3,
Reactor Coolant Vent System, and the surveillance requirements
specified in TS Section 4.18, Reactor Coolant Vent System Paths.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not affect any system that is a
contributor to initiating events for previously evaluated
anticipated operational occurrences and design basis accidents.
Neither do the changes significantly affect any system that is used
to mitigate any previously evaluated anticipated operational
occurrences and design basis accidents. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
The proposed changes do not alter the design, function, or
operation of any plant component and does not install any new or
different equipment, therefore the possibility of a new or different
kind of accident from those previously analyzed has not been
created.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The proposed changes do not alter the initial conditions assumed
in deterministic analyses associated with either the RCS [reactor
coolant system] boundary or fuel cladding, therefore these changes
do not involve a significant reduction in the margins 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.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: August 25, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 2.1.2, ``THERMAL POWER, High
Pressure and High Flow,'' and the Bases for TS 2.1, ``Safety Limits.''
These changes are being made to implement an appropriately conservative
Safety Limit Minimum Critical Power Ratio (SLMCPR) for the upcoming
Cycle 9 Hope Creek core and fuel designs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The derivation of the revised SLMCPRs for Hope Creek for
incorporation into the Technical Specifications, and its use to
determine cycle-1 specific thermal limits, have been performed using
NRC approved methods. These calculations do not change the method of
operating the plant and have no effect on the probability of an
accident initiating event or transient.
There are no significant increases in the consequences of an
accident previously evaluated. The basis of the MCPR Safety Limit is
to ensure that no mechanistic fuel damage is calculated to occur if
the limit is not violated. The new SLMCPRs preserve the existing
margin to transition boiling and the probability of fuel damage is
not increased. Therefore, the proposed change does not involve an
increase in the probability or consequences of an accident
previously evaluated.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
[[Page 50939]]
The proposed changes contained in this submittal result from an
analysis of the Cycle 9 core reload using the same fuel types as
previous cycles. These changes do not involve any new method for
operating the facility and do not involve any facility
modifications. No new initiating events or transients result from
these changes. Therefore, the proposed Technical Specification
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 margin of safety as defined in the Technical Specification
bases will remain the same. The new SLMCPRs are calculated using NRC
approved methods, which are in accordance with the current fuel
design, and licensing criteria. The MCPR Safety Limit remains high
enough to ensure that greater than 99.9% of all fuel rods in the
core will avoid transition boiling if the limit is not violated,
thereby preserving the fuel cladding integrity. Therefore, the
proposed Technical Specification changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: Robert A. Capra.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of amendment requests: March 6, 1998.
Description of amendment requests: The proposed amendment would
modify the Technical Specifications (TS) to eliminate reference to
shutdown cooling (SDC) system isolation bypass valve inverters. The
proposed change would allow the licensee to replace the inverters with
transfer switches.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The staff's evaluation of the three criteria are
presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The SDC system isolation bypass valves are not considered as event
initiators in the accidents analyzed in the safety analysis report.
Therefore, the proposed change in how the valves are aligned to
available power supplies does not affect the probability of an accident
previously evaluated.
The SDC system isolation bypass valves are realigned post-accident
to place the shutdown cooling system in operation. The proposed change
will modify the power supply for these valves from an inverter that is
supplied from the safety-related DC buses to the safety-related AC
buses through a manual transfer switch. This will allow the power
supplies for opposite trains' valves for SDC suction supplies to be
powered from opposite trains of electrical power. The operations
required to actually place SDC in operation from the control room are
unaffected. The proposed change does not affect the course of any
accident previously analyzed, and therefore the consequences of any
accident previously evaluated are unaffected by the proposed change.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The SDC system isolation bypass valves are used during accident
mitigations, and are not considered as credible accident initiators.
Thus, modifying the manner in which power is supplied to the valves
will 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.
Current accident analyses assume proper operation of the SDC system
to mitigate the consequences of an accident to maintain postulate
offsite release below the limits of 10 CFR Part 100. The proposed
change only modifies the manner in which power is made available to the
valves, while retaining the current design for redundancy and
diversity.
The proposed change does not, therefore, affect the current margins
of safety.
Based on the above staff analysis, it appears that the three
standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes
to determine that the amendment requests involve no significant hazards
consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.
Tennessee Valley Authority, Docket No. 50-260 and 50-296, Browns Ferry
Nuclear Plant Units 2, 3, Limestone County, Alabama
Date of amendment request: September 4, 1998.
Description of amendment request: The proposed amendment would
revise the licensing bases for the Browns Ferry Nuclear Plant (BFN)
Units 2 and 3 to credit containment pressure in excess of atmospheric
pressure (containment overpressure) in the analysis for Emergency Core
Cooling Systems (ECCS) pump required net positive suction head (NPSH)
during design basis accident conditions. The proposed licensing bases
change would be implemented by a change to the BFN Updated Final Safety
Analysis Report (UFSAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
NRC Bulletin 96-03 requested BWR [Boiling Water Reactor] owners
implement appropriate measures to minimize the potential clogging of
the ECCS suppression chamber strainers by potential debris generated
by a LOCA [loss-of-coolant-accident]. TVA's [Tennesse Valley
Authority's] proposed resolution of this issue for BFN takes credit
for containment overpressure to maintain adequate ECCS pump NPSH.
Containment overpressure is a result of the conditions which will
exist in the containment following the pipe break inside
containment. Therefore, the use of containment overpressure in the
analysis of the consequences of the LOCA does not affect the
precursors for the LOCA, nor does it affect the precursors for any
other accident or transient analyzed in Chapter 14 of the BFN
Updated Final Safety Analysis Report (UFSAR). Therefore, there is no
increase in the probability of any accident previously evaluated.
The worst radiological consequences for the design basis
accidents analyzed in UFSAR Chapter 14 are a result of a
circumferential break of one of the recirculation loop lines inside
containment. The analysis of the radiological consequences of this
event assumes a two percent per day leakage from the containment.
The results of this analysis are presented in Section 14.6.3 of the
UFSAR and indicate substantial margin when compared to 10 CFR Part
100 limits.
[[Page 50940]]
The radiological consequences of the design basis accident are
not increased by taking credit for the post-LOCA suppression chamber
airspace pressure. Without loss of primary containment, no mechanism
exists to increase the accident consequences since current leakage
bounds this condition. The initial analysis does not assume
differential pressure between the drywell and the suppression
chamber even though one exists due to the equilibrium conditions
caused by the suppression chamber airspace temperature.
Specifically, the suppression chamber airspace pressure credited in
the ECCS pump NPSH analyses is provided by an increase in
suppression chamber vapor pressure due to the increased pool
temperature, including an evaluation of the effects of containment
initial conditions and leakage.
By crediting the post-LOCA suppression chamber airspace pressure
in the calculation of NPSH, no requirement is created to purposely
maintain a higher containment pressure than would otherwise occur;
no requirement is incurred to delay operating containment heat
removal equipment; no requirement is incurred to deliberately
continue any condition of high containment pressure in order to
maintain adequate NPSH; and no requirement is incurred for the
purposeful addition of nitrogen into the containment to increase the
available pressure. Therefore, the proposed amendment does not
involve a significant increase in the consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed use of the post-LOCA suppression chamber airspace
pressure in the calculation of NPSH for the ECCS pumps does not
introduce any new modes of plant operation or make physical changes
to plant systems. Rather, the post-LOCA suppression chamber airspace
pressure is a byproduct of the conditions that will exist in the
containment after a line break inside containment. Therefore,
crediting the post-LOCA suppression chamber airspace pressure in the
calculation of NPSH does not create the possibility of a new or
different accident.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The integrity of the primary containment and the operation of
the ECCS systems limit the offsite doses to values less than those
specified in 10 CFR 100 in the event of a reactor coolant system
line break inside primary containment. In order for the ECCS pumps
to meet their design basis performance requirements, the NPSH
available to the pumps throughout the duration of the accident
response must meet their specific NPSH requirements. Excess NPSH
margin will not improve the performance of the ECCS pumps.
The post-LOCA suppression chamber airspace pressure is a
byproduct of the conditions that will exist in the containment after
a line break inside containment. The credit taken for this pressure
in ECCS NPSH analyses has been performed in such a manner as to
assure that the actual containment overpressure will always exceed
the value assumed in the analyses. The NPSH margin will exceed that
credited in the NPSH analyses and ECCS pump performance will meet
applicable requirements. Therefore, the proposed license amendment
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
its review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, 405 E.
South Street, Athens, Alabama 35611.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: August 5, 1998 (TS 98-008).
Description of amendment request: The proposed amendment would
revise the Watts Bar Nuclear Plant (WBN) Technical Specifications (TS)
and associated TS Bases to allow up to 4 hours to make the residual
heat removal suction relief valve available as a cold overpressure
mitigation (COMS) relief path. This condition will be applicable when
entering Mode 4 from Mode 3 during a plant shutdown.
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 4 hour allowance to place the RHR [residual heat removal]
relief valve in service in the proposed TS change is bounded by the
current COMS TS. The COMS TS currently allows cooldown of the unit
while in Mode 4 with only one operable relief path for up to 7 days.
Operation in this condition is allowed by Action E.1 of LCO
[limiting condition for operation] 3.4.12. The 7 day completion time
considers the facts that only one of the RCS [reactor coolant
system] relief valves is required to mitigate an overpressure
transient and that the likelihood of an active failure of the
remaining relief path during this 7 day time period is very low.
Thus a failure of the single available relief path concurrent with
an overpressurization event during the proposed 4 hour time period
for alignment and preparation of the RHR system for service is more
remote. Therefore, the proposed TS change does not increase the
probability of an accident previously evaluated. Further, this
change does not result in hardware or procedural changes which will
affect the probability of the occurrence of an accident. Considering
this, the proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Action E.1 of LCO 3.4.12 addresses a condition where one relief
path is inoperable while in Mode 4. The completion time for Action
E.1 is 7 days. The 4 hour period of operation in Mode 4 that will be
allowed by the addition of Note 4 to the Applicability statement of
LCO 3.4.12 is well within the bounds of the analysis for operation
allowed by Action E.1. This 4 hour time allowance for placement of
the RHR suction relief valve in service therefore, does not cause
the initiation of any accident nor create any new [credible]
limiting failure for safety-related systems and components. Since
the 4 hour period is only a fraction of the 7 day time period
previously authorized for operation with only a single relief path,
it is not probable that an accident different from those previously
evaluated will be created. Therefore, the change has no adverse
effect on the ability of the safety-related systems to perform their
intended safety functions.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The Technical Specifications currently allow one of the two
required relief valves to be unavailable for 7 days (Condition E of
LCO 3.4.12) while in Mode 4. In this condition (one of the two
relief valves inoperable), the proposed change would permit a mode
change from Mode 3 to Mode 4 while providing 4 hours to place the
RHR system into service. Consequently, this change does not reduce
the margin of safety since the probability of an event occurring
during the 4 hour period is less than the probability of an event
occurring during the 7 days permitted by Action E.1. Considering
this, the proposed change does not significantly reduce the margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET I0H, Knoxville, Tennessee 37902.
[[Page 50941]]
NRC Project Director: Frederick J. Hebdon.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: August 6, 1998 (TS 98-007).
Description of amendment request: The proposed amendment would
revise the Watts Bar Nuclear Plant (WBN) Technical Specifications (TS)
and associated TS Bases to clarify the intent of the surveillance
requirements (SRs) for turbine driven auxiliary feedwater (AFW) pump.
The proposed revision would allow three SRs to be performed prior to
achieving 1092 psig in the steam generator (SG).
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 license amendment would revise the subject TDAFWP
[turbine driven auxiliary feedwater pump] TS surveillance
requirements to be consistent with the intent of the current
Westinghouse MERITS TS, NUREG 1431, Revision 1. TS 3.3.2 and 3.7.5
would be revised to permit testing of the TDAFWP at SG pressures
less than the no-load pressure of 1092 psig [pounds per square inch-
gauge]. Under these conditions, the AFW system will continue to
satisfy requirements for the analyzed design basis accidents and
anticipated operational transients dependent on AFW. The design
basis for the AFW system and specifically the TDAFWP will be
maintained such that the AFW system and its equipment will continue
to perform its safety functions because the TDAFWP test will
demonstrate, on recirculation flow near pump shutoff head, the
ability to deliver full rated flow to the SGs. The proposed TS
change does not result in any modifications to the plant and does
not alter any fission barriers or challenge fuel integrity, nor are
other safety systems degraded by the subject change. Potential
radiological releases are not impacted by this TS change and there
are no new release pathways created. Therefore, the proposed TS
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated for WBN.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS change does not result in a modification to the
plant and has no adverse affect on the ability of any safety-related
system to perform its intended function. No new accident scenarios
are created and no new failure modes/mechanisms or limiting single
failures are created as a result of the proposed change that would
prevent the AFW system from performing its safety functions. A lower
test pressure than the current value of 1092 psig would have an
insignificant impact on the stroke time of the Terry turbine trip
and throttle valve, 1-FCV-1-51. Therefore, the proposed TS change
will not result in any new or different kind of accident from any
accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
This TS change does not change an acceptance limit nor does it
reduce a margin of safety associated with the acceptance criteria
for any WBN accident. The safety analyses performed for WBN is not
based on the SG pressure at which the TDAFWP test is conducted.
Specifically, the proposed TS change clarifies requirements for the
TDAFW pump testing consistent with industry practice. The capability
of the SRs to detect any degradation to the TDAFWP is unaffected.
The capability of the SRs to demonstrate automatic start and
adequate response time of the TDAFWP is not adversely impacted. The
test remains a requirement of the TS, but clarifies that the test
may be conducted at a SG pressure less than no-load conditions. The
proposed TS change does not reduce the margin of safety limits
established to protect any fission product barriers. Therefore, the
proposed TS change will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: May 8, 1998, as supplemented on July 10,
1998.
Description of amendment request: The licensee proposed to change
the maximum torus water temperature during normal operation from 100
deg.F to 90 deg.F; limit the temperature during testing to 100 deg.F
for no more than 24 hours; and, should temperature exceed 110 deg.F
prevent operation until the temperature is reduced to below 90 deg.F
(changed from 100 deg.F). Basis for proposed no significant hazards
consideration determination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(a) The proposed change to decrease the normal operating
suppression pool temperature limit from 100 deg.F to 90 deg.F will
assure that the consequences of accidents previously evaluated will
not be significantly increased.
A reduction in the normal operating suppression pool temperature
limit provides more margin for the suppression pool as a heat sink
to absorb energy from the reactor vessel following an accident. The
effect of higher calculated suppression pool temperatures following
an accident as a result of the effect of increased feedwater
addition and decreased [residual heat removal] RHR heat exchanger
heat removal does not affect the consequences of accidents
previously evaluated.
Certain types of Mark I containment loading conditions are
increased at lower suppression pool temperatures, but since the
analysis of Mark I loads for Vermont Yankee was based on initial
suppression pool temperatures between 70 deg.F and 90 deg.F, the
proposed decrease in the normal operating limit to 90 deg.F will
not affect the consequences of those particular events.
(b) The proposed change to decrease the normal operating
suppression pool temperature limit from 100 deg.F to 90 deg.F will
not affect the probability of accidents occurring. The accidents and
transients described in the [final safety analysis report] FSAR are
initiated by failures of components which are not in contact with
the suppression pool water, therefore a change in the suppression
pool temperature will have no affect on the probability of those
accidents occurring.
(c) The proposed change to restrict operation during testing
that adds heat to the suppression pool to no more than 24 hours
while above the normal operating temperature limit will have no
affect on the consequences of accidents previously evaluated since
accidents are not assumed to be initiated during these modes of
operation. This assumption is made in order to assure that plants
have testing flexibility at power. In addition to the time limit
placed on pool temperature, the plant enters the appropriate
limiting condition for operation whenever the RHR system is placed
in the suppression pool cooling mode during power operation.
(d) The proposed change to restrict operation during testing
that adds heat to the suppression pool to no more than 24 hours
while above the normal operating temperature limit will have no
affect on the probability of an accident occurring. The accidents
and transients described in the FSAR are initiated by failures of
components which are not in contact with the suppression pool water,
therefore a change in
[[Page 50942]]
the duration of time at any particular suppression pool temperature
will have no affect on the probability of those accidents occurring.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment, will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change to decrease the normal operating suppression
pool temperature limit from 100 deg.F to 90 deg.F does not change
any accident initiators or the types of accidents analyzed. No new
modes of equipment operation or physical plant equipment
modifications are proposed. The change in predicted peak suppression
pool temperature results from more conservatively calculating the
effects of currently analyzed accidents. Therefore this change will
not create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change to restrict operation during testing that
adds heat to the suppression pool to no more than 24 hours with
water temperature above the normal operating temperature limit will
allow for appropriate testing of safety related equipment to ensure
operability. This testing allowance does not create any new
initiating events or transients and does not involve any new modes
of operation. Therefore, this change does not create the possibility
of a new or different kind of accident from those previously
evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment, will not involve a
significant reduction in a margin of safety.
The proposed change to decrease the normal operating suppression
pool temperature limit from 100 deg.F to 90 deg.F assures that the
suppression pool can adequately perform its safety function without
a significant decrease in the margin of safety. Each of the
accidents affected by suppression pool temperature have been
evaluated. The evaluation showed that a higher peak suppression pool
temperature was predicted based on analysis assumptions that are
more conservative tha[n] those used in the current FSAR, but that
the increase in peak temperature does not have a[n] impact on
containment loads and equipment operability. The principal effect of
an increase in peak suppression pool temperature is the reduction of
[net positive suction head] NPSH margin for the low pressure
[emergency core cooling system] ECCS pumps. Operator action is
credited in throttling the ECCS pump flow rates after 10 minutes for
the most limiting scenarios in order to assure that available NPSH
exceeds required NPSH. Operator action after 10 minutes is
consistent with Vermont Yankee's design basis and Emergency
Operating Procedures. The proposed reduction in the normal operating
suppression pool temperature limit from 100 deg.F to 90 deg.F will
provide more time for operators to take actions, if required.
Operation of the facility in accordance with the proposed change
to restrict operation during testing that adds heat to the
suppression pool to no more than 24 hours while above the normal
operating temperature limit will not involve a significant reduction
in a margin of safety because it restricts the amount of time that
the facility can be operated at a suppression pool temperature above
the normal operating limit.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Project Director: Cecil O. Thomas.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: October 10, 1996.
Description of amendment request: The amendment would add to the
WNP-2 Facility Operating License No. NPF-21, the authority to store on
the WNP-2 site, byproduct, source, and special nuclear materials
currently addressed by the WNP-1 Materials License 46-17694-02.
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:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed amendment does not remove or modify existing
requirements or safety limits. The requirements of the [Atomic
Energy] Act and 10 CFR Parts 30, 40, and 70 will govern storage of
sealed byproduct and neutron sources. Operation of WNP-2 requires
possession and use of similar materials, and control of such
materials is currently being exercised pursuant to the requirements
of the Act and 10 CFR Parts 30, 40, and 70. The additional inventory
of radioactive materials is a very small percentage of that already
being controlled under Operating License NPF-21. Stored materials
such as those proposed are not assumed as an initiator of, or
contributor to, a previously analyzed accident. Consequently, the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
The requirements of the Act and 10 CFR Parts 30, 40, and 70 will
govern storage of sealed byproduct and neutron sources. These
materials will be stored indefinitely, and will not be put to active
use. Operation of WNP-2 requires possession and use of similar
materials, and control of such materials is currently being
exercised pursuant to the requirements of the Act and 10 CFR Parts
30, 40, and 70. Consequently, the proposed amendment does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The additional inventory of radioactive materials included in
sealed byproduct and neutron sources to be stored is a very small
percentage of that already being controlled under Operating License
NPF-21. The storage of materials does not impact the normal or
emergency operation of the plant. No change to the manner in which
the plant is operated is proposed. No modification to the facility
is proposed. Consequently the proposed amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Project Director: William H. Bateman.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: October 15, 1996, as supplemented by
letter dated December 4, 1997.
Description of amendment request: This amendment would modify the
secondary containment and standby gas treatment system (SGTS) technical
specifications to more accurately reflect the existing design by
revising the secondary containment and SGTS surveillance requirements
to reflect a revised flow rate, revising the secondary containment
integrity surveillance requirements by establishing an acceptable
operating region as a function of secondary containment differential
pressure and SGTS system
[[Page 50943]]
flow, and deleting the existing requirement to maintain the secondary
containment at greater than or equal to 0.25 inch of vacuum water gauge
at all times.
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?
Secondary containment and the Standby Gas Treatment (SGT) system
are not initiators or precursors to any accident. The SGT system
acts as part of secondary containment to minimize and control
airborne radiological releases from the plant following a design
basis accident. Therefore, operation of WNP-2 in accordance with the
proposed changes will not cause a significant increase in the
probability of an accident previously evaluated.
The proposed amendment to the Technical Specifications impacts
the capability to demonstrate that the secondary containment and SGT
system designs will maintain radioactive releases within 10 CFR 100
guidelines and 10 CFR 50, Appendix A, General Design Criteria 19
limits. As a result, a new (current) design basis accident dose
analysis was performed using the source term criteria outlined in
Regulatory Guide 1.3, ``Assumptions Used for Evaluating the
Potential Radiological Consequences of a Loss of Coolant Accident
for Boiling Water Reactors,'' to evaluate the proposed changes. The
new analysis provides a conservative representation of the timing
and release of radioactivity during a design basis accident.
The proposed amendment also deletes the normal (nonsafety-
related) secondary containment ventilation system surveillance
requirement to verify every 24 hours that the pressure within
secondary containment is less than or equal to 0.25 inch of vacuum
water gauge. This surveillance requirement is not necessary as
current Technical Specification Limiting conditions for Operation
(LCOs) as well as the WNP-2 Final Safety Analysis Report (FSAR)
adequately address secondary containment integrity requirements and
ensure secondary containment effluent is monitored. Deletion of the
surveillance requirement has no impact on the secondary containment
drawdown analysis or the design basis dose analysis. Thus, the
analyses assumptions and conclusions remain valid.
The secondary containment and SGT system designs must
accommodate a post-accident single failure and remain operable. In
addition, certain plant specific parameters, such as SGT capacity,
secondary containment in-leakage, outside meteorological conditions,
secondary containment heat loads, available cooling capacity,
emergency diesel start time and loading sequence, and drawdown time
for secondary containment must be considered in the design analyses
and dose assessments. The current design in conjunction with an
assumed secondary containment leakage of 2240 cfm and a drawdown
time of 20 minutes provide assurance that the radiological doses for
a design basis accident are maintained below the 10 CFR 100
guidelines and 10 CFR 50, Appendix A, General Design Criteria 19
limits.
The dose analysis supporting the proposed amendment to the
Technical Specifications includes analytical changes to the SGT flow
rate, secondary containment drawdown time, mixing, and bypass
leakage, and established a 95% meteorological basis. These
analytical changes, in combination, result in a calculated increase
in the offsite thyroid dose values and a decrease in the whole body
dose values. Although the calculated offsite thyroid dose values are
higher than previously calculated, they remain within the 10 CFR 100
guidelines and 10 CFR 50, Appendix A, General Design Criteria 19
limits. In accordance with Standard Review Plan (NUREG-0800),
Section 15.6.5, ``Loss-of-Coolant Accidents Resulting From a
Spectrum of Postulated Piping Breaks Within the Reactor Coolant
Pressure Boundary,'' the radiological consequences of a design basis
accident are considered acceptable if they are within the guidelines
of 10 CFR 100. Since the offsite thyroid dose values remain within
these acceptance criteria, and since there is no increase in the
control room thyroid dose values or any of the whole body dose
value, the changes are considered acceptable and operation of WNP-2
in accordance with the proposed amendment to the Technical
Specifications will not cause a significant increase in the
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Secondary containment and the SGT system are not initiators or
precursors to any accident. The SGT system acts as part of secondary
containment to minimize and control airborne radiological releases
from the plant following a design basis accident.
The dose analysis supporting the proposed amendment to the
Technical Specifications includes analytical changes to the SGT flow
rate, secondary containment drawdown time, mixing, and bypass
leakage, and establish a 95% meteorological basis. These analytical
changes do not alter any safety-related equipment or functions or
create any new failure modes. The changes will improve the
capability of secondary containment and the SGT system to mitigate
the consequences of a design basis accident by ensuring that
secondary containment pressure can be drawn down from 0 inches water
gauge to at least 0.25 inch of vacuum water gauge during the most
adverse environmental conditions. The proposed changes reflect
consideration of SGT capacity, secondary containment in-leakage,
outside meteorological conditions, secondary containment heat loads,
available cooling capacity, emergency diesel start time and loading
sequence, and drawdown time for the limiting secondary containment
elevation. Required instrumentation have been evaluated to ensure
proper operation under normal and accident environmental conditions,
including but not limited to pressure, humidity, seismic,
temperature, and radiation. The evaluation method is based on
American National Standards Institute/Instrument Society of America
(ANSI/ISA) Standard S67.04-1988, ``Setpoints for Nuclear Safety-
Related Instrumentation,'' and guidelines in ISA draft Recommended
Practice RP67.04, ``Methodologies for the Determination of Setpoints
for Nuclear Safety-Related Instrumentation.''
The proposed amendment to the Technical Specification does not
change plant equipment or functions, but serves to clarify and
credit existing design features. Fault tree and single failure
analyses were performed to ensure that the SGT system design,
including the equipment and components, credited in the licensing
basis for the proposed amendment meet the single failure criteria
for credible failure modes. The proposed amendment also deletes the
normal (nonsafety-related) secondary containment ventilation system
surveillance requirement to verify every 24 hours that the pressure
within secondary containment is less than or equal to 0.25 inch of
vacuum water gauge. Deletion of this surveillance requirement does
not invalidate existing analyses or change plant equipment or
functions. Thus, no new failure modes are created.
Based on equipment failure and qualification analyses performed
and the above conclusions, the proposed amendment to the Technical
Specifications does not change any safety-related equipment or
functions, or create any new failure modes. Therefore, operation of
WNP-2 in accordance with the proposed amendment to the Technical
Specifications will 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?
Consistent with the current Bases for the Technical
Specifications and the WNP-2 FSAR, secondary containment and the SGT
system act to minimize and control airborne radiological releases
from the plant to within 10 CFR 100 guidelines and 10 CFR 50,
Appendix A, General Design Criteria 19 limits following a design
basis accident.
The proposed amendment to the Technical Specifications impacts
the capability to demonstrate that the secondary containment and SGT
system designs will maintain radioactive releases within 10 CFR 100
guidelines and 10 CFR 50, Appendix A, General Design Criteria 19
limits. As a result, a new (current) design basis accident dose
analysis was performed using the source term criteria outlined in
Regulatory Guide 1.3 to evaluate the proposed changes. The new
analysis provides a conservative representation of the timing and
release of radioactivity during a design basis accident.
The proposed amendment also deletes the normal (nonsafety-
related) secondary containment ventilation system surveillance
requirement to verify every 24 hours that the pressure within
secondary containment is less than or equal to 0.25 inch of vacuum
water gauge. This surveillance requirement is
[[Page 50944]]
not necessary as current Technical Specification LCOs as well as the
WNP-2 FSAR adequately address secondary containment integrity
requirements and ensure secondary containment effluent is monitored.
Deletion of the surveillance requirement has no impact on the
secondary containment drawdown analysis or the design basis dose
analysis. Thus, it follows that deletion of the surveillance
requirement will not impact the offsite and control room dose safety
margins established by these analyses.
The dose analysis includes analytical changes which increase SGT
system flow rate and secondary containment drawdown time, credit
mixing within secondary containment, increase bypass leakage, and
establish a 95% meteorological basis. The combined effect of these
analytical changes results in an increase in the calculated offsite
thyroid dose values. The calculated control room thyroid dose values
and all of the whole body dose values are shown to decrease.
Although the new thyroid dose values are higher than previously
calculated, they remain within the 10 CFR 100 guidelines and 10 CFR
50, Appendix A, General Design Criteria 19 limits. The calculated
thyroid dose values at the plant exclusion area boundary (EAB) (1.2
miles) increased from 72 Rem to 114.2 Rem and the calculated thyroid
dose at the low population zone (LPZ) (3 miles) increased from 251
Rem to 275.6 Rem.
The LPZ is defined as all land within a 3 mile radius of the
plant site and 0 persons reside within this area. The nearest
residence is 4.1 miles from the plant site. There are no schools or
hospitals within 5 miles of the plant site and the nearest
population center is at 12 miles. Considering the low population
density in the area immediately surrounding the plant site, the
increase in thyroid dose will have a small impact on the health and
safety of the public.
Since the offsite thyroid dose values remain within the 10 CFR
100 guidelines and 10 CFR 50, Appendix A, General Design Criteria 19
limits, and since there is a small impact on the health and safety
of the public, the increase in the offsite thyroid dose values are
considered acceptable and operation of WNP-2 in accordance with the
proposed amendnment to the Technical Specifications will not cause a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Project Director: William H. Bateman.
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.
Illinois Power Company, Docket No. 50-461, Clinton Power Station,
DeWitt County, Illinois Date of Application for Amendment: August 24,
1998
Brief description of amendment request: The proposed amendment
concerns the ``ready-to-load'' requirement for the Division 3 diesel
generator (DG). The Division 3 DG requires operator action to reset the
mechanical governor to meet the ``ready-to-load'' requirement.
Date of publication of individual notice in Federal Register:
September 10, 1998 (63 FR 48529).
Expiration date of individual notice: October 13, 1998.
Local Public Document Room location: Vespasian Warner Public
Library, 310 N. Quincy Street, Clinton, IL 61727.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 28, 1998.
Brief description of amendment request: The proposed amendment
would modify Technical Specification 4.0.5 to state that the inservice
testing requirement for exercise testing in the closed direction for
specified Unit 1 containment isolation valves shall not be required
until the next plant shutdown to Mode 5 of sufficient duration to allow
the testing or until the next refueling outage scheduled in March 1999.
Date of publication of individual notice in Federal Register:
September 9, 1998 (63 FR 48254)
Expiration date of individual notice: September 24, 1998.
Local Public Document Room location: Wharton County Junior College,
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company, Toledo Edison Company Docket No. 50-440, Perry Nuclear Power
Plant, Unit 1, Lake County, Ohio
Date of amendment request: June 30, 1998.
Description of amendment request: The proposed amendment would
transfer operating authority for the Perry Nuclear Power Plant, Unit
No. 1, from The Cleveland Electric Illuminating Company and Centerior
Service Company to a new operating company, called the FirstEnergy
Nuclear Operating Company. The proposed action has been submitted
pursuant to 10 CFR 50.80 and 10 CFR 50.90.
Date of publication of individual notice in Federal Register:
August 4, 1998 (63 FR 41600).
Expiration date of individual notice: September 3, 1998.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: June 29, 1998, as supplemented
July 14, 1998.
Brief description of amendment request: This amendment would
reflect the approval of the transfer of the authority to operate Davis-
Besse Nuclear Power Station, Unit 1, under the license to a new
company, FirstEnergy Nuclear Operating Company.
Date of publication of individual notice in Federal Register:
August 4, 1998.
Expiration date of individual notice: September 3, 1998.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
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
[[Page 50945]]
complies with the standards and requirements of the Atomic Energy Act
of 1954, as amended (the Act), and the Commission's rules and
regulations. The Commission has made appropriate findings as required
by the Act and the Commission's rules and regulations in 10 CFR Chapter
I, which are set forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: October 31, 1996.
Brief description of amendment: This amendment changes Technical
Specification 3/4.7.5 by reducing the maximum allowable water
temperature for the Ultimate Heat Sink from 95 deg.F to 94 deg.F and
increasing the minimum main reservoir level from 205.7 feet mean sea
level to 215 feet mean sea level.
Date of issuance: September 8, 1998.
Effective date: September 8, 1998.
Amendment No: 80.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64382).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: May 16, 1997, as supplemented
June 29, 1998. The June 29, 1998, supplemental letter provided
clarifying information only, and did not change the initial no
significant hazards consideration determination.
Brief description of amendment: This amendment changes Technical
Specification 3/4.6.2.3 by reducing the Containment Fan Coolers cooling
water flow rate requirement from 1425 gallons per minute (gpm) to 1300
gpm.
Date of issuance: September 8, 1998.
Effective date: September 8, 1998.
Amendment No: 81.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 25, 1998 (63 FR
14485).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of application for amendments: January 14, 1998, as
supplemented by letter dated July 17, 1998.
Brief description of amendments: The amendments change the
Braidwood, Unit 1, Technical Specification limits on Reactor Coolant
System Dose Equivalent Iodine-131 from 0.35 microcuries/gram to 0.05
microcuries/gram for the remainder of Cycle 7.
Date of issuance: September 3, 1998.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 95 and 95.
Facility Operating License Nos. NPF-72 and NPF-77: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 11, 1998 (63 FR
11914). The July 17, 1998, submittal provided additional clarifying
information that did not change the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
September 3, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wilmington Public Library, 201
S. Kankakee Street, Wilmington, Illinois 60481.
Duke Energy Corporation, et al., Docket No. 50-413, Catawba Nuclear
Station, Unit 1, York County, South Carolina
Date of application for amendment: August 6, 1998.
Brief description of amendment: The amendment deletes Surveillance
Requirement 4.8.1.1.2.i.2, regarding diesel fuel oil system pressure
testing, from the unit Technical Specifications for Unit 1 on the basis
that the staff had previously approved alternative surveillance based
on Code Case N-498-1 of the American Society of Mechanical Engineers.
Date of issuance: September 9, 1998.
Effective date: As of the date of issuance.
Amendment No.: 171.
Facility Operating License No. NPF-35: The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. (63 FR 43962 dated August 17, 1998). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by September 16, 1998, but indicated that if the Commission
makes a final no significant hazards consideration determination, any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and a final no significant hazards consideration
determination are contained in a Safety Evaluation dated September 9,
1998.
Attorney for licensee: Paul R. Newton, Legal Department (PB05E),
Duke Energy Corporation, 422 South Church Street, North Carolina.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
[[Page 50946]]
Duke Energy Corporation (DEC), 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: August 14, 1998.
Brief description of amendments: The amendments revise Technical
Specification Section 4.6.5.1.b.2 regarding surveillance requirements
for the ice condenser. One current requirement specifies that a visual
inspection of flow passages be performed once per 9 months to ensure
that there is no significant ice and frost accumulation (less than 0.38
inch). DEC proposed to relax the visual inspection frequency of the
lower plenum support structures and turning vanes to once per 18
months, while the remaining parts of the ice condenser will continue to
be inspected at 9-month intervals.
Date of issuance: September 10, 1998.
Effective date: As of the date of issuance.
Amendment Nos.: Unit 1--172; Unit 2--163.
Facility Operating License Nos. NPF-35 and NPF-52: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. (63 FR 45872 dated August 27, 1998). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by September 28, 1998, but indicated that if the Commission
makes a final no significant hazards consideration determination, any
such hearing would take place after issuance of the amendments.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and a final no significant hazards consideration
determination are contained in a Safety Evaluation dated September 10,
1998.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
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: August 14, 1998.
Brief description of amendments: The amendments revise Surveillance
Requirement 4.6.5.1.b.3 of the Technical Specifications, relaxing the
visual inspection interval of the ice condenser lower plenum and
turning vanes from the current 9-month to 18-month intervals.
Date of issuance: September 10, 1998.
Effective date: As of the date of issuance.
Amendment Nos.: Unit 1-180; Unit 2-162.
Facility Operating License Nos. NPF-2 and NPF-8: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. (63 FR 45870 dated August 27, 1998). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by September 28, 1998, but indicated that if the Commission
makes a final no significant hazards consideration determination, any
such hearing would take place after issuance of the amendments.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and a final no significant hazards consideration
determination are contained in a Safety Evaluation dated September 10,
1998.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: March 11, 1993, as supplemented
August 26, October 26, November 29, and December 6, 1993, October 3,
1995, February 27, May 2, and September 3, 1997, and May 7, 1998.
Brief description of amendments: The amendments completely revise
the current Technical Specifications related to the electrical
distribution system and incorporate new requirements for system
operation, limiting conditions for operation, and surveillance
requirements.
Date of Issuance: September 4, 1998.
Effective date: As of the date of issuance, to be implemented
coincident with implementation of the Improved Technical
Specifications.
Amendment Nos.: Unit 1-232; Unit 2-232; Unit 3-231.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 3, 1997 (62 FR
63975).
The May 2, 1997, and May 7, 1998, letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 4, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: April 28, 1998.
Brief description of amendment: The amendment proposed to revise
the Improved Technical Specification 5.6.2.8 to change the scope and
frequency of volumetric and surface inspections for the reactor coolant
pump flywheels. The amendment approves the requested change to reflect
the frequency and scope of these inspections as specified in Topical
Report WCAP-14535A.
Date of issuance: August 31, 1998.
Effective date: August 31, 1998.
Amendment No.: 170.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40555)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application foramendment: June 29, 1998, as supplemented
July 27, 1998.
Brief description of amendment: The amendment reduces the scope of
a
[[Page 50947]]
previous amendment request dated February 22, 1996. It retains the
provision to delete the requirement that the biennial inspection of the
emergency diesel generators (EDGs) be performed during shutdown,
permits skipping diesel starting battery capacity test for recently
installed batteries, and increases the minimum loading during diesel
testing from 20% to 80%. In addition, there are wording changes to
enhance clarity and a typograhpical error is corrected.
Date of Issuance: September 8, 1998.
Effective date: September 8, 1998, to be implemented within 30
days.
Amendment No.: 197.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40556). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated September 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room Location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
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: February 22, 1996.
Brief description of amendments: The amendments revise the
Technical Specifications to reference NRC Regulatory Guide 1.9,
Revision 3, rather than NRC Regulatory Guide 1.108, Revision 1, for the
determination of a valid diesel generator test.
Date of issuance: September 2, 1998.
Effective date: September 2, 1998, with full implementation within
45 days.
Amendment Nos.: 222 and 206.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15990).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
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: June 10, 1998.
Brief description of amendments: The amendments defer the
implementation date of Amendments Nos. 216/200 to become effective when
modifications are completed but not later than December 31, 2000.
Date of issuance: August 31, 1998.
Effective date: August 31, 1998, with full implementation not later
than December 31, 2000.
Amendment Nos.: 221 and 205.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the licenses.
Date of initial notice in Federal Register: July 31, 1998 (63 FR
40940).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: June 22, 1995, as supplemented
on May 13, 1998.
Brief description of amendments: The amendments revise Technical
Specifications 3.4.1.4 and 3.9.8.2 by deleting footnotes and associated
information regarding service water system header operation to allow
residual heat removal system operation to be consistent with current
regulations and the Standard Technical Specifications--Westinghouse
Plants (NUREG-1431).
Date of issuance: September 8, 1998.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment Nos.: 214 and 194.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45183).
The May 13, 1998, letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination, and was within the scope of the original application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of application for amendments: July 22, 1998.
Brief description of amendments: The amendments revise the
technical specifications to extend the allowed outage time (AOT) for
off-site circuits and for the emergency diesel generator.
Date of issuance: September 9, 1998.
Effective date: September 9, 1998, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 2-141; Unit 3-133.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 31, 1998 (63 FR
40941).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear
Plant, Unit 3, Limestone County, Alabama
Date of application for amendment: June 2, 1995, revised March 6,
1997, as supplemented April 11, May 13, and August 20, 1997, and March
13, 1998. (TS-353).
Brief description of amendment: Revises Technical Specifications
(TS) to permit implementation of upgrade of power range neutron monitor
instrumentation. Other changes also have been incorporated to thermal
limits specifications to implement average power range monitor and rod
block monitor TS improvements, and maximum extended load line limit
analyses.
Date of issuance: September 3, 1998.
Effective date: September 3, 1998.
Amendment No.: 213.
Facility Operating License No. DPR-68: Amendment revises the TS. .
Date of initial notice in Federal Register: August 16, 1995 (60 FR
[[Page 50948]]
42609). The revision dated March 6, 1997; the proposal for the same
changes to be made to the Improved Standard TS format dated April 11,
1997; and the supplemental information dated May 13 and August 20,
1997, and March 13, 1998, did not affect the staff's original finding
of no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 3, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2. Hamilton County, Tennessee
Date of application for amendments: February 13, 1998 (TS 97-04).
Brief description of amendments: The amendments change the
Technical Specifications (TS) by relocating the snubber requirements
from Section 3.7.9 of the TS, and its bases, to the Sequoyah Nuclear
Plant Technical Requirements Manual. This change does not alter the
requirements for operability or surveillance testing of the snubbers.
This amendment also deletes License Condition 2.C.(19), for Unit 1
only. This condition is a one-time snubber-related action that was
completed and no longer needs to be included in the SQN Operating
License.
Date of issuance: August 28, 1998.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: Unit 1-235 ; Unit 2-225.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TS.
Date of initial notice in Federal Register: April 8, 1998 (63 FR
17235).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 28, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: December 23, 1997.
Brief description of amendment: This amendment revised Technical
Specification (TS) Section 4.4.5, ``Reactor Coolant System--Steam
Generators--Surveillance Requirements (SRs).'' SR 4.4.5.8 was modified
to provide flexibility in the scheduling of steam generator inspections
during refueling outages.
Date of issuance: September 2, 1998.
Effective date: September 2, 1998.
Amendment No.: 226.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4327).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: June 30, 1998.
Brief description of amendment: The licensee proposes to delete the
calibration requirements for emergency core cooling actuation
instrumentation--core spray (CS) subsystem and low pressure coolant
injection (LPCI) system auxiliary power monitor since the relays
operate from a switched input and functional testing is sufficient to
demonstrate the relay pickup/dropout capability.
Date of Issuance: September 1, 1998.
Effective date: September 1, 1998, to be implemented within 30
days.
Amendment No.: 162.
Facility Operating License No. DPR-28. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40563).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated September 1, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Dated at Rockville, Maryland, this 17th day of September 1998.
For The Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-25281 Filed 9-22-98; 8:45 am]
BILLING CODE 7590-01-P