[Federal Register Volume 64, Number 36 (Wednesday, February 24, 1999)]
[Notices]
[Pages 9183-9209]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-4391]
[[Page 9183]]
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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 January 30, 1999, through February 11, 1999.
The last biweekly notice was published on February 10, 1999.
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 March 26, 1999, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
[[Page 9184]]
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.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: January 28, 1999.
Description of amendment request: The H. B. Robinson, Unit No. 2,
Technical Specifications (TSs) are proposed to be changed to replace
and add analytical methodologies used to determine acceptable core
designs and provide inputs to methodologies that develop the core
operating limits in the Core Operating Limits Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes in a methodology have been previously
generically reviewed and approved for use by the NRC for determining
core neutronics design and gadolinimum oxide thermal conductivity.
Analyzed events are assumed to be initiated by the failure of plant
structures, systems, or components. The fuel design parameters
developed in accordance with the new methodologies are bounded by
the limitations in the NRC acceptance in its safety evaluations of
the new methodologies. The topical reports associated with the new
methodologies demonstrate that the integrity of the fuel will be
maintained during normal operations and that design requirements
preclude fuel rods containing gadolinium oxide from being limiting
in accident and related safety analyses. The proposed change does
not have a detrimental impact on the integrity of any plant
structure, system, or component. The proposed change will not alter
the operation of any plant equipment, or otherwise increase its
failure probability. Therefore, the probability of occurrence for a
previously analyzed accident is not significantly increased.
The consequences of a previously analyzed accident are dependent
on the initial conditions assumed for the analysis, the behavior of
the fuel during the analyzed accident, the availability and
successful functioning of the equipment assumed to operate in
response to the analyzed event, and the setpoints at which these
actions are initiated. The proposed changes to methodology continues
to meet applicable design and safety analyses acceptance criteria
for neutronics design analysis and gadolinimum oxide thermal
conductivity. The topical reports associated with the new
methodologies demonstrate that the integrity of the fuel will be
maintained as is assumed or is bounded initially in accident
analyses and that design requirements preclude fuel rods containing
gadolinimum oxide from being limiting in accident and related safety
analyses. The proposed change does not affect the performance of any
equipment used to mitigate the consequences of an analyzed accident.
As a result, no analyses assumptions are violated and there are no
adverse effects on the factors that contribute to offsite or onsite
dose as the result of an accident. The proposed change does not
affect setpoints that initiate protective or mitigative actions. The
proposed change ensures that plant structures, systems, or
components are maintained consistent with the safety analysis and
licensing bases. Based on this evaluation, there is no significant
increase in the consequences of a previously analyzed event.
Therefore, the proposed change does not involve any increase in
the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures, or components. The proposed changes in
methodology continue to meet applicable criteria for neutronics
design analysis and assure that design requirements preclude fuel
rods containing gadolinimum oxide from being limiting. The proposed
change does not involve a physical alteration of the plant other
than allowing for fuel design in accordance with NRC approved
methodologies. No new or different equipment is being installed. No
installed equipment is being operated in a different manner. There
is no alteration to the parameters within which the plant is
normally operated or in the setpoints that initiate protective or
mitigative actions. As a result no new failure modes are being
introduced. There are no changes in the methods governing normal
plant operation, nor are the methods utilized to respond to plant
transients altered. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The margin of safety is established through the design of the
plant structures, systems, and components, through the parameters
within which the plant is operated, through the establishment of the
setpoints for the actuation of equipment relied upon to respond to
an event, and through margins contained within the safety analyses.
The proposed change is to methodologies that continue to meet
applicable criteria for neutronics design analysis and continues to
assure that design requirements preclude fuel rods containing
gadolinimum oxide from being limiting. The proposed change does not
impact the condition or performance of structures, systems,
setpoints, and components relied upon for accident mitigation. The
proposed change does not significantly impact any safety analysis
assumptions or results. Therefore, the proposed change does not
result in a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 9185]]
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
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: Cecil B. Thomas.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of amendment request: November 25, 1998.
Description of amendment request: The proposed amendments would
revise Improved Technical Specifications 3.8.4 and 3.8.9 to support on-
line replacement of the Braidwood 125 Volt DC AT&T batteries with new
Charter Systems Inc. batteries.
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?
During the replacement of the existing batteries, a temporary
battery bank will provide the same function as the AT&T batteries
being removed. Even though this temporary battery will not be
seismically mounted, due to its location in the Turbine Building, it
is the safety related AT&T battery which was previously qualified
and used to perform this function on Unit 1.
While the temporary battery is being connected, the DC bus will
be supplied by the existing crosstie with Unit 1. Similar crosstie
conditions are allowed under the present Improved Technical
Specifications.
The DC system is normally supplied by the AC system through the
ESF [Engineered Safety Feature] battery charger. The essential
function of the DC system battery is to supply control power
necessary to start and load the Diesel Generators. Once the Diesel
Generators are on line, the DC system will be supplied via the
battery charger. However, the ESF batteries have been sized for one
hour to provide additional assurance that the critical DC loads are
available in the event of a loss of a battery charger.
During the 10 day Completion Time when the temporary battery and
the ESF charger are supporting the bus, the ability of that DC
Division to mitigate an event/accident is unchanged except for its
ability to cope with a seismic event. However, the probability of a
seismic event concurrent with the 10 day Completion Time is
extremely small. During a seismic event, one DC division may be
compromised, however, the unit has adequate DC power available in
the form of the other division to mitigate all Design Basis
accidents. This loss of one DC division is bounded by the loss of an
entire AC division, a condition which the plant is currently
evaluated to withstand.
During the 8 hour Completion Time to connect and disconnect the
temporary battery, there is no adverse impact on Unit 1. The
compensatory measures to manually open the crosstie will ensure the
Unit 1 DC battery can supply its required loads for the entire one
hour duty cycle. The Unit 2 DC bus, which is crosstied, will be de-
energized in the event of a Unit 2 accident based on the
compensatory measures. This action would only be required if the
associated Diesel Generator were to fail to re-energize its
associated charger. This condition is consistent with the other
crosstie scenarios currently permitted by the Technical
Specifications. Thus, the 8 hour Completion Time is consistent with
the two hour Completion Time with respect to the ability to safely
shutdown the Unit. Only the duration of the Completion Time is
different.
Based on the above, the overall design, function, and operation
of the DC system and equipment has not been significantly modified
by these changes. The proposed changes do not affect any accident
initiators or precursors and do not alter the design assumptions for
the systems or components used to mitigate the consequences of an
accident as analyzed in UFSAR Chapter 15.
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?
During the replacement of the existing batteries, a temporary
battery bank will provide the same function as the batteries being
removed. Even though this temporary battery is not seismically
mounted, it is the safety related AT&T battery which was previously
qualified and utilized to perform this function on Unit 1. Because
this temporary battery is identical to the battery that is currently
installed, and will be connected and used in the same way, no new
electrical or functional failure modes are created.
The temporary battery will be located in the turbine building,
which is non-seismic. The temporary battery will not be seismically
mounted. Thus, a seismic failure of the batteries is possible. Since
the temporary battery is located in the turbine building the
potential for battery failure to initiate an accident is not
present, and failure of the battery cannot create a different
response from any previously postulated accident.
Due to the location of the main generator in relationship to the
temporary batteries, a turbine blade failure would not hit the
battery unless it penetrated the turbine casing and ricocheted in
the direction of the battery, which is an unlikely scenario due to
the orientation of the temporary battery. Likewise, an unmitigated
Outside Containment Steam Line Break of either unit would be
interrupted by the successful closure of all MSIVs [Main Steam
Isolation Valves] thereby leaving the battery and the DC bus intact
and available. Also any affects of a postulated storm on the turbine
building have been previously addressed and would not change as a
result of the batteries being temporary located there.
While the temporary battery is being connected, the DC bus will
be supplied by the existing crosstie with Unit 1. To prevent any
occurrence on Unit 2 from adversely affecting Unit 1, this crosstie
will be manually disconnected based on specific criteria that may be
indicative of a Unit 2 accident (specifically a Unit 2 LOOP). Once
the crosstie is opened, the Unit 2 bus will be de-energized and the
other Unit 2 division will be required to mitigate the accident.
This loss of one DC division is bounded by the loss of one division
(AC or DC), a condition which the plant is currently evaluated to
withstand.
The DC system and its equipment will continue to perform the
same function and be operated in the same fashion. The proposed
changes do not introduce any new accident initiators or precursors,
or any new design assumptions for the systems or components used to
mitigate the consequences of an accident. Therefore, the possibility
of a new or different kind of accident from any accident previously
evaluated has not been created.
Therefore, this proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
During the replacement of the existing batteries, a temporary
safety related battery bank will perform the same function as the
batteries being removed. Even though this temporary battery is not
seismically mounted, it is the safety related battery which was
previously qualified and used to perform this function on Unit 1 and
is identical to the safety related battery that is currently
installed. Therefore, it has the same capacity, margin and
capability to fulfill the requirements of the Unit 2 DC bus as the
existing qualified battery. The proposed replacement activity will
not prevent the plant from responding to either a seismic event or
design basis accident. In both cases, the design mitigation
capability will be maintained. Due to the limited duration of the
activity and the planned contingency actions, a significant
reduction in the margin of safety will not result.
While the temporary battery is being connected, the DC bus will
be supplied by the existing crosstie with Unit 1. This condition is
currently allowed for a limited time by the Improved Technical
Specifications.
The inherent design conservatism of the DC system and its
equipment has not been altered. The DC system and its equipment will
continue to be operated with the same degree of conservatism.
Accordingly, there is no 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
[[Page 9186]]
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Wilmington Public Library, 201
S. Kankakee Street, Wilmington, Illinois 60481.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Project Director: Stuart A. Richards.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: December 29, 1998.
Description of amendment request: The proposed amendments would
revise the Technical Specification Tables 3.3.1-1 and 3.3.2-1, to
revise twelve Reactor Trip System and Engineered Safety Feature
Actuation System Allowable Values.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
These changes to the twelve AVs [Allowable Values] do not
involve an increase in the probability of an accident previously
evaluated. The AVs provide the basis for determining instrument
channel operability and do not change the system function, or
channel operation or calibration. Operation within the AV ensures
the instrument channel's ability to provide the required reactor
trip or engineered safety feature actuation signal during plant
operation. In all cases, the proposed changes only make the twelve
AVs more restrictive with respect to the current AVs, and do not
effect the response characteristics of the instrumentation because
actual trip setpoints are unchanged. There is no change being made
to the approved design, nor is there any operational change being
made which would increase the probability of occurrence of an
accident previously evaluated. The RTS [Reactor Trip System] and
ESFAS [Engineered Safety Feature Actuation System] systems which are
actuated by the corresponding instrumentation setpoints will operate
in the same manner as before and within their design limits.
These changes to the twelve AVs do not involve an increase in
the consequences of an accident previously evaluated. These changes
have no effect on plant operation. There is no physical or
operational change being made which would alter the sequence of
events, plant response, or assumptions or conclusions of the
affected analyses. The use of the AVs as a basis for determining
instrument or channel operability does not change system operation
or channel function. The proposed changes do not change the
established trip setpoints for these functions. No design analyses
have changed or will be affected. The twelve revised AVs are more
restrictive than the current AVs and continue to ensure that the
safety limits are not violated during anticipated transients, and
that the consequences of design basis accidents remain acceptable.
The change to the AVs does not degrade or prevent any actions from
taking place in response to an accident. The use of NRC approved or
endorsed methodology in developing the proposed AVs ensures that the
present analytical limits for all accidents will be maintained.
These proposed changes to the AVs for RTS and ESFAS instrumentation
will continue to ensure that the associated RTS trip or ESFAS
actuation signals will be generated when required within the bounds
of the plant safety analyses. There is no change in the type or
amount of any effluents released, and no change in either the onsite
or offsite dose consequences as a result of this change.
Therefore, based on this evaluation, this proposed amendment
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
These proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes to the twelve AVs for RTS and ESFAS
instrumentation will not affect the trip setpoints at which a
reactor trip or engineered safety feature actuation is initiated.
The trip setpoints contained in the Technical Requirements Manual
are not being changed and will continue to be maintained. The only
changes being made are to the AVs used as a basis for determining
instrument channel operability. Because the trip setpoints are
unchanged, RTS or ESFAS setpoint actuation is not affected by the
revised AVs.
An RTS trip or ESFAS actuation signal that may initiate between
its trip setpoint and the associated AV is acceptable because an
allowance has been made in the affected instrument uncertainty
calculation to accommodate this deviation. It allows for potential
drift while ensuring plant operation in a safe manner. Using this
methodology provides plant operational flexibility and yet remains
within the allowances accounted for in the various accident
analyses. No new equipment is being installed, and no installed
equipment is being operated in a new or different manner with these
twelve AV changes. The revised AVs do not alter the intended design
or operation of systems or instrument channels.
As no physical plant equipment changes are being made, no new
equipment failure modes are being introduced as a result of these
proposed changes. There is no change in plant operation that affects
previously evaluated failure modes and no change in plant response
to a transient condition. These changes do not represent a new
failure mode over what has been previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
There is no significant reduction in the margin of safety from
these proposed changes. These proposed changes move twelve AVs
closer to the trip setpoints compared to the existing AVs, which
increases the margin of safety. An RTS trip or ESFAS actuation
signal that may initiate between its trip setpoint and the
associated AV is acceptable because an allowance has been made in
the affected instrument uncertainty calculation to accommodate this
deviation. The revised AVs have been calculated using NRC approved
or endorsed methodology, which is consistent with existing safety
analyses that define the margin of safety. Safety analyses
assumptions and results are not affected.
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: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Project Director: Stuart A. Richards.
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: January 21, 1999.
Description of amendment request: This amendment request proposes
to relocate Technical Specification (TS) Section 3/4.6.I to the Updated
Final Safety Analysis Report (UFSAR) and plant procedures. TS Section
3/4.6.I contains reactor coolant chemistry limiting conditions for
operation (LCO) and surveillance requirements (SR) for conductivity,
chloride concentration and pH.
[[Page 9187]]
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 changes simplify the TS, meet regulatory
requirements for relocated TS's, and implement the recommendations
of the Commission's Final Policy Statement on TS improvements. The
Chemistry requirements will be relocated to the Updated Final Safety
Analysis Report (UFSAR) and to applicable station procedures. Future
changes to these requirements will be controlled by 10 CFR 50.59.
The proposed changes are administrative in nature and do not involve
any modification to any plant equipment or affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any previously
evaluated accident.
Consequently, this proposed amendment does not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety related system performs
its function. Therefore, this proposed TS amendment will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed amendment represents the relocation of current
requirements which are based on generic guidance or previously
approved provisions for other stations. The proposed changes are
administrative in nature and do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analysis. The proposed changes have been
evaluated and found to be acceptable for use at Quad Cities Nuclear
Power Station. Since the proposed changes are administrative in
nature, and are based on NRC accepted provisions which have been
adopted at other nuclear facilities, and maintain the necessary
levels of system reliability, the proposed 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: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Project Director: Stuart A. Richards.
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 amendment request: January 28, 1999.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to correct Surveillance
Requirement (SR) 3.7.13.4 and the associated Bases. This SR currently
is incorrect and does not reflect the Fuel Handling Ventilation Exhaust
System (FHVES) as designed. Specifically, the FHVES flow rate
requirement has been inadvertently stated at half the design value
(18,221 instead of 36,443 cfm [cubic feet per minute]). The proposed
amendments would only revise the SR to the correct design value; no
physical change to the FHVES design is involved.
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:
First Standard
Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Approval of this amendment will have no effect
on accident probabilities or consequences. The FHVES is not an
accident initiating system; therefore, there will be no impact on
any accident probabilities by the approval of this amendment. The
design of the system is not being modified by this proposed
amendment. The amendment merely aligns TS requirements with the
existing design and function of the system. Therefore, there will be
no impact on any accident consequences.
Second Standard
Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of NRC approval of this amendment request. No changes
are being made to the plant which will introduce any new accident
causal mechanisms. This amendment request does not impact any plant
systems that are accident initiators; neither does it impact any
accident mitigating systems.
Third Standard
Implementation of this amendment would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The performance of these fission
product barriers will not be impacted by implementation of this
proposed amendment. The FHVES is already capable of performing as
designed. No safety margins will be impacted.
Based upon the preceding analysis, Duke Energy has concluded
that the proposed amendment does not involve a significant hazards
consideration.
The staff reviewed the licensee's analysis, and agrees that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Project Director: Herbert N. Berkow.
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 amendment request: December 16, 1998, supplemented January
25, 1999.
Description of amendment request: The proposed amendments would
completely replace the High Pressure Injection (HPI) section of the
Improved Technical Specifications that were issued on December 16,
1998. The proposed changes would: (1) expand the applicability for the
requirements regarding the third HPI pump, discharge crossover valves,
and the HPI suction headers; (2) specify the HPI conditions and allowed
times that require the discharge headers be cross-connected or
separated; (3) incorporate limiting conditions for operation when
specified equipment was inoperable during specified plant conditions;
(4) specify changes in HPI system discharge path valve lineup when
certain equipment is inoperable; (5) change the requirement to reduce
reactor power when an HPI system is inoperable from 60 percent power to
75 percent power and specify the length of time operation may continue
at this power level; (6) address the failure to cross-connect the HPI
[[Page 9188]]
discharge headers as an independent condition; (7) add a requirement to
verify by administrative means that the Atmospheric Dump Valve flow
path for each steam generator is operable every 12 hours under certain
conditions; (8) add a requirement that the HPI pump and crossover
valves be restored to operable status within 30 days; (9) delete the
requirement to restore the capability to automatically actuate the HPI
within 24 hours; (10) add a Required Action to reduce reactor power to
less than or equal to 75 percent power within 3 hours in the event an
HPI train cannot be actuated by automatic or manual means; (11) expand
the Completion Time for restoring an inoperable HPI train to 72 hours;
(12) require that Limiting Condition for Operation 3.0.3 be entered
immediately if two HPI trains or two HPI (low pressure injection) -LPI
flow paths are inoperable; (13) change the surveillance requirement to
manually cycle open each LPI-HPI flow path discharge valve every 18
months to require that the HPI discharge crossover valves be cycled
every 18 months; and (14) add or modify various administrative and
Bases changes that support the proposed changes. The licensee supplied
data resulting from risk-informed analyses that were performed in
accordance with Regulatory Guides 1.174 and 1.177 to support the
evaluation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated:
No. The proposed change do not involve a physical alteration of
the plant. No new or different equipment is being installed, and no
installed equipment is being operated in a new or different manner.
No set points for parameters which initiate protective or mitigative
action are being changed.
The proposed changes do not have any impact upon the ability of
the HPI [High Pressure Injection] System to add soluble poison to
the Reactor Coolant System. The remaining potential impact is upon
the ability to mitigate the consequences of a small break LOCA
[Loss-of-Coolant Accident], which is addressed below. The small
break LOCA is the limiting design basis accident with respect to HPI
System operability requirements.
The Technical Specification requirements for the HPI System are
supported by a spectrum of small break LOCA analyses based on the
approved Evaluation Model described in FTI [Framatome Technologies
Incorporated] topical report BAW-10192PA. These small break LOCA
analyses demonstrate that the acceptance criteria of 10 CFR 50.46
are satisfied.
The requirements of LCO [Limiting Condition for Operation] 3.5.2
assure that flow can be provided via two HPI trains (i.e., one HPI
train responds automatically upon an ESPS [Engineered Safeguards
Protective System] signal, and the second HPI train is aligned
within 10 minutes via operator actions in the Control Room)
following a small break LOCA and a single active failure. The full
power small break LOCA analyses supporting this proposed license
amendment have been performed in accordance with the approved
Evaluation Model described in FTI topical report BAW-10192P.
If enhanced steam generator cooling is not credited in the
accident analysis, two HPI trains are required to mitigate specific
small break LOCAs with Thermal Power [less than or equal to] 75% RTP
[Reactor Thermal Power]. However, if equipment not qualified as QA-1
(i.e., an ADV [Atmosphic Dump Valve] flow path for one steam
generator) is credited for enhanced steam generator cooling, the
safety analyses have determined that the capacity of one HPI train
is sufficient to mitigate a small break LOCA on the discharge of the
reactor coolant pumps if Thermal Power [less than or equal to] 75%
RTP. An ADV flow path for each steam generator is credited as a
compensatory measure in Actions B and C of LCO 3.5.2 to permit
operation to continue with THERMAL POWER [less than or equal to] 75%
RTP: a) for 30 days with an HPI pump of one or more HPI discharge
crossover valve(s) inoperable; and b) for 72 hours with one HPI
train inoperable. This provides additional defense-in-depth, because
the ADV flow path for each steam generator is required to be
operable while only one is needed to perform the function.
Additionally, a risk-informed assessment (provided as Attachment 7
to Duke's license amendment request dated December 18, 1998)
concluded that operating the plant in accordance with the Required
Actions was acceptable.
The proposed changes involve crediting an additional operator
action (i.e., steaming that steam generator through an ADV flow
path) that has not previously been reviewed and approved by the
staff for licensing basis small break LOCA analyses. Additionally,
while the EFW System has been credited in past SBLOCA [small break
LOCA] analyses as described in responses to NUREG-0565, actions to
raise steam generator levels to the loss of subcooled margin
setpoint were only assumed in the smaller SBLOCAs. These operator
actions have been included in the Emergency Operating Procedure
(i.e., AP/1, 2, or 3/A/1800/001) for many years.
The times for completing these operator actions (i.e., feeding a
steam generator via EFW [Emergency Feedwater] and steaming that
steam generator through an ADV flow path) are new to the small break
LOCA analysis and the licensing basis, and are considered
reasonable. Crediting the performance of these operator actions
within the specified time frames in the SBLOCA analyses does not
result in any substantive change to the operator's response to [an]
SBLOCA.
In summary, the technical analyses described in this license
amendment justify the adequacy of this specification and assure that
operability of the HPI System is maintained in a manner consistent
with the requirements of the design basis accidents. Therefore, it
is concluded that this amendment request will not significantly
increase the probability or consequences of an accident previously
evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
No. The proposed changes do not involve a physical alteration of
the plant. No new or different equipment is being installed, and no
installed equipment is being operated in a new or different manner.
No set points for parameters which initiate protective or mitigative
action are being changed. As a result, no new failure modes are
being introduced.
The requirements of ITS [Improved Technical Specification] 3.5.2
continue to assure that operability of the HPI System is maintained
in a manner consistent with the requirements of the design basis
accidents. The requirements are supported by small break LOCA
analyses which demonstrate that the acceptance criteria of 10 CFR
50.46 are satisfied.
The proposed change involve crediting an additional operator
action (i.e., steaming that steam generator through an ADV flow
path) that has not previously been reviewed and approved by the
staff for licensing basis small break LOCA analyses. Additionally,
while the EFW System has been credited in past SBLOCA analyses as
described in responses to NUREG-0565, actions to raise steam
generator levels to the loss of subcooled margin setpoint were only
assumed in the smaller SBLOCAs. These operator actions have been
included in the Emergency Operating Procedure (i.e., AP/1, 2, or 3/
A/1800/001) for many years.
The times for completing these operator actions (i.e., feeding a
steam generator via EFW and steaming that steam generator through an
ADV flow path) are new to the small break LOCA analysis and the
licensing basis, and are considered reasonable. Crediting the
performance of these operator actions within the specified time
frames in the SBLOCA analyses does not result in any substantive
change to the operator's response to [an] SBLOCA.
Therefore, this proposed amendment will not create the
possibility of any new or different kind of accident.
(3) Involve a significant reduction in a margin of safety.
No. The requirements of ITS 3.5.2 continue to assure that
operability of the HPI System is maintained in a manner consistent
with the requirements of the design basis accidents. The
requirements are supported by small break LOCA analyses which
demonstrate that the acceptance criteria of 10 CFR 50.46 are
satisfied. These analyses were performed in accordance with the
Evaluation Model described in FTI topical report BAW-10192P.
Therefore, it is concluded that the proposed amendment request
will not result in a significant decrease in the margin of safety.
[[Page 9189]]
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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC.
NRC Project Director: Herbert N. Berkow.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of amendment request: January 18, 1999
Description of amendment request: The proposed amendments would:
(1) delete license condition 2.C.(3) from the Beaver Valley Power
Station, Unit No. 1 (BVPS-1) operating license and delete some
references to two-loop operation from BVPS-1 Technical Specifications
(TSs); (2) revise BVPS-1 and Beaver Valley Power Station, Unit No. 2
(BVPS-2) TS 2.2.1, 3.3.2.1, associated tables 2.2-1 and 3.3.4, and
associated bases, to use consistent format and wording between units;
(3) revise BVPS-1 and BVPS-2 TS 2.2.1, 3.3.2.1, associated tables 2.2-1
and 3.3.4, and associated bases, to include revised nominal trip
setpoints and allowable values which are more conservative than those
currently listed; (4) delete or revise TS to reflect the current
configuration of Unit 1 plant hardware; and (5) make miscellaneous
editorial changes to BVPS-1 and BVPS-2 TS and associated Bases to
define terms, revise formatting, modify titles, and add license numbers
to pages.
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 [as modified by the NRC staff
based upon information provided elsewhere in the licensee's submittal].
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
This proposed amendment includes changes to nominal Reactor Trip
System (RTS) and Engineered Safety Feature Actuation System (ESFAS)
trip setpoints and allowable values that have been determined with
the use of an approved methodology. The new values ensure that all
automatic protective actions will be initiated at or before the
condition assumed in the safety analysis. This change, which
includes modification of the requirements stated in Limiting Safety
System Setting (LSSS) 2.2.1 and Limiting Condition for Operation
(LCO) 3.3.2.1, will allow the nominal trip setpoints to be adjusted
within the calibration tolerance band allowed by the setpoint
methodology. There will be no adverse effect on the ability of the
channels to perform their safety functions as assumed in the safety
analyses. Since there will be no adverse effect on the trip
setpoints or the instrumentation associated with the trip setpoints,
there will be no significant increase in the probability of any
accident previously evaluated.
Other changes in trip system function, content and format are
proposed based on the current configuration of the trip system
hardware at Beaver Valley Power Station (BVPS) Unit No. 1.
Similarly, since the ability of the instrumentation to perform its
safety function is not adversely affected, there will be no
significant increase in the consequences of any accident previously
evaluated.
Since the safety analysis is unaffected by this change there is
no change in the consequences of any previously evaluated accident.
The editorial changes do not affect plant safety. The
administrative change, for BVPS Unit 1 only, pertaining to two loop
operation and Reactor Coolant System isolation valve position, does
not affect plant safety. The Technical Specification requirements in
LCOs 3.4.1.1 and 3.4.1.4.1 will continue to [prohibit two-loop
operation and] ensure safe plant operation by properly controlling
the operation and position of the reactor coolant loops and Reactor
Coolant System isolation valves.
[The administrative change to delete line item 7.d, pertaining to
Auxiliary Feedwater (AFW) Pump Auto-start on Emergency Bus
Undervoltage, from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not
affect plant safety because this function is not directly initiated
by bus undervoltage. Rather, the automatic start of the motor-driven
AFW pumps is accomplished by the combination of 1) Emergency Bus
feed breaker opening 2) valid start signal from ESFAS, and 3)
Emergency Diesel Generator (EDG) sequencer actuation. Requirements
for these items are included in the ESFAS related TS, Table 3.3-3
and 3.3-4 items 7.a, 7.c, 7.e, and EDG related TS 4.8.1.1.2.b.3 (b).
Therefore, since there is no change made to the plant hardware or
its operation and requirements related to the AFW pump auto-start
function are maintained elsewhere in the BVPS-1 TS, deleting line
item 7.d from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not
change the probability or consequences of any accident previously
evaluated.]
Therefore, this change does not involve any significant increase
in the probability of occurrence of any accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed amendment includes changes to the format and
magnitudes of nominal trip setpoints and allowable values that
preserve all safety analysis assumptions related to accident
mitigation. The protection system will continue to initiate the
protective actions as assumed in the safety analysis. The proposed
changes to LSSS 2.2.1 and LCO 3.3.2.1 will continue to ensure that
the trip setpoints are maintained consistent with the setpoint
methodology and the plant safety analysis. This proposed amendment
does not involve additional hardware changes. Plant operation will
not be changed.
Other proposed changes are made so that the Technical
Specifications more accurately reflect the plant-specific trip
system hardware in BVPS Unit No. 1.
Furthermore, the proposed changes do not alter the functioning
of the RTS and ESFAS. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed RTS and ESFAS trip setpoints
are calculated with an approved methodology. The proposed changes to
LSSS 2.2.1 and LCO 3.3.2.1 will continue to ensure that the trip
setpoints are maintained consistent with the setpoint methodology
and the plant safety analysis. Therefore, the response of the RTS
and ESFAS to accident transients reported in the Updated Final
Safety Analysis Report is unaffected by this change. No additional
hardware changes are involved. Therefore, accident analysis
acceptance criteria are not affected. Other proposed changes are
made so that the protection system Technical Specifications more
accurately reflect the plant-specific trip system hardware in BVPS
Unit No. 1.
The editorial changes do not affect plant safety. The
administrative change, for BVPS Unit 1 only, pertaining to two loop
operation and Reactor Coolant System isolation valve position, does
not affect plant safety. The Technical Specification requirements in
LCOs 3.4.1.1 and 3.4.1.4.1 will continue to [prohibit two-loop
operation and] ensure safe plant operation by properly controlling
the operation and position of the reactor coolant loops and Reactor
Coolant System isolation valve.
[The administrative change to delete line item 7.d, pertaining to
Auxiliary Feedwater (AFW) Pump Auto-start on Emergency Bus
Undervoltage, from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not
affect plant safety because this function is not directly initiated
by bus undervoltage. Rather, the automatic start of the motor-driven
AFW pumps is accomplished by the combination of (1) Emergency Bus
feed breaker opening, (2) valid start signal from ESFAS, and (3) EDG
sequencer actuation. Requirements for these items are included in
the ESFAS related TS, Table 3.3-3 and 3.3-4 items 7.a, 7.c, 7.e, and
EDG related TS 4.8.1.1.2.b.3 (b). Therefore, since there is no
change made to the plant hardware or its operation and requirements
related to the AFW pump auto-start function are maintained elsewhere
in the BVPS-1 TS,
[[Page 9190]]
deleting line item 7.d from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2
will not involve a significant reduction in a margin of safety.]
Therefore, operation of the facility in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: S. Singh Bajwa.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: December 16, 1998.
Description of amendment request: The licensee has proposed an
amendment of Facility Operating License No. NPF-47, Appendix A--
Technical Specifications, Section 2.1.1.2, entitled ``Reactor Core
[Safety Limits].'' The proposed amendment will change the two
recirculation loop Minimum Critical Power Ratio (MCPR) limit from 1.13
to 1.12 and the single recirculation loop MCPR limit from 1.14 to 1.13.
The revised limits are necessary to address the operation of Cycle 9
following the refueling outage which is scheduled to begin April 1999.
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 request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The plant/cycle specific SLMCPRs have been calculated using
methods identical to those used by General Electric (GE) to assess
the SLMCPR for other Boiling Water Reactors (BWRs). Similar methods
were used to determine the value of the SLMCPR for the previous
cycle. These methods are within the existing design and licensing
basis and cannot increase the probability or severity of an
accident. The basis of the SLMCPR calculation is to ensure that
greater than 99.9% of all fuel rods in the core avoid transition
boiling and fuel damage in the event of the occurrence of
Anticipated Operational Occurrences (AOO) or a postulated accident.
The SLMCPR is used to establish the Operating Limit Minimum
Critical Power Ratio (OLMCPR). Neither the SLMCPR nor the OLMCPR are
initiators or affect initiators of an accident previously evaluated
and therefore changes to the SLMCPR do not increase the probability
of any accident previously evaluated. The proposed changes involve
the use of an accepted methodology in calculating the SLMCPR and,
since there is no change in the definition of the SLMCPR, these
changes will not affect the consequences of any accident previously
evaluated. In addition, the proposed changes do not involve any
change in the way the plant is operated. Existing procedures will
ensure that the SLMCPR is not violated. Therefore, these changes
have no effect on the consequences of an accident.
On these bases, there will be no increase in the probability or
consequences of an accident previously analyzed as a result the
proposed changes.
2. The request does not create the possibility of occurrence of
a new or different kind of accident from any accident previously
evaluated.
The proposed changes consist of SLMCPR calculated from an
accepted method of analysis that has been used by many BWRs. These
changes do not involve any alteration of the plant and do not affect
the plant operation. Neither the SLMCPR nor the OLMCPR can initiate
an event, therefore a change to the SLMCPR does not create the
possibility of occurrence of a new or different kind of accident
from any accident previously evaluated.
3. The request does not involve a significant reduction in the
margin of safety.
The SLMCPR is a Technical Specification numerical value to
ensure that 99.9% of all fuel rods in the core will avoid transition
boiling if the limit is not violated. The proposed SLMCPR change
results from SLMCPR analysis using the accepted methods as
identified in the Attachment.
The margin of safety resides between the SLMCPR and the point at
which fuel fails. Maintaining the MCPR above the proposed SLMCPR
will maintain the margin of safety associated with GE's SLMCPR
methodology. Existing plant procedures will continue to ensure that
the SLMCPR is not violated.
Therefore, this request does not involve a reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005.
NRC Project Director: John N. Hannon.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 23, 1998.
Description of amendment request: The proposed changes will modify
the Limiting Condition for Operation for Technical Specifications
3.3.3.7.1 for the chlorine detection system at Waterford Steam Electric
Station, Unit 3. A change in the alarm/trip setpoint from 3 parts per
million (ppm) to 2 ppm is requested. Additionally, the proposed request
corrects a typographical error in Table 3.3-4.
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. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: The chlorine detection system has no effect on the
accidents analyzed in Chapter 15 of the Final Safety Analysis
Report. Its only effect is on habitability of the control room,
which will be enhanced by specifying a more conservative setpoint in
the Technical Specifications (TS). Analysis using more conservative
assumptions show that a setpoint of 2 parts per million (ppm)
chlorine is acceptable.
Correcting the typographical error on TS page 3/4 3-19 has no
effect on the probability or consequences of an accident previously
evaluated.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: The proposed Technical Specification change in itself
does not change the design or configuration of the plant. Using a
more conservative setpoint performs the same function as the old
setpoint, but it accomplishes this function with increased
conservatism.
Correcting the typographical error on TS page 3/4 3-19 will not
create the possibility of a new or different type of accident from
any accident previously evaluated.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Will operation of the facility in accordance with this
proposed change
[[Page 9191]]
involve a significant reduction in a margin of safety?
Response: The chlorine detection system has no effect on a
margin of safety as defined by Section 2 of the Technical
Specifications. Its only effect is on habitability of the control
room, which will be enhanced by a more conservative setpoint
provided by this change to the Technical Specifications.
Correcting the typographical error on TS page 3/4 3-19 does not
involve a significant reduction in a margin of safety.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400
L Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: John N. Hannon.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: January 25, 1999.
Description of amendment request: The proposed change request will
modify Technical Specification (TS) 3.5.1 to allow up to 72 hours to
restore safety injection tank (SIT) to operable status if one SIT is
inoperable due to boron concentration not within the limits or the
inability to verify level and pressure. The proposed change would also
allow up to 24 hours to restore SIT to operable status if one SIT is
inoperable due to other reasons when Reactor Coolant System pressure is
greater than or equal to 1750 psia. The ACTIONS for an inoperable SIT
are being subdivided based on pressurizer pressure to be consistent
with the current Waterford 3 requirements and applicability.
Additionally, the Surveillance requirement to sample the SIT after a 1%
volume increase is being changed to not be required if the source of
the makeup is the refueling water storage pool. This amendment request
is a collaborative effort of participating Combustion Engineering
Owners Group members based on a review of plant operations,
deterministic and design basis considerations, and plant risk, as well
as previous generic studies and conclusions drawn by the NRC Staff and
contained within NUREG-1366, ``Improvements to Technical Specifications
Surveillance Requirements,'' and NUREG-1432, Revision 1, ``Standard
Technical Specifications for Combustion Engineering (CE) Plants.'' TS
Bases 3/4.5.1 will be revised to support above changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: The Safety Injection Tanks (SITs) are passive
components in the Emergency Core Cooling System. The SITs are not an
accident initiator in any accident previously evaluated. Therefore,
this change does not involve an increase in the probability of an
accident previously evaluated.
The SITs were designed to mitigate the consequences of Loss of
Coolant Accidents (LOCA). These proposed changes do not affect any
of the assumptions used in deterministic LOCA analyses. Hence the
consequences of accidents previously evaluated do not change.
In order to fully evaluate the affect of the SIT Allowed Outage
Time (AOT) extension from 1 hour to 24 hours when one SIT is
inoperable for reasons other than boron concentration or inability
to measure level or pressure, probabilistic safety analysis (PSA)
methods were utilized. The results of these analyses show no
significant increase in the core damage frequency. As a result,
there would be no significant increase in the consequences of an
accident previously evaluated. These analyses are detailed in CE
NPSD-994, Combustion Engineering Owners Group ``Joint Applications
Report for Safety Injection Tank AOT/STI Extension.''
The proposed change to extend the AOT from 1 hour to 72 hours
when unable to measure level or pressure is acceptable because SIT
operability is not based on instrumentation availability. Therefore,
this does not involve a significant increase in the consequences of
an accident as evaluated and are endorsed by the Nuclear Regulatory
Commission (NRC) in NUREG-1366, ``Improvements to Technical
Specifications Surveillance Requirements.'' The inability to measure
level or pressure is acceptable because the SIT instrumentation
provides no safety actuation.
The AOT extension from 1 hour to 72 hours, based upon boron
concentration outside the prescribed limits does not involve a
significant increase in the consequences of an accident as evaluated
and approved by the NRC in NUREG-1432, ``Standard Technical
Specifications for Combustion Engineering Plants.'' These changes
are acceptable because the reduced concentration effects on core
subcriticality during reflood are minor.
The change in sampling requirements to not require sampling if
the makeup source is of the same concentration limit as the SIT is
acceptable as the concentration will remain within the TS limits.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: The proposed change does not alter the design or
configuration of the plant. It also does not alter the mitigation
capabilities of any safety system or components. This change
increases the AOTs for the condition of SIT inoperability. The boron
concentration is maintained by make-up from a source of water with
the required concentration of the SITs.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: The proposed changes do not affect the limiting
conditions for operation or their bases that are used in the
deterministic analyses to establish the margin of safety. PSA and
deterministic evaluations were used to evaluate these changes. The
PSA evaluations demonstrated that the applicable changes are either
risk neutral or risk beneficial. These evaluations are detailed in
CE NPSD-994. The deterministic evaluations show that the SITs would
be able to perform their safety function. These changes are
consistent with NUREG-1366 and NUREG-1432. The margin of safety is
not significantly affected by makeup from a source of the same
concentration limit as the SIT or increase in the AOT for boron
concentration of one SIT not within limits.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400
L Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: John N. Hannon.
[[Page 9192]]
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: January 25, 1999.
Description of amendment request: The proposed changes modify
Technical Specifications Section 6.0 to remove certain administrative
controls and instead rely on the change controls of 10 CFR 50.54(a)(3)
and to add a requirement to Section 6.0 concerning the responsibilities
of the General Manager Plant Operations. The requested changes are
consistent with the Improved Standard Technical Specifications for
Combustion Engineering plants, NUREG-1432.
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. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: The requested changes are purely administrative in
nature. The proposed changes do not affect the operation of any
structures, systems, or components or the assumptions of any
accident analyses. The requested changes only affect Section 6.0 of
the Waterford 3 Technical Specifications which describe the
administrative controls to be implemented at the site. The requested
changes either add an additional administrative requirement or
remove quality assurance program details from the Technical
Specifications. The details are being removed from the Technical
Specifications and instead rely on the change controls of 10 CFR
50.54(a)(3). This submittal makes no changes to the regulatory
controls governing changes. The requested changes are purely
administrative in nature.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: The proposed changes to the Technical Specification
requirements are purely administrative in nature and do not involve
a change in plant design or affect the configuration or operation of
any structure, system, or component.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: The proposed changes do not affect the operation of
any structures, systems, or components or the assumptions of any
accident analyses. The requested changes are purely administrative
in nature.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety. The NRC staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400
L Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: John N. Hannon.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: January 22, 1999.
Description of amendment request: The proposed amendment would
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)
Section 4.3, ``Fuel Storage,'' by updating the criticality requirements
(k-infinity and U-235 enrichment limits) for storage of fuel assemblies
in the spent fuel racks. This change would allow for storage of nuclear
fuel assemblies with new designs, including GE-12 with a 10X10 pin
array.
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:
After reviewing this proposed amendment, we have concluded:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The probability of occurrence of the accident/abnormal
conditions evaluated in UFSAR Section 9.1.2.3 is not significantly
increased by this change because no modification in fuel handling
equipment, fuel pool cooling equipment, fuel storage racks, or fuel
handling practices is taking place. Only the k-infinity and
enrichment limits for the stored fuel are being changed.
The postulated accident/abnormal conditions evaluated in UFSAR
Section 9.1.2.3 have been re-evaluated for the proposed changes in
k-infinity and enrichment limits. The results demonstrate that the
consequences are negligible. The analyses performed show that the
requirement to maintain K-eff less than 0.95 (substantially
subcritical) is satisfied for normal and postulated abnormal
conditions using methods and assumptions that are consistent with
the existing UFSAR. Seismic adequacy and structural integrity of the
pool and racks are not affected by the introduction of GE-12 fuel.
Local and bulk pool temperatures remain bounded by the current UFSAR
analysis for fuel exposures with GE-12 fuel expected through two
cycles of operation (i.e., through Cycle 18 operation). Based upon a
scoping study comparing the hydraulic diameters of GE-10 and GE-12
fuel, large margins to pool boiling conditions at the final
discharge exposures of GE-12 fuel will be maintained. Therefore, the
consequences of the accident are not significantly increased by this
change.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
No new types of accidents are being introduced because no
modification in fuel handling equipment, fuel pool cooling
equipment, fuel storage racks or fuel handling procedures is being
made. The design basis function of the spent fuel racks is to
maintain the fuel configuration substantially subcritical and within
allowable temperatures under both normal and postulated abnormal
conditions. This design basis function will be maintained with the
proposed k-infinity and enrichment limits.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
The margin of safety is not significantly reduced. This margin
is based on the requirement to limit the K-eff of fuel in the spent
fuel racks to less than 0.95. The proposed changes in k-infinity and
enrichment limits have been shown to meet this requirement, using
methods and assumptions that are consistent with the existing UFSAR.
Seismic adequacy and structural integrity of the pool and racks are
not affected by the introduction of GE-12 fuel. Local and bulk pool
temperatures remain bounded by the current UFSAR analysis for fuel
exposures with GE-12 fuel expected through two cycles of operation
(i.e., through Cycle 18 operation). Based upon a scoping study
comparing the hydraulic diameters of GE-10 and GE-12 fuel, large
margins to pool boiling conditions at the final discharge exposures
of GE-12 fuel will be maintained.
Based upon the above, we have determined that the proposed
amendment will not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401.
[[Page 9193]]
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: Cynthia A. Carpenter.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: October 15, 1998, as supplemented on
December 21, 1998.
Description of amendment request: The proposed amendment would
revise the Duane Arnold Energy Center (DAEC) Technical Specifications
(TS) by adding a new TS 3.7.9, ``Control Building/Standby Gas Treatment
System (CB/SBGT) Instrument Air System.'' The proposed amendment would
also revise (TS) 3.6.1.3, ``Primary Containment Isolation Valves
(PCIVs),'' Condition E, by adding a time limit for plant operation if a
penetration flow path is isolated by a single purge valve with
resilient seal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The amendment is adding new requirements for the CB/SBGT
Instrument Air System that are commensurate with the safety
functions it supports and consistent with other support systems in
the Technical Specifications. These requirements provide appropriate
actions and time limits for plant operation with one or both CB/SBGT
Instrument Air subsystems inoperable. The probability of an event
while in this condition is low, and the consequences are bounded by
the failure of the supported systems. The CB/SBGT Instrument Air
System is not assumed to be an initiator of an analyzed event.
The amendment is also adding a time limit for plant operation if
a purge valve with resilient seal is used to satisfy TS 3.6.1.3
Required Action E.1 (isolate the affected penetration flow path).
While primary containment integrity is provided by the purge valve,
it is prudent to limit operation in this condition due to the
potential for increased leakage from a single active failure.
These additions will provide assurance that affected systems
will be OPERABLE when required and as assumed in the design basis.
This change will not physically alter the plant (no new or
different type of equipment will be installed). This change will not
alter the operation of process variables, structures, systems, or
components as described in the safety analysis. This change will not
alter assumptions relative to the mitigation of an accident or
transient event. This change will not increase the probability of
initiating, or the consequences of an analyzed event.
(2) The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The amendment adds new requirements for the CB/SBGT Instrument
Air System and adds a time limit for plant operation if a purge
valve with resilient seal is used to satisfy TS 3.6.1.3 Required
Action E.1.
This change will not physically alter the plant (no new or
different type of equipment will be installed). This change will not
alter the operation of process variables, structures, systems, or
components as described in the safety analysis. Thus, a new or
different kind of accident will not be created.
(3) The proposed amendment will not involve a significant
reduction in a margin of safety.
The amendment is adding new requirements for the CB/SBGT
Instrument Air System to provide appropriate actions and time limits
for plant operation with one or both CB/SBGT Instrument Air
subsystems inoperable.
The amendment is also adding a time limit for plant operation if
a purge valve with resilient seal is used to satisfy TS 3.6.1.3
Required Action E.1 (isolate the affected penetration flow path).
While primary containment integrity is provided by the purge valve,
it is prudent to limit operation in this condition due to the
potential for increased leakage from a single active failure in the
remaining OPERABLE components.
This change will not physically alter the plant (no new or
different type of equipment will be installed). This change will not
alter the operation of process variables, structures, systems, or
components as described in the safety analysis. This change will not
alter assumptions relative to the primary success path for
mitigation of an accident or transient event.
These additions will provide assurance that the accident
mitigation functions will perform as assumed in the safety analysis.
Thus, the margin of safety will not be reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401.
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: Cynthia A. Carpenter.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: January 29, 1999.
Description of amendment request: The amendment would revise the
technical specifications (TS) to relocate three cycle-specific
parameter limits; shutdown margin with Tcold>210 deg.F,
moderator temperature coefficient, and minimum boric acid storage tank
level versus concentration, to the Core Operating Limits Report (COLR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The safety analysis most impacted by a change to the negative
Moderator Temperature Coefficient (MTC) limit is the Main Steam Line
Break (MSLB) event. The Steam Line Break Cooldown curves for an MTC
are calculated and then input to the cycle-specific MSLB analysis
(if necessary) during the reload analysis process, using an NRC-
approved methodology. The required/acceptable Shutdown Margin (SDM)
is dependent upon the core loading pattern used (i.e., cycle-
specific core physics parameters) and is largely dependent on the
cycle-specific MTC and available scram worth. The SDM is determined
based on the analysis of the Hot Zero Power (HZP) MSLB event in
which the return-to-critical and return-to-power conditions are
evaluated to provide acceptable results. With the ongoing changes in
MTC as a result of core loadings for FCS and higher U-235
enrichments, the end-of-cycle MTC is becoming more negative than the
present Technical Specifications limit. Since the MTC is fuel cycle
specific and influences the required SDM, it is appropriate to move
both of these values to the COLR, consistent with Generic Letter 88-
16. Note that no change to the SDM for Tcold
210 deg.F is being proposed.
The cycle-specific reload analysis is performed for every
operating cycle and the results, as incorporated into the COLR
pursuant to the 10 CFR 50.59 process, are transmitted to the NRC.
FCS will continue to provide COLR updates to the NRC. The relocation
of the negative MTC and the ``BAST level versus BAST Concentration''
curves into the COLR, consistent with the NRC recommendations of
Generic Letter 88-16, will not modify the methodology used in
generating the limits, nor the manner in which they are implemented.
These limits will continue to be determined by analyzing the same
postulated events as previously analyzed. FCS will continue to
operate within the limits specified in the COLR and will take the
same corrective actions when or if these limits are exceeded as
required by
[[Page 9194]]
current Technical Specifications. The potential increase of the
absolute magnitude of the negative MTC with Shutdown Margin decrease
is evaluated during the COLR reload analysis process in accordance
with OPPD's NRC-approved topical report. Therefore, this proposed
amendment is administrative in nature and has been concluded not to
increase the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to FCS Technical Specifications were the
result of a recommendation from a Generic Letter. Future changes to
the parameters being relocated to the COLR can only be performed
with approved Reload Analyses. No new or different kind of accident
is created by this administrative change because the actual
operation of FCS remains unchanged. Therefore the possibility of an
accident or malfunction of a different type than previously
evaluated in the safety analysis report would not be created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
As indicated above, the implementation of this proposed COLR
change, consistent with the guidance of Generic Letter 88-16, makes
use of the existing safety analysis methodologies and the resulting
limits and setpoints for plant operation. Additionally, the safety
analysis acceptance criteria for operation with this proposed
amendment have not changed from the criteria used in the current
reload analysis. Therefore, the margin of safety as defined in the
bases of Technical Specifications is not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502.
NRC Project Director: William H. Bateman.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: January 12, 1999.
Description of amendment request: The proposed change involves
revising Technical Specification (TS) Section 3/4.4.2, ``Safety/Relief
Valves,'' and TS Bases Sections B 3/4.4.2, B 3/4.5.1 and B 3/4.5.2, to
increase the allowable as-found main steam Safety Relief Valve (SRV)
code safety function lift setpoint tolerance from plus or minus 1% to
plus or minus 3%. This change will also require the as-left SRV code
safety function lift setting to be set within plus or minus 1% of the
specified nominal lift setpoint prior to reinstallation in the plant.
In support of this proposed TS change, the required number of OPERABLE
SRVs in Operational Conditions (OPCONs) 1, 2, and 3 will be changed
from 11 to 12. The number of SRVs in each lift pressure grouping will
remain the same. This proposed TS change does not alter the SRV nominal
lift setpoints or the SRV lift setpoint test frequency currently
specified by TS Section 3/4.4.2. The proposed change does not change
the SRV testing commitment specified in LGS Updated Final Safety
Analysis Report (UFSAR) Chapter 5.2.2.10, ``Inspection and Testing.''
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 Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS changes allow for an increase in the as-found
main steam Safety Relief Valve (SRV) setpoint tolerance from plus or
minus 1% to plus or minus 3%. The proposed changes also reduce the
allowable number of SRVs to be out-of-service from three (3) to two
(2). The proposed changes do not alter the SRV nominal lift
setpoints or SRV lift setpoint test frequency. The actuation of an
SRV is the precursor to the inadvertent opening of a SRV transient,
as discussed in Updated Final Safety Analysis Report (UFSAR) Chapter
15.1.4. Increasing the allowable as-found SRV code safety function
lift setpoint tolerance from plus or minus 1% to plus or minus 3%
does have the potential for the minimum SRV simmer margin to be
reduced from 113.3 psig to 89.9 psig. A reduction in simmer margin
will not directly result in an increase of the probability on an
inadvertent self actuation of an SRV. A reduction in simmer margin
will reduce the seating force which may initiate leakage. However,
this leakage is monitored and corrective actions can be implemented
prior to progressing to the point of the potential of an inadvertent
actuation. This reduction in SRV simmer margin has been evaluated by
the SRV manufacturer and determined to be acceptable; therefore, the
probability of an inadvertent SRV actuation remains unchanged.
Actuation of an SRV is not a precursor for any other event evaluated
in the Safety Analysis Report (SAR).
The proposed TS changes have been evaluated on both a generic
and plant specific basis. The NRC has approved the general approach
of this change; however, implementation is contingent on several
plant specific evaluations. The required plant specific analyses and
evaluations included transient analysis of the anticipated
operational transients (AOTs); analysis of the design basis
overpressurization event; evaluation of the performance of high
pressure systems, motor operated valves, and vessel instrumentation
and associated piping; and evaluation of the containment response
during Loss-of-Coolant Accident (LOCA) and hydrodynamic loads on the
SRV discharge lines and containment. In addition to the plant
specific analyses and evaluations required by the NRC, the following
items were also considered: ECCS/LOCA [Emergency Core Cooling
System] performance, SRV simmer margin, high pressure--low pressure
interfaces, i.e., High Energy Line Break (HELB), Station Blackout
(SBO), and Fire Safe Shutdown (FSSD), and the short term
pressurization phase of an ATWS [anticipated transient without
scram] event. These analyses and evaluations show that there is
adequate margin to the design core thermal limits and reactor vessel
pressure limits using the plus or minus 3% SRV code safety function
lift setpoint tolerance and two (2) SRVs out-of-service. The
analyses and evaluations also show that the operation of the high
pressure injection systems will not be adversely affected, that SRV
discharge piping stresses will not be exceeded, and that the
containment response during a LOCA will be acceptable.
Evaluations of the impact of the proposed change on the
Equipment Important to Safety have been performed and no adverse
conditions were identified. The reactor pressure vessel and attached
systems and piping have been evaluated for the impact of this
proposed TS change. A plant specific analysis has been performed
which indicates that neither the American Society of Mechanical
Engineers (ASME) Code upset limits or the TS Safety Limits for the
reactor pressure vessel will be exceeded for the limiting event,
i.e., Main Steam Isolation Valve (MSIV) closure with flux Scram. The
reactor pressure vessel and attached piping design values will not
be exceeded. The current high pressure--low pressure interface
evaluation utilized nominal SRV setpoints, and therefore, is
unaffected. Therefore, the probability of a malfunction of the
reactor pressure vessel and attached systems and piping is not
increased.
The nuclear fuel has been evaluated for the impact of the
proposed change. Plant specific analyses were performed which
indicate that for all abnormal operational transients adequate
margin to the limiting thermal limit parameter, i.e., Minimum
Critical Power Ratio (MCPR), is maintained. Emergency Core Cooling
System (ECCS)/LOCA performance is maintained adequate to meet the
requirements of 10CFR50.46. Therefore, the probability of the
malfunction of the nuclear fuel is not increased.
The SRVs have been evaluated for the impact of the proposed TS
changes. No physical changes to the SRVs will be made as a result of
the proposed TS changes. Adequate simmer margin will be maintained
with the increased tolerance to ensure that an inadvertent lifting
of a SRV does not occur.
[[Page 9195]]
The increase in SRV discharge flow and reactor vessel pressure due
to the potential for higher SRV lift setpoints are bounded by the
SRV steam flows and reactor vessel pressure currently used in the
evaluation of SRV discharge piping, quencher, quencher support, and
hydrodynamic loads on the suppression pool and submerged structures;
therefore, the probability of a malfunction of a SRV or associated
components and structures is not increased.
The Containment response during a LOCA has been evaluated for
the impact of the proposed change. The major factor in the
Containment response to a LOCA is the rate of reactor vessel water
inventory loss. The rate of reactor vessel water inventory loss is
mainly dependent on reactor decay heat which is not affected by the
proposed change. Therefore, the probability of the malfunction of
the Containment is not increased.
The High Pressure Coolant Injection (HPCI) system has been
evaluated for the impact of the proposed TS changes. The analysis
determined that the HPCI system would not be capable of developing
its design flowrate of 5600 gpm at a reactor pressure of 1205 psig
(lowest SRV nominal setpoint +3% tolerance) unless the HPCI turbine/
pump maximum rated speed was increased. However, increasing the HPCI
turbine/pump maximum rated speed is prevented due to HPCI pump
discharge piping overpressurization concerns. Further analysis has
shown that the HPCI system is capable of meeting its required ECCS
function design flowrate, and its required non-ECCS flowrate,
without any change to the current system operating parameters.
Therefore, the probability of a malfunction of the HPCI System is
not increased.
The Reactor Core Isolation Cooling (RCIC) system has been
evaluated for the impact of the proposed change. The analysis
determined that in order for the RCIC system to be capable of
injecting its design flowrate of 600 gpm at a reactor pressure of
1205 psig (lowest SRV setpoint of 1170 psig +3% tolerance) the
maximum rated speed of the RCIC turbine/pump is required to be
increased from 4575 rpm to 4625 rpm. This increase in the RCIC
turbine/pump maximum rated speed will reduce the margin to the
overspeed trip from 123% to 122.1%. This reduction in the margin to
the overspeed trip is acceptable due to the implementation of plant
Modification P00210, ``RCIC System Startup Transient Improvement,''
which reduced the amount of turbine/pump speed overshoot during
system startup. The RCIC overspeed trip setpoint will not be
changed; therefore, a failure of the RCIC turbine/pump (missile
hazard or system overpressurization) due to overspeed is not
increased. All other RCIC System components will continue to operate
within the currently specified design and operating limits.
Therefore, the probability of a malfunction of the RCIC System is
not increased.
The Standby Liquid Control (SLC) system has been evaluated for
the impact of the proposed change. The SLC system capability of
shutting down the reactor during a postulated event in which all or
some of the control rods cannot be inserted or during a postulated
Anticipated Transient Without Scram (ATWS) event is not impacted by
this proposed change. Therefore, the probability of a malfunction of
the SLCS is not increased.
The Control Rod Drive (CRD) system has been evaluated for the
impact of the proposed change. The CRD system capability of
controlling reactor power during normal plant operation and rapidly
inserting control rod blades (Scram) during abnormal plant
conditions is not impacted by the proposed change. Therefore, the
probability of a malfunction of the CRD system is not increased.
The Reactor Vessel Instrumentation System has been evaluated for
the impact of the proposed change. The Reactor Vessel
Instrumentation System will continue to be operated within the
current design pressure/temperature requirements; therefore, the
probability of a malfunction of the Reactor Vessel Instrumentation
System is not increased.
The LGS, Units 1 and 2, Generic Letter 89-10 Motor-Operated
Valve (MOV) Program has been evaluated for the proposed change. The
LGS MOV Program currently uses SRV nominal setpoints for
differential pressure determinations for valves in which reactor
pressure at the SRV setpoint is limiting. Use of nominal SRV
setpoints is consistent with current industry practice. Therefore,
the probability of a malfunction of a MOV is not increased.
Reducing the number of SRVs allowed to be out-of-service does
not make the consequences of a malfunction of a SRV more severe,
since the number of SRVs required to maintain the reactor vessel
within ASME Code and TS Safety Limits will be maintained OPERABLE.
The proposed change does not result in any changes to the
interactions of any system, structure, or component. All systems,
structures, and components will continue to function as designed.
Therefore, the proposed TS changes do not significantly increase
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes allow for an increase in the as-found
SRV setpoint tolerance from plus or minus 1% to plus or minus 3%.
The proposed TS changes also reduce the allowable number of SRVs to
be out-of-service from three (3) to two (2). Generic and plant
specific analyses and evaluations indicate that the plant response
to any previously evaluated event will remain unchanged. All plant
systems, structures, and components will continue to be capable of
performing their required safety function as required by event
analysis guidance.
The proposed TS changes do not alter the SRV nominal lift
setpoints or SRV lift setpoint test frequency. The operation and
response of the affected Equipment Important to Safety is unchanged.
All systems, structures, and components will continue to be operated
within acceptable operating and/or design parameters. No system,
structure, or component will be subjected to a condition that has
not been evaluated and determined to be acceptable using the
guidance required for specific event analysis.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed TS changes allow for an increase in the as-found
SRV setpoint tolerance from plus or minus 1% to plus or minus 3%.
The proposed TS changes also reduce the allowable number of SRVs to
be out-of-service from three (3) to two (2). The proposed TS changes
do not alter the SRV nominal lift setpoints or SRV lift setpoint
test frequency. The operation and response of the affected Equipment
Important to Safety is unchanged. All systems, structures, and
components will continue to be operated within acceptable operating
and/or design parameters. While the calculated peak reactor vessel
pressure for the ASME overpressure event and the ATWS Pressure
Regulator Failure-Open (PREGO) event are higher than those
calculated without the increase in setpoint tolerance, both are
still within the respective licensing acceptance limits associated
with these events. These licensing acceptance limits have been
determined by the NRC to provide a sufficient margin of safety.
The increase in the RCIC system turbine/pump maximum rated speed
is within the capability of the system design. The reduction in the
margin to the overspeed trip is not a reduction in the margin of
safety, since the operation of the RCIC System has demonstrated
minimal speed overshoot on system initiation due to the installation
of plant Modification P00210, ``RCIC System Startup Transient
Improvement.''
The inability of the HPCI system to be capable of injecting 5600
gpm at a reactor pressure of 1205 psig (lowest SRV nominal setpoint
of 1170 psig +3% tolerance) is not a reduction in the margin of
safety, since analysis for events that would result in high reactor
vessel pressure indicate that the HPCI System is capable of
providing adequate coolant injection.
The increase in SRV steam flow and reactor vessel pressure does
not reduce the margin of safety associated with the SRVs and
associated components and structures since the increased SRV steam
flow rate and reactor vessel pressure are bounded by the current
design analysis.
The margin of safety for fuel thermal limits and 10CFR50.46
limits is unaffected by the proposed change.
The margin of safety for the Containment is unaffected by the
proposed change.
The capability of the SLC system to perform its safety function
during all required events, using the required guidance for event
analysis, is maintained. Therefore, the proposed changes do not
reduce the margin of safety provided by the SLC system.
Therefore, these proposed TS 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
[[Page 9196]]
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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Project Director: William M. Dean.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: January 25, 1999.
Description of amendment request: The proposed Technical
Specification (TS) Change Request revises the TS Surveillance
Requirement frequencies for Sections 4.8.1.1.2.e.1, 4.8.1.1.2.e.8.a,
and 4.8.1.1.2.e.8.b for the Emergency Diesel Generator maintenance
inspection outages, the 24-hour endurance run, and for the hot restart
test from 18 to 24 months.
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 Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The maintenance inspection interval change and the corresponding
interval change for the associated 24 hour endurance test and hot
restart test which are normally performed in conjunction with the
diesel preventive maintenance overhaul inspections, as well as the
programmatic improvements addressed here do not involve physical
changes that would affect the ability of the EDGs [emergency diesel
generators] to perform their safety function. The Emergency Diesel
Generator System is not an accident initiator.
The Surveillance Testing requirements of Technical Specification
Section 3/4.8 will continue to verify the operability and
reliability of the Emergency Diesel Generator system.
The proposed changes do not affect the ability of the EDGs to
mitigate the consequences of an accident, including the Loss of
Coolant Accident (LOCA) coupled with Loss Of Offsite Power accident
analyses as presented in Chapter 15 of the LGS [Limerick Generating
Station] UFSAR [Updated Final Safety Analysis Report]. EDG
unavailability due mostly to outage inspections is more than 2 times
higher than EDG unplanned unavailability. An extension of the outage
inspection frequency to 24 months will result in increased EDG
availability to mitigate the consequences of a potential accident.
When this program is taken in its entirety the extended maintenance
intervals coupled with the defined enhancements is judged to result
in an overall increase in EDG availability and reliability.
Therefore, the probability or consequences of an accident previously
evaluated is not increased.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The Emergency Diesel Generator system is not an accident
initiator. The operation and design of the onsite emergency power
system (including the EDGs) is not being changed; only the overhaul
inspection interval coupled with the program improvements and the
corresponding interval change for the associated 24 hour endurance
test and hot restart test, (which are normally performed in
conjunction with the diesel preventive maintenance overhaul
inspections), are changed. The EDG system meets the single failure
criteria at the EDG unit level, i.e., the SAR [safety analysis
report] states that with one EDG failed or out-of-service, the
standby AC system is capable of furnishing sufficient power for the
minimum Class 1E load demand, assuming a limiting design basis
accident has occurred. The proposed changes involve a routine
preventive maintenance and inspection time interval change along
with the corresponding surveillance test interval changes, and also
include programmatic improvements to reduce the likelihood of a
failure of an individual EDG unit; the proposed changes do not
involve any physical design or operational changes that could create
a malfunction extending beyond an individual EDG nor do they
increase the potential for a common-mode EDG failure. Therefore, it
is not possible to create a new or different type of accident
through implementation of these changes.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The changes to bring the frequencies of the EDG overhaul, the 24
hour endurance test and the associated hot restart test into
alignment with the current 2 year operating cycle, and the detailed
programmatic changes to achieve conformance with the FMOG [Fairbanks
Morse Owners Group] recommended maintenance program, will increase
the reliability and availability of the EDG system. This will
enhance the margin of safety as the amount of time the EDGs are out-
of-service will decrease and the system will be single-failure proof
for more clock hours when the nuclear reactor(s) are operating. The
changes discussed here do not result in operation of the emergency
diesel generator system nor any other plant system in a manner
beyond their original design basis, and thus does not reduce any
explicit or implicit Technical Specification 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Project Director: William M. Dean.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: February 12, 1997.
Brief description of amendment: The proposed amendment would delete
a portion of the Trojan site from the 10 CFR 50 license when that
portion of the site, designated for use as an independently licensed
spent fuel storage installation (ISFSI), receives a part 72 license.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensees' analysis
against the standards of 10 CFR 50.92(c). The licensee's analysis is
summarized below:
The proposed changes would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change is administrative in nature and has no impact on
the probability or consequences of accidents previously evaluated. The
physical structures, systems, and components of the Trojan Nuclear
Plant and the operating procedures for their use are unaffected by this
proposed change. The proposed action would eliminate the ISFSI area
from the Part 50 license when the Part 72 license is issued. The 10 CFR
72 licensing controls for the area will assure an adequate level of
safety for the area during normal operation of the ISFSI and during
abnormal events or accidents. Therefore the proposed Part 50 amendment
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes would not create the possibility of a new or
different kind of accident from any accident previously evaluated. The
proposed action would eliminate the ISFSI area from the Part 50 license
when the Part 72 license is issued. The proposed change is
administrative in
[[Page 9197]]
nature and has no impact on plant systems, structures, or components or
on any procedures for operating the plant equipment. The ISFSI will be
separately licensed under Part 72 and physically separated from the
Part 50 licensed structures and equipment. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from those previously evaluated.
The proposed changes do not involve reduction in the margin of
safety. The Trojan Permanently Defueled Technical Specifications (PDTS)
contain four limiting conditions of operation that address: 1) Spent
Fuel Water Level, 2) Spent Fuel Pool Boron Concentration, 3) Spent Fuel
Pool Temperature, and 4) Spent Fuel Pool load restrictions. These PDTS
will remain in effect as long as spent fuel is stored in the Spent Fuel
Pool, which is in accordance with their applicability statements. The
ISFSI area is physically separated from the Spent Fuel Pool area and
the Fuel Building and will have no effect on spent fuel water level,
spent fuel pool boron concentration, spent fuel pool temperature, or
loads over the Spent Fuel Pool. The proposed change is administrative
and does not affect plant equipment, operating parameters, or
procedures. Based on the above, the proposed change will not reduce the
margin of safety.
Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Attorney for licensees: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S. W. Salmon Street, Portland, Oregon 97204.
NRR Project Director: Seymour H. Weiss.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: January 7, 1999.
Brief description of amendment: The proposed amendment would allow
loading and handling of spent fuel transfer and storage casks in the
Trojan Fuel Building.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensees' analysis
against the standards of 10 CFR 50.92(c). The licensee's analysis is
summarized below:
The proposed changes would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
With the permanent cessation of operations, the number of potential
accidents was reduced to those types of accidents associated with the
storage of irradiated fuel and radioactive waste storage and handling.
Additional events were postulated for decommissioning activities due to
the difference in the types of activities that were to be performed.
The postulated accidents in the Defueled Safety Analysis Report (DSAR)
are generally classified as: (1) radioactive release from a subsystem
or component, (2) fuel handling accident and, (3) loss of spent fuel
decay heat removal capability. The postulated events described in the
Decommissioning Plan are grouped as: (1) decontamination,
dismantlement, and materials handling events, (2) loss of support
systems (offsite power, cooling water, and compressed air), (3) fire
and explosions, and (4) external events (earthquake, external flooding,
tornadoes, extreme winds, volcanoes, lightning, toxic chemical
release). These types of accidents are discussed below.
Radioactive release from a subsystem or component involves failure
of a radioactive waste gas decay tank (WGDT) or failure of a chemical
and volume control system holdup tank (HUT). For a failure of a WGDT,
the radioactive contents are assumed to be principally the noble gases
krypton and xenon, the particulate daughters of some of the krypton and
xenon isotopes and trace quantities of halogens. For the failure of a
HUT, the assumptions were full power operations with 1-percent failed
fuel, 40 weeks elapsed since power operation, and 60,000 gallons of
120 deg. F liquid released over a 2-hour period. However, the WGDT's
and HUT's are no longer active and have been emptied. Therefore, cask
loading and transfer activities cannot increase the probability of
occurrence of a failure or the consequence of a failure of the WGDT's
or HUT's.
The fuel handling accident involves a stuck or dropped fuel
assembly that results in damage of the cladding of the fuel rods in one
assembly and the release of gaseous fission products. Spent fuel
handling and loading will involve moving the spent fuel assemblies one
by one, from the Spent Fuel Pool to the baskets which will be located
in the Cask Loading Pit. The fuel handling equipment will be the same
as had been previously analyzed with the exception of special tools
which will be used to manipulate failed fuel. These special tools will
be similar in size and weight to the existing tools used for underwater
manipulation and therefore will not present a new hazard. In addition,
the same administrative controls and physical limitations imposed on
any fuel handling operation will be used for spent fuel loading and
handling. The potential release, 100 percent of gap noble gas, from a
fuel assembly is not affected (although the fission product inventory
in a fuel assembly continues to decrease with time). Thus there is no
increase in the probability of occurrence or consequences of a fuel
handling accident over what would be expected for any routine fuel
handling operation.
The loss of spent fuel decay heat removal capability involves the
loss of forced spent fuel cooling with and without concurrent Spent
Fuel Pool inventory loss. The only requirement to assure adequate decay
heat removal capability for the spent fuel is to maintain the water
level in the Spent Fuel Pool so that the fuel assemblies remain covered
(i.e. the capability to make up water to the Spent Fuel Pool must be
available when required). The potential events which could result in a
loss of spent fuel decay heat removal include external events
(explosions, toxic chemical, fires, ship collision with intake
structure, oil or corrosive liquid spills in the river, cooling tower
collapse, seismic events, severe meteorological events), and internal
events including Spent Fuel Pool makeup water system malfunctions
(Service Water System, electrical power, instrument air). Spent fuel
loading and handling will not require the use of explosive materials
(the gases used for electric arc welding are inert), toxic chemicals or
flammable materials (routine use of contamination control materials is
not considered to present a significant hazard). The probability of
other external events (e.g. cooling tower collapse) is not effected by
the spent fuel handling and loading activities inside the Fuel
Building. Spent fuel loading and handling activities will not directly
interface with the Spent Fuel Pool makeup water systems, therefore does
not affect their probability of failure. (The Cask Loading Pit will be
filled with borated water from the Spent Fuel Pool that will be cooled
by the Spent Fuel Cooling System, but use of this water in the Cask
Loading Pit does not increase the failure probability of
[[Page 9198]]
the Spent Fuel Pool or makeup water systems.) As described in the
licensees' safety evaluation, the safe load path and handling height
limitations will ensure that a load drop does not adversely affect the
Spent Fuel Pool or the makeup water systems. Therefore there is no
significant increase in the probability or consequences of a loss of
spent fuel decay heat removal capability.
The events postulated in the Decommissioning Plan are similar to
the DSAR with the exception of the decontamination, dismantlement, and
materials handling events. Decontamination events involve gross liquid
leakage from in-situ decontamination equipment (e.g. tanks) or
accidental spraying of liquids containing concentrated contamination.
Dismantlement events involve segmentation of components and structures,
or removal of concrete by rock splitting, explosives, or electric and/
or pneumatic hammers. Dismantlement events potentially result in
airborne contamination. Material handling events involve the dropping
of contaminated components, concrete rubble, filters, or packages of
particulate materials. Licensee administrative controls will be
implemented to ensure that spent fuel loading and handling activities
and decommissioning activities will not be performed concurrently if
they interact with each other and could increase the probability or
consequences of a postulated event of accident. Therefore, neither the
probability nor the consequences of decontamination, dismantlement, and
materials handling events will not be significantly increased.
The proposed changes would not create the possibility of a new or
different kind of accident from any accident previously evaluated. As
described in the licensees' safety evaluation the potential accidents
associated with fuel handling and loading were similar to fuel handling
accidents, material handling events and pressurized line break
previously analyzed. Additionally the potential consequences were a
small fraction of Environmental Protection Agency (EPA) Protective
Action Guides (PAG's). Therefore, fuel loading and handling does not
present new or different types of accidents.
The proposed changes do not involve a significant reduction in the
margin of safety. The Trojan Permanently Defueled Technical
Specifications (PDTS) contain four limiting conditions of operation
that address: (1) Spent fuel water level, (2) spent fuel pool boron
concentration, (3) spent fuel pool temperature, and (4) spent fuel pool
load restrictions. These PDTS will remain in effect as long as spent
fuel is stored in the Spent Fuel Pool, which is in accordance with
their applicability statements. The spent fuel loading and handling
activities will not affect these PDTS or their bases.
The Cask Loading Pit, where the spent fuel will be loaded into the
basket, is immediately adjacent to the Spent Fuel Pool. The gate
between the Cask Loading Pit and Spent Fuel Pool will be open to allow
transfer of spent fuel assemblies from storage racks in the Spent Fuel
Pool to the basket in the Cask Loading Pit. Opening the gate between
them will allow free exchange of water between the Cask Loading Pit and
the Spent Fuel Pool. The Cask Loading Pit will be filled with borated
water at approximately the same concentration and temperature as the
Spent Fuel Pool prior to opening the gate. This will maintain the
limiting conditions for operation for Spent Fuel Pool boron
concentration, temperature, and water level and the margin of safety
will not be affected.
Spent fuel loading and handling activities will involve lifting and
moving heavy loads (e.g. transfer cask, basket). Loads that will be
carried over fuel in the Spent Fuel Pool racks and the heights at which
they will be carried will be limited to preclude impact energies over
240,000 in-lbs if the loads were dropped. This is in accordance with
limiting condition for operation 3.1.4 ``Spent Fuel Pool Load
Restrictions.'' With this precaution, the limiting condition for
operation pertaining to load restrictions over the Spent Fuel Pool will
be satisfied and the margin of safety will be unaffected. The safe load
paths for heavy loads being lifted outside the Spent Fuel Pool will be
sufficiently far from the Spent Fuel Pool so as to not have an
interaction in the unlikely event of a load drop. In addition
mechanical stops and electrical interlocks on the Fuel Building
overhead crane will provide additional assurance that heavy loads are
not carried over the Spent Fuel Pool racks.
Based on the above, the spent fuel loading and handling activities
will not reduce the margin of safety.
Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Attorney for licensees: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S. W. Salmon Street, Portland, Oregon 97204.
NRR Project Director: Seymour H. Weiss.
Portland General Electric Company, et l., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: January 27, 1999.
Brief description of amendment: The proposed amendment would allow
unloading of spent fuel transfer casks in the Trojan Fuel Building.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The licensee's analysis is
summarized below:
The proposed changes would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
With the permanent cessation of operations, the number of potential
accidents was reduced to those types of accidents associated with the
storage of irradiated fuel and radioactive waste storage and handling.
Additional events were postulated for decommissioning activities due to
the difference in the types of activities that were to be performed.
The postulated accidents in the Defueled Safety Analysis Report (DSAR)
are generally classified as: (1) Radioactive release from a subsystem
or component, (2) fuel handling accident and, (3) loss of spent fuel
decay heat removal capability. The postulated events described in the
Decommissioning Plan are grouped as: (1) Decontamination,
dismantlement, and materials handling events, (2) loss of support
systems (offsite power, cooling water, and compressed air), (3) fire
and explosions, and (4) external events (earthquake, external flooding,
tornadoes, extreme winds, volcanoes, lightning, and toxic chemical
release). These types of accidents are discussed below.
Radioactive release from a subsystem or component involves failure
of a radioactive waste gas decay tank (WGDT) or failure of a chemical
and volume control system holdup tank (HUT). For a failure of a WGDT,
the radioactive contents are assumed to be principally the noble gases
krypton and xenon, the particulate daughters of some
[[Page 9199]]
of the krypton and xenon isotopes and trace quantities of halogens. For
the failure of a HUT, the assumptions were full power operations with
1-percent failed fuel, 40 weeks elapsed since power operation, and
60,000 gallons of 120 deg. F liquid released over a two hour period.
However, the WGDT's and HUT's are no longer active and have been
emptied. Therefore, cask loading and transfer activities cannot
increase the probability of occurrence of a failure or the consequence
of a failure of the WGDT's or HUT's.
The fuel handling accident involves a stuck or dropped fuel
assembly that results in damage of the cladding of the fuel rods in one
assembly and the release of gaseous fission products. Spent fuel cask
unloading will involve moving the spent fuel assemblies one by one,
from the baskets which will be located in the cask loading pit to the
spent fuel pool. The fuel handling equipment will be the same as had
been previously analyzed. In addition, the same administrative controls
on physical limitations imposed on fuel handling and fuel loading
operations will be used for fuel unloading. The potential release, 100
percent of noble gases within the gap, from a fuel assembly is not
affected (although the inventory in a radioactive stored fuel assembly
continues to decrease with time). Thus, there is no increase in the
probability of occurrence or consequences of a fuel handling accident
over what would be expected for any routine fuel handling operation or
loading of fuel into a cask.
The loss of spent fuel decay heat removal capability involves the
loss of forced spent fuel cooling with and without concurrent spent
fuel pool inventory loss. The only requirement to assure adequate decay
heat removal capability for the spent fuel is to maintain the water
level in the spent fuel pool so that the fuel assemblies remain covered
(i.e., the capability to make up water to the spent fuel pool must be
available when required). The potential events that could result in a
loss of spent fuel decay heat removal include external events
(explosions, toxic chemical, fires, ship collision with intake
structure, oil or corrosive liquid spills in the river, cooling tower
collapse, seismic events, and severe meteorological events), and
internal events including spent fuel pool makeup water system
malfunctions (service water system, electrical power, and instrument
air). Spent fuel cask unloading will not require the use of explosive
materials, toxic chemicals or flammable materials (routine use of
contamination control materials is not considered to present a
significant hazard). The probability of other external events (e.g.
cooling tower collapse) is not effected by the spent fuel unloading
activities inside the fuel building. Spent fuel cask unloading
activities will not directly interface with the spent fuel pool makeup
water systems, and therefore does not affect their probability of
failure. (The cask loading pit will be filled with borated water from
the spent fuel pool that will be cooled by the spent fuel cooling
system, but use of this water in the cask loading pit does not increase
the failure probability of the spent fuel pool or makeup water
systems). As described in the licensees' safety evaluation, the safe
load path and handling height limitations will ensure that a load drop
does not adversely affect the spent fuel pool or the makeup water
systems. Therefore, there is no significant increase in the probability
or consequences of a loss of spent fuel decay heat removal capability.
The events postulated in the Decommissioning Plan are similar to
the DSAR with the exception of the decontamination, dismantlement, and
materials handling events. Decontamination events involve gross liquid
leakage from in-situ decontamination equipment (e.g. tanks) or
accidental spraying of liquids containing concentrated contamination.
Dismantlement events involve segmentation of components and structures,
or removal of concrete by rock splitting, explosives, or electric and/
or pneumatic hammers. Dismantlement events potentially result in
airborne contamination. Material handling events involve the dropping
of contaminated components, concrete rubble, filters, or packages of
particulate materials. Licensee administrative controls will be
implemented to ensure that spent fuel cask unloading activities and
decommissioning activities will not be performed concurrently if they
interact with each other and could increase the probability or
consequences of a postulated event of accident. Therefore, neither the
probability nor the consequences of decontamination, dismantlement, and
materials handling events will be significantly increased.
The proposed changes would not create the possibility of a new or
different kind of accident from any accident previously evaluated. As
described in the licensee's safety evaluation the potential accidents
associated with fuel cask unloading were similar to fuel handling
accidents, material handling events and pressurized line break
previously analyzed. Additionally the potential consequences were a
small fraction of Environmental Protection Agency (EPA) Protective
Action Guides (PAGs). Therefore, fuel loading and handling does not
present new or different types of accidents.
The proposed changes do not involve a significant reduction in the
margin of safety. The Trojan Permanently Defueled Technical
Specifications (PDTS) contain four limiting conditions of operation
that address: (1) spent fuel pool water level, (2) spent fuel pool
boron concentration, (3) spent fuel pool temperature, and (4) spent
fuel pool load restrictions. These PDTS will remain in effect as long
as spent fuel is stored in the spent fuel pool, which is in accordance
with their applicability statements. The spent fuel cask unloading
activities will not affect these PDTS or their bases.
The cask loading pit, where the spent fuel will be unloaded from
basket, is immediately adjacent to the spent fuel pool. The gate
between the cask loading pit and spent fuel pool will be open to allow
transfer of spent fuel assemblies from the basket in the cask loading
pit to the storage racks in the spent fuel pool. Opening the gate
between them will allow free exchange of water between the cask loading
pit and the spent fuel pool. The cask loading pit will be filled with
borated water at approximately the same concentration and temperature
as the spent fuel pool prior to initial cask loading. This will
maintain the limiting conditions for operation for spent fuel pool
boron concentration, temperature, and water level and the margin of
safety will not be affected.
Spent fuel cask unloading activities may involve lifting and moving
heavy loads (e.g. transfer cask, basket). Loads that will be carried
over fuel in the spent fuel pool racks and the heights at which they
will be carried will be limited to preclude impact energies over
240,000 in-lbs if the loads were dropped. This is in accordance with
limiting condition for operation 3.1.4 ``Spent Fuel Pool Load
Restrictions.'' With this precaution, the limiting condition for
operation pertaining to load restrictions over the spent fuel pool will
be satisfied and the margin of safety will be unaffected. The safe load
paths for heavy loads being lifted outside the spent fuel pool will be
sufficiently far from the spent fuel pool so as to not have an
interaction in the unlikely event of a load drop. In addition,
mechanical stops and electrical interlocks on the fuel building
overhead crane will provide additional assurance that heavy loads are
not carried over the spent fuel pool racks.
[[Page 9200]]
Based on the above, the spent fuel cask unloading activities will
not reduce the margin of safety.
Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Attorney for licensees: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
NRR Project Director: Seymour H. Weiss.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: October 16, 1998, as supplemented
January 28, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to
relocate the Chemical Volume and Control System (CVCS) TS 3.2 from the
TSs to the IP3 Operational Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?
Response: Relocation (i.e., removal from TS) of TS 3.2, the
bases and the associated surveillances in Table 4.1-1 (items 12, 26,
and 27), Table 4.1-2 (item 2), and Table 4.1-3 (item 12) will not
involve a significant increase [in] the probability or consequences
of an accident since the relocation of the Technical Specifications
to administrative controls governed by 10 CFR 50.59 does not affect
the availability or function of charging and boric acid flow paths.
CVCS is not an initiator of an accident (the dilution event is
equipment malfunction that is manually terminated) and the proposed
change does not alter overall system operation, physical design,
system configuration, or operational setpoints. There will be no
significant increase in the consequences of an accident because the
required boration flow paths will continue to be available for
boration to the reactor coolant system.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: Relocation (i.e., removal from TS) of TS 3.2, the
bases and the associated surveillances in Table 4.1-1 (items 12, 26,
and 27), Table 4.1-2 (item 2), and Table 4.1-3 (item 12) will not
create the possibility of a new or different kind of accident from
any previously evaluated since it does not alter the overall system
operation, physical design, system configuration, or operational
setpoints. The plant systems for boration are operated in the same
manner as before and, consequently, the relocation does not
introduce any new accident initiators or failure mechanisms and does
not invalidate the existing dilution event response. The boration
function is not an accident initiator.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: Relocation (i.e., removal from TS) of TS 3.2, the
bases and the associated surveillances in Table 4.1-1 (items 12, 26,
and 27), Table 4.1-2 (item 2), and Table 4.1-3 (item 12) will not
involve a significant reduction in margin of safety. The relocation
is a change to the administrative controls that are used to assure
system availability and those administrative controls are governed
by 10 CFR 50.59. The manner in which the system is operated does not
change and there is no change to physical design, system
configuration, or operational setpoints. Previous analyses of system
malfunction remain unchanged. The current Technical Specification
does not meet the criteria in 10 CFR 50.36(c)(2)(ii) for inclusion
in the technical specifications.
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: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: December 30, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Limiting Condition for Operation
(LCO) 3.7.3 and Table 3.7.3-1. The proposed changes would modify the
flood protection actions required during periods of elevated river
water level.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS revisions related to flood protection TS Action
Statements involve no hardware changes and no changes to existing
structures, systems or components. The proposed changes to the flood
protection TS Action Statements ensure that the supported systems
can perform their required safety functions under worst case design
basis conditions, consistent with limitations imposed by other TS.
The proposed flood protection TS ACTION Statements ensure that the
plant is directed to enter a safe shutdown condition whenever the
capability to withstand worst case design basis conditions is
affected. Since the flood protection changes will still ensure that
the plant remains capable of meeting applicable design basis
requirements and retains the capability to mitigate the consequences
of accidents described in the [Hope Creek] HC [Updated Final Safety
Analysis Report] UFSAR, the proposed changes were determined to be
acceptable. As a result, these changes will neither increase the
probability of an accident previously evaluated nor increase the
radiological dose consequences of an accident previously evaluated.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to the flood protection TS contained in
this submittal will not adversely impact the operation of any safety
related component or equipment. Since the proposed changes involve
no hardware changes and no changes to existing structures, systems
or components, there can be no impact on the potential occurrence of
any accident due to new equipment failure modes. The resulting
operational limits imposed by the flood protection LCO ensure that
the plant can either perform its design basis safety functions or an
appropriately conservative shutdown action statement is entered.
Furthermore, there is no change in plant testing proposed in this
change request that could initiate an event. Therefore, these
changes 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.
The proposed changes for the flood protection TS retain the
plant's continued capability to withstand worst case design basis
conditions. The proposed flood protection TS ACTION Statements
ensure that the plant is directed to: (1) enter a safe shutdown
condition whenever the capability to withstand worst case design
basis conditions is lost; or (2) enter a conservatively short period
of continued operation when supported system redundancy is reduced.
Since the plant will still remain capable of meeting all applicable
design basis requirements and retaining the
[[Page 9201]]
capability to withstand worst case design basis events described in
the HC UFSAR, the proposed changes were determined to not result in
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: William M. Dean.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Description of amendment request: The proposed changes revise the
descriptive details of Technical Specification 4.7.1.2.1.a, regarding
performance testing of the Auxiliary Feedwater (AFW) pumps, to more
closely adhere to NUREG-1431, Improved Standard Technical
Specifications for Westinghouse Plants. This involves relocating the
surveillance-required numerical values for the AFW pump performance
test discharge pressure and flow rate to the South Texas Project
Updated Final Safety Analysis Report (UFSAR).
Date of amendment request: January 20, 1999.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change, which relocates descriptive details (i.e.,
numerical values for AFW pump discharge pressure and flow rate) of
the surveillance testing applicable to the AFW pumps, does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The affected AFW pump testing
pressure and flow descriptive details that are being removed from
SRs 4.7.1.2.1.a.1 and 4.7.1.2.1.a.2 are not related to any assumed
initiators of analyzed events and are not assumed to mitigate
accident or transient events. The requirement to perform testing on
a monthly, staggered basis is not altered by the proposed change,
and will remain in the Technical Specifications. The descriptive
details of the surveillance testing will be relocated from the
Technical Specifications to the USFAR and will be maintained
pursuant to 10CFR50.59. The proposed revised wording of SRs
4.7.1.2.1.a.1 and 4.7.1.2.1.a.2 (i.e., to verify the developed head
of each pump is greater than or equal to the required developed
head) and the relocation of pump testing details to the UFSAR is
consistent with the AFW pump test requirements in NUREG-1431. In
addition, the surveillance testing details are addressed in existing
surveillance procedures that are also controlled by 10CFR50.59 and
subject to the change control provisions imposed by plant
administrative procedures, which endorse applicable regulations and
standards. Therefore, this proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change relocates descriptive details (i.e.,
numerical values for AFW pump discharge pressure and flow rate) of
surveillance testing applicable to the AFW pumps, which do not meet
the criteria for inclusion in Technical Specifications as identified
in 10CFR50.36(c)(3). The requirement to perform testing on a
monthly, staggered basis is not altered by the proposed change, and
will remain in the Technical Specifications. Additionally,
relocation of the descriptive testing details is consistent with the
wording of the AFW pump test requirements in NUREG-1431, which does
not specify minimum numerical pressure and flow limits. The proposed
change does not involve a physical alteration of the plant (no new
or different type of equipment will be installed) or make changes in
the methods governing normal plant operation. The change will not
impose different requirements, and any future changes to these
relocated surveillance testing details or to the applicable
surveillance procedures will be evaluated per the requirements of
10CFR50.59. This change will not alter assumptions made in the
safety analysis and licensing basis. Therefore, this change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change, which relocates descriptive details (i.e.,
numerical values for AFW pump discharge pressure and flow rate) of
the surveillance testing applicable to the AFW pumps, will not
reduce a margin of safety since it has no impact on any safety
analysis assumptions. The requirement to perform AFW pump testing on
a monthly, staggered basis will not be altered by the proposed
change, and will remain in the Technical Specifications.
Furthermore, the proposed change will not affect the operability
requirements of the AFW system as delineated in Specification
3.7.1.2. Since any future changes to these relocated surveillance
testing details or to the applicable surveillance procedures will be
evaluated per the requirements of 10CFR50.59, there is no reduction
in a margin of safety. Finally, this proposed change is also
consistent with NUREG-1431, previously approved by the NRC Staff.
Revising the Technical Specifications to reflect the approved NUREG-
1431 content ensures no significant reduction in the margin of
safety. Therefore, this 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: John N. Hannon.
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: January 15, 1999 (TS 98-07).
Brief description of amendments: The proposed amendments would
change the Sequoyah (SQN) Technical Specification (TS) requirements by
adding a new action statement to TS 3.1.3.2, ``Position Indicating
Systems--Operating,'' that eliminates the need to enter TS 3.0.3
whenever two or more individual rod position indicators (RPIs) may be
inoperable per bank, while maintaining the appropriate overall level of
protection and adding flexibility to the initial determination of the
position of the non-indicating rod(s).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), 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 change to TS 3.1.3.2 does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The potential for the new action statement to
impact the probability or consequences of the safety analyses for
the plant lies only in the area of operator-exacerbated reactivity
events due to
[[Page 9202]]
a loss of RCCA [rod control cluster assembly] position indication.
RCCA events such as: One or more dropped RCCAs, a dropped RCCA
bank or a RCCA ejection (FSAR [Final Safety Analysis Report]
Sections 15.2.3 and 15.4.6, respectively) are not impacted since the
new action statement does not involve a design change. Events such
as: Uncontrolled RCCA bank withdrawal at power, statically
misaligned RCCA or withdrawal of a single RCCA (FSAR Sections
15.2.2, 15.2.3, and 15.3.6, respectively) involve, or potentially
involve, operator action and are of interest. The uncontrolled RCCA
bank withdrawal at power is an ANS [American Nuclear Society]
Condition II transient that has been analyzed using a positive
reactivity insertion rate greater than that for the simultaneous
withdrawal of the two control banks having the maximum combined
worth at maximum speed. Whether the event is caused by a failure in
the rod control system or by operator error has no effect on the
positive reactivity insertion rate assumed in the analysis. The
protection systems assumed in the analysis are unaffected since
there is no change to the design. Loss of the RPIS would not result
in more frequent control rod movement by plant operators. Therefore,
the new action statement would not affect the analysis of this event
and departure from nucleate boiling ratio (DNBR) design basis would
still be met.
The most severe misalignment situation, with respect to DNBR,
arises from cases in which one RCCA is fully inserted or where Bank
D is fully inserted to its insertion limits with one RCCA fully
withdrawn. For these cases, as discussed in FSAR Section 15.2.3.2,
the DNBR remains above the safety analysis limit values. Also, the
control bank insertion limit alarms remain available to warn
operators that bank insertion limits have been reached.
A compensatory action associated with this new action statement,
placing the control rods under manual control, addresses concerns
associated with automatic rod motion due to the rod control system
and inadvertent operator contribution to these events.
The worst-case event of those described above, the withdrawal of
a single RCCA, is an ANS Condition III event. It has been analyzed
in FSAR Section 15.3.6, assuming that operators ignore RCCA position
indication or that multiple rod control system failures occur. No
single electrical or mechanical failure in the rod control system
could cause the accidental withdrawal of a single RCCA from an
inserted bank at full power operation. The operator could
deliberately withdraw a single RCCA in the control bank. This
feature is necessary in order to retrieve an accidentally dropped
rod. This new action statement does not change the plant design;
therefore, there would be no change in the probability of the event
being induced by the unlikely, simultaneous electrical failures
(FSAR Section 7.7.2.2).
The change in the time to determine the position of the non-
indicating rods, indirectly with the movable incore detectors, does
not involve a design change nor does it affect the immediate
response of the operator to the event, therefore, it does not affect
the results of the analyses described above.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Since there is no change to the design associated with the
proposed change, it does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change involves a loss of the RPIS [Rod Position
Indication System] and establishes compensatory measures to maintain
control rod position consistent with the assumptions used in the
existing accident and transient analyses. The new action statement
provides sufficient time for troubleshooting while avoiding
unnecessary plant shutdowns per TS 3.0.3.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change to TS 3.1.3.2 does not involve a significant
reduction in a margin of safety. As discussed in Section IV.A above,
the results of the FSAR Chapter 15 safety analyses for the
applicable events, are not affected by the proposed changes.
Therefore, the safety margins demonstrated by these analyses remain
unchanged. The additional time to obtain the flux maps is consistent
with the 12-hour time frame allowed to verify shutdown margin when a
rod is misaligned from its group step counter height by more than
plus or minus 12 steps in TS 3.1.3.1 and remains within a shiftly
basis. Therefore, it does not reduce the margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Project Director: Cecil O. Thomas.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: January 29, 1999 (TSCR 211).
Description of amendment request: The proposed amendments reflect
changes to sections 15.6 and 15.7 of the Point Beach Nuclear Plant
(PBNP), Units 1 and 2, Technical Specifications (TS). The proposed
changes are considered administrative in nature and reflect personnel
title changes, an increase in minimum operating crew shift staffing,
relocation of the Manager's Supervisory Staff composition and
functional requirements to owner controlled documents, and revisions to
the procedure review and approval process.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a significant increase in
the probability or consequences of an accident previously evaluated.
These changes are administrative and therefore do not involve a
significant increase in the probability of an accident previously
evaluated because no such accidents are affected by the proposed
revisions. The proposed TS changes do not introduce any new accident
initiators since no accidents previously evaluated have as their
initiators anything related to the administrative changes described
above.
In addition, initiating conditions and assumptions are unchanged
and remain as previously analyzed for accidents in the PBNP Final
Safety Analysis Report. The proposed TS changes do not involve any
physical changes to systems or components, nor do they alter the
typical manner in which the systems or components are operated. All
Limiting Conditions [for] Operation, Limiting Safety System
Settings, and Safety Limits specified in the TS remain unchanged.
Therefore, these changes do not increase the probability of
previously evaluated accidents.
These changes do not involve a significant increase in the
consequences of an accident previously evaluated because the source
term, containment isolation or radiological releases are not being
changed by these proposed revisions. Existing system and component
redundancy and operation is not being changed by these proposed
changes. The assumptions used in evaluating the radiological
consequences in the PBNP Final Safety Analysis Report are not
invalidated; therefore, these changes do not affect the consequences
of previously evaluated accidents.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
These changes do not introduce nor increase the number of
failure mechanisms of a new or different type than those previously
evaluated since there are no physical changes being made to the
facility. The design and design basis of the facility remain
unchanged. The plant safety analyses remain unchanged. Therefore,
the possibility of a new or different kind of accident from any
accident previously evaluated is not introduced.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not involve a significant reduction in
a margin of safety.
[[Page 9203]]
The proposed changes do not involve a significant reduction in
the margin of safety because existing component redundancy is not
being changed by these proposed changes. There are no new or
significant changes to the initial conditions contributing to
accident severity or consequences, and safety margins established
through the design and facility license including the Technical
Specifications remain unchanged. Therefore, there are no significant
reductions in a margin of safety introduced by [these] proposed
amendment[s].
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: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 29, 1998.
Description of amendment request: This amendment would revise the
Wolf Creek Technical Specification (TS) Figures 3.4-2, 3.4-3, and 3.4-4
to incorporate revised reactor coolant system heatup and cooldown limit
curves and a revised cold overpressure mitigation system (COMS) power
operated relief valve (PORV) setpoint limit curve.
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.
Incorporating the revised heatup and cooldown pressure/
temperature limit curves and the COMS PORV setpoint limit curve into
the WCGS Technical Specifications does not affect the probability or
consequences of an accident previously evaluated.
The revised limit curves are calculated using the most limiting
RTNDT for the reactor vessel components and include a
radiation-induced shift corresponding to the end of the period for
which the curves are generated. The COMS PORV Setpoint Limit Curve
is calculated using the most limiting mass injection transient,
taking into account operation of the NCP [normal charging pump]
during shutdown modes. The changes do not affect the basis,
initiating events, chronology, or availability/operability of safety
related equipment required to mitigate transients and accidents
analyzed for WCGS.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Adopting the revised limit curves redefines the range of
acceptable operation for the Reactor Coolant System. This
redefinition is a result of the analysis of reactor vessel
surveillance specimens removed from the reactor in a continuing
surveillance program which monitors the effects of neutron
irradiation on the WCGS reactor vessel materials under actual
operating conditions. Included in the revised limit curves is
consideration for NCP operation during shutdown modes. Incorporating
these revised curves does not create the possibility of an accident
of a different type from any previously evaluated for WCGS.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The revision of these limit curves continues to maintain the
margin of safety required for prevention of non-ductile failure of
the WCGS reactor vessel during low temperature operation as required
by 10 CFR 50, Appendices G and H. The revised curves primarily
affect RCS [reactor coolant system] operation below 350 deg.F by
limiting the available pressure/temperature window for heatup and
cooldown. The revised limit curves compensate for the in-service
radiation induced embrittlement of the reactor vessel and accounts
for the requirement that the closure flange region temperature must
exceed the nil-ductility temperature by at least 120 deg.F when
pressure exceeds 20% of the preservice hydrostatic test pressure.
The revised COMS PORV Setpoint Limit Curve, which includes
consideration of NCP operation during shutdown modes, ensures
overpressure protection of the RCS and reactor vessel.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: January 12, 1999.
Description of amendment request: This license amendment request
proposes to revise Wolf Creek Generating Station (WCGS) Technical
Specification 3/4.7.5, Ultimate Heat Sink, to add a new action
statement. Specifically, the new action statement will require
verification of operability of the two residual heat removal (RHR)
trains, or initiation of power reduction with only one RHR train
operable, when the plant inlet water temperature is between 90 and 94
degrees Farenheit. The current TS requires shutdown when plant inlet
water temperature exceeds 90 degrees Farenheit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not involve any physical alteration of
plant systems, structures or components. The proposed change
provides an allowed time for the plant to continue operation with
plant inlet water temperature in excess of the current technical
specification limit of 90 degrees Fahrenheit, up to 94 degrees
Fahrenheit, which is less than the design limit of 95 degrees
Fahrenheit for plant components. The plant inlet water temperature
is not assumed to be an initiating condition of any accident
analysis evaluated in the updated safety analysis report (USAR).
Therefore, the allowance of a limited time for the water temperature
to be in excess of the current limit does not involve an increase in
the probability of an accident previously evaluated in the USAR. The
UHS [ultimate heat sink] supports operability of safety related
systems used to mitigate the consequences of an accident. Plant
operation for brief periods with plant inlet water temperature
greater than 90 degrees Fahrenheit up to 94 degrees Fahrenheit will
not adversely affect the operability of these safety-related systems
and will not adversely impact the ability of these systems to
perform their safety-related functions. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated in the USAR.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not involve any physical alteration of
plant systems,
[[Page 9204]]
structures or components. The temperature of the plant inlet water
being greater than 90 degrees Fahrenheit but less than or equal to
94 degrees Fahrenheit for a short period does not introduce new
failure mechanisms for systems, structures or components not already
considered in the USAR. Therefore, the possibility of a new or
different kind of accident from any accident previously evaluated is
not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change will allow an increase in plant inlet water
temperature above the current technical specification limit of 90
degrees Fahrenheit for the Ultimate Heat Sink, and delay the
requirement to shutdown the plant when the plant inlet water system
temperature limit is exceeded for 12 hours. The proposed change does
not alter any safety limits, limiting safety system settings, or
limiting conditions for operation, and the proposed temperature
increase will remain below the design limit cooling water input
value for safety-related equipment, except for the unlikely event of
a combination of a worst dam failure occurring with a loss of
coolant accident during a period of severe meteorological
conditions. Thus, the proposed change does not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
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.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of application for amendment: January 22, 1999.
Brief description of amendment: The amendment would revise
Technical Specification Surveillance Requirement 3.8.1.7 to better
match plant conditions during diesel generator (DG) testing by
clarifying which voltage and frequency limits are applicable during the
transient and steady state portions of the DG start.
Date of publication of individual notice in Federal Register:
February 1, 1999 (64 FR 4902).
Expiration date of individual notice: March 3, 1999.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401.
Illinois Power Company, Docket, No. 50-461, Clinton Power Station,
DeWitt County, Illinois
Date of application for amendment: January 20, 1999.
Brief description of amendment request: The proposed amendment
requests changes to the Technical Specification degraded voltage relay
setpoints.
Date of publication of individual notice in Federal Register:
January 28, 1999 (64 FR 4474).
Expiration date of individual notice: March 1, 1999.
Local Public Document Room location: Vespasian Warner Public
Library, 310 N. Quincy Street, Clinton, IL 61727.
PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit
2, Luzerne County, Pennsylvania
Date of amendment request: November 23, 1998.
Brief description of amendment request: The requested changes would
change the allowable values for both the core spray system and the low
pressure coolant injection system reactor steam dome pressure-low
functions.
Date of publication of individual notice in Federal Register:
February 1, 1999 (64 FR 4904).
Expiration date of individual notice: March 3, 1999.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendments: November 30, 1998, as
supplemented by letter dated January 8, 1999.
Brief description of amendments: The amendments relocate the
requirement for removal of the Reactor Protection System (RPS) shorting
links to the Updated Final Safety Analysis Report (UFSAR).
[[Page 9205]]
Date of issuance: February 8, 1999.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 170; 165 & 183; 180.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 7, 1999. (64 FR
1032).
The January 8, 1999, 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 February 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: March 27, 1998 (NRC-98-0033).
Brief description of amendment: The amendment revises technical
specifications (TS) 3.5.2 and 3.5.3 and the associated Bases, raising
the minimum water level for the core spray system in the condensate
storage tank (CST). The amendment also removes incorrect information
from TS 3.5.3 regarding water inventory in the CST reserved for the
high pressure coolant injection and reactor core isolation cooling
systems.
Date of issuance: February 8, 1999.
Effective date: February 8, 1999, with full implementation within
90 days.
Amendment No.: 131.
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19967).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 25, 1998, as supplemented by
letter dated November 30, 1998.
Brief description of amendment: The amendment changes the Appendix
A Technical Specifications (TSs) by modifying TS 3.9.8.1, ``Shutdown
Cooling and Coolant Circulation-High water Level,'' and TS 3.9.8.2,
``Shutdown Coolant Circulation-Low Water Level,'' to change the minimum
water level above the fuel assemblies seated in the reactor vessel at
which the Shutdown Cooling System (SDC) is required to be maintained
operable, or be in operation. Also TS 3.8.1.2, ``Electric Power Systems
A.C. Sources Shutdown,'' and appropriate Bases are revised to make
wording consistent with the TS 3.9.8.1 and 3.9.8.2.
Date of issuance: February 2, 1999.
Effective date: This license amendment is effective as of its date
of issuance, to be implemented within 60 days.
Amendment No.: 148.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 6, 1998 (63 FR
25109).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: October 27, 1998.
Brief description of amendment: This amendment revises TS 3/
4.8.2.3, ``Electrical Power Systems--DC Distribution--Operating,'' and
the associated bases. The surveillance requirements for battery testing
have been revised.
Date of issuance: February 9, 1999.
Effective date: February 9, 1999.
Amendment No.: 229.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64125).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 1999.
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.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: June 30, 1998, as supplemented
on December 9, 1998.
Brief description of amendment: This amendment revised Technical
Specification 3.1.7, ``Standby Liquid Control System,'' by increasing
the boron concentration in the Standby Liquid Control System for Cycle
8 fuel design.
Date of issuance: February 8, 1999.
Effective date: February 8, 1999.
Amendment No.: 97.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40562).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: October 27, 1998.
Brief description of amendments: The amendments revised Turkey
Point Units 3 and 4 Technical Specifications to add the qualifications
for the multi-discipline supervisor.
Date of issuance: February 3, 1999.
Effective date: February 3, 1999.
Amendment Nos.: 199 and 193.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69341).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
[[Page 9206]]
University, University Park, Miami, Florida 33199.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: July 21, 1998, as supplemented
October 6, December 16, and December 31, 1998.
Brief description of amendment: The amendment changes various
Reactor Protection System (RPS) and Engineered Safety Feature Actuation
System setpoints and allowable values; corrects the specified maximum
reactor power level limited by the high power level RPS trip; adds a
new Technical Specification associated with the automatic isolation of
steam generator blowdown; and makes several editorial changes to
correct various errors and to provide needed clarification. The
amendment also makes changes to the applicable Bases pages and expands
the Bases to discuss the new requirements for the automatic isolation
of steam generator blowdown. However, the staff has not completed its
evaluation of the requested change in the trip setpoint and allowable
values for the steam generator water level. This portion of the request
will be addressed later.
Date of issuance: February 8, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 226.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43208).
The October 6, December 16, and December 31, 1998, letters provided
clarifying information that did not change the scope of the July 21,
1998, application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: March 3, 1998, as supplemented
May 7, 1998.
Brief description of amendment: The amendment revises the Millstone
Unit 3 licensing basis by eliminating the requirement to have the
recirculation spray system directly inject into the reactor coolant
system following a design-basis accident, with the exception of loss-
of-coolant accident (LOCA) scenarios involving a long-term passive
failure. The Millstone Unit 3 licensing basis maintains the direct
injection requirement for scenarios, as a contingency, for situations
where it may be needed--as in the case of a LOCA with a long-term
passive failure or for beyond design-basis scenarios.
Date of issuance: January 20, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days from the date of issuance.
Amendment No.: 165.
Facility Operating License No. NPF-49: Amendment revised the
Millstone Unit 3 licensing basis.
Date of initial notice in Federal Register: March 25, 1998 (63 FR
14487).
The May 7, 1998, letter provided clarifying information that did
not change the scope of the March 3, 1998, application and the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment and final no
significant hazards consideration determination are contained in a
Safety Evaluation dated January 20, 1999.
No significant hazards consideration comments received: No public
comments received.
A petition to intervene was received from the Citizens Regulatory
Commission that was dismissed and terminated by the NRC Atomic Safety
Licensing Board.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: April 16, 1998.
Brief description of amendment: The amendment changes the Technical
Specifications to modify a testing requirement for the emergency diesel
generators.
Date of issuance: February 9, 1999.
Effective date: February 9, 1999.
Amendment No.: 187.
Facility Operating License No. DPR-64: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998, (63
FR 56256).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: June 16, 1998.
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 6 to relocate the Safety Review Committee
Reviews, Audits and Records from TS to the Quality Assurance Program
Section of the Final Safety Analysis Report.
Date of issuance: February 8, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 251.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 15, 1998 (63 FR
38204).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: October 19, 1998.
Brief description of amendment: This amendment eliminates
restrictions imposed by Technical Specification (TS) 3.0.4 for the
Filtration, Recirculation and Ventilation System
[[Page 9207]]
during fuel movement and CORE ALTERATION activities. Specifically, TS
Limiting Conditions for Operation 3.6.5.3.1 and 3.6.5.3.2 have been
revised to add a note stating that the provisions of TS 3.0.4 are not
applicable for initiation of handling of irradiated fuel in the
secondary containment and CORE ALTERATIONS provided that the plant is
in OPERATIONAL CONDITION 5, with reactor water level equal to or
greater than 22 feet 2 inches.
Date of issuance: February 4, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 113.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 4121).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 4, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: September 8, 1998, as
supplemented December 8, 1998.
Brief description of amendment: This amendment revised Appendix C,
``Additional Conditions,'' and will allow the performance of single
cell charging and the use of non-Class 1E single cell battery chargers,
with proper electrical isolation, for charging connected cells in
OPERABLE Class 1E batteries. The single cell chargers will be used to
restore individual cell parameters to the normal limits specified in
Technical Specification Table 4.8.2.1-1.
Date of issuance: February 9, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 114.
Facility Operating License No. NPF-57: This amendment revised
Appendix C of the license.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53954).
The December 8, 1998, supplement provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: April 28, 1998, as supplemented
September 29, 1998, and December 8, 1998.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.4.2.1 to replace the 1% setpoint
tolerance limit for safety/relief valves (SRVs) with a 3%
setpoint tolerance limit. In addition, the amendment revises TS 4.4.2.2
to state that all SRVs will be re-certified to meet a 1%
tolerance prior to returning the valves to service after setpoint
testing.
Date of issuance: February 10, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 115.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 17, 1998 (63 FR
33108).
The September 29, 1998, and December 8, 1998, supplements provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the scope of
the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 10, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: June 25, 1998, as supplemented
August 25, 1998, and December 15, 1998.
Brief description of amendment: The amendment revised Technical
Specification (TS) Surveillance Requirement 4.5.1.d.2.b by deleting the
requirement to perform in-situ functional testing of the Automatic
Depressurization System safety relief valves (SRVs) during startup
testing activities. The amendment also revised TS Surveillance
Requirement 4.4.2.1 such that the 18-month channel calibration for the
SRV acoustic monitors will no longer require an exception to the
provisions of TS 4.0.4, nor adjustments to SRV full open noise levels.
Date of issuance: February 10, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 116.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43212).
The August 25, 1998, and December 15, 1998, supplements provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the scope of
the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 10, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Public Service Electric & Gas Company, Docket No. 50-311, Salem Nuclear
Generating Station, Unit No. 2, Salem County, New Jersey
Date of application for amendment: October 12, 1998.
Brief description of amendment: This amendment allowed a one-time
extension of the Technical Specification (TS) surveillance interval to
the end of fuel Cycle 10 for certain TS surveillance requirements
(SRs). Specifically, the amendment extended the surveillance interval
in (a) SR 4.3.2.1.3 for the instrumentation response time testing of
each engineered safety features actuation system function, (b) SRs
4.8.2.3.2.f and 4.8.2.5.2.d for service testing of the 125-volt DC and
the 28-volt DC distribution system batteries, respectively, and (c) SR
4.8.2.5.2.c.2 for verification that the 125-volt DC battery connections
are clean, tight, and coated with anti-corrosion material. Because of
the length of the last outage and delays in restart, the SRs would have
become overdue prior to reaching the next refueling outage (2R10). The
SRs are to be completed during the 2R10 outage, prior to returning the
unit to Mode 4 (hot shutdown) upon outage completion.
Date of issuance: February 1, 1999.
[[Page 9208]]
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No.: 198.
Facility Operating License No. DPR-75: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59594).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 1, 1999.
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: November 14, 1997, as
supplemented by letters dated March 13, 1998, and November 10, 1998.
Brief description of amendments: The amendments would revise the
licensing basis as described in Section 3.5, ``Missile Protection,'' of
the Updated Final Safety Analysis Report to allow the use of NUREG-
0800, ``Standard Review Plan'' methodology in evaluating tornado-
generated missiles.
Date of issuance: February 9, 1999.
Effective date: February 9, 1999, to be implemented in the next
periodic update of the Updated Final Safety Analysis Report (UFSAR) in
accordance with 10 CFR 50.71(e) that occurs after 60 days of the date
of issuance.
Amendment Nos.: Unit 2--148; Unit 3--140.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the UFSAR.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68315).
The March 13, 1998, and November 10, 1998, supplemental letters
provided additional clarifying information and did not change the
original no significant hazards consideration determination. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated February 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
Southern Nuclear Operating Company, Inc., et al. Docket Nos. 50-424 and
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: June 26, 1998, as supplemented
by letters dated September 18 and November 30, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) as follows: (1) The Applicability of
Limiting Condition for Operation (LCO) 3.3.6, ``Containment Ventilation
Isolation Instrumentation,'' is revised to refer to TS Table 3.3.6-1;
the TS table is revised to add a column entitled ``Applicable Modes or
Other Specified Conditions.'' Then, the applicable modes for Manual
Initiation, Automatic Actuation Logic and Actuation Relays, and Safety
Injection are revised to include only Modes 1, 2, 3, and 4. Consistent
with this change, LCO 3.3.6, Condition C and Required Action C.2 are
revised to reflect that system level manual initiation and automatic
actuation are not required during core alterations and/or during
movement of irradiated fuel assemblies within containment. Appropriate
Bases changes are included to reflect the TS changes. (2) LCO 3.9.4 is
revised to allow the emergency air lock to be open during core
alterations and/or during movement of irradiated fuel assemblies within
containment. In addition, the LCO statement is revised to reflect that
containment ventilation isolation (CVI) would be accomplished by
manually closing the individual containment purge supply and exhaust
isolation valves as opposed to a system level manual or automatic
initiation, consistent with the proposed change to LCO 3.3.6.
Surveillance Requirement (SR) 3.9.4.2 is revised to reflect the change
to CVI. Appropriate Bases changes are included to reflect the TS
changes. (3) LCO 3.7.6 is revised to delete the words ``Redundant
CSTs'' from the title and LCO 3.7.6a is deleted. Appropriate Bases
changes are included to reflect the changes.
Date of issuance: January 29, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1--105; Unit 2--83.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53955). The supplement dated November 30, 1998, provided clarifying
information that did not change the scope of the application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 29, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: July 13, 1998, as supplemented
by letters dated December 16, 1998, and January 13, 1999.
Brief description of amendments: The amendments revise Technical
Specification Section 1.1, Definitions, for ``Engineered Safety Feature
[ESF] Response Time'' and ``Reactor Trip System [RTS] Response Time''
to provide for verification of response time for selected components
provided that the components and the methodology for verification have
been previously reviewed and approved by the NRC.
Date of issuance: February 8, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: 106 and 84.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53957).
The December 16, 1998, and January 13, 1999, letters provided
clarifying information that did not change the scope of the July 13,
1998, application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia.
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: October 29, 1998.
Brief description of amendments: Relocates portions of Technical
Specification 4.8.1.1.2.g requirements regarding maintenance of the
diesel generator fuel oil storage tank to the Technical Requirements
Manual.
[[Page 9209]]
Date of issuance: February 8, 1999.
Effective date: The license amendment is effective as of its date
of issuance, to be implemented within 30 days of issuance.
Amendment Nos.: Unit 1--Amendment No. 102; Unit 2--Amendment No.
89.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69347).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: November 16, 1998.
Brief description of amendments: The amendments revise the Sequoyah
Nuclear Plant Technical Specification (TS) emergency diesel generator
surveillance requirements. The U.S. Nuclear Regulatory Commission staff
has found the proposed changes to be acceptable.
Date of issuance: February 9, 1999.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: 242 and 232.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TSs.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66603).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power
Station, Franklin County, Massachusetts
Date of application for amendment: August 20, 1998.
Brief description of amendment: Revises Technical Specifications
(TS) through deletion of definition of SITE BOUNDARY, moves site map
from TS to Final Safety Analysis Report and deletion of an uneeded
reference to the site map.
Date of issuance: February 3, 1999.
Effective date: February 3, 1999.
Amendment No.: 150.
Possession Only License No. DPR-3: Amendment revised the Technical
Specifications.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53962). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Greenfield Community College,
1 College Drive, Greenfield, Massachusetts 01301.
Dated at Rockville, Maryland, this 17th day of February 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-4391 Filed 2-23-99; 8:45 am]
BILLING CODE 7590-01-P
