[Federal Register Volume 62, Number 117 (Wednesday, June 18, 1997)]
[Notices]
[Pages 33117-33142]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-15827]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person. This biweekly notice includes
all notices of amendments issued, or proposed to be issued from May 23,
1997, through June 6, 1997. The last biweekly notice was published on
June 4, 1997 (62 FR 30629).
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
[[Page 33118]]
take this action, it will publish in the Federal Register a notice of
issuance and provide for opportunity for a hearing after issuance. The
Commission expects that the need to take this action will occur very
infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By July 18, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. 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.
Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of amendment request: May 16, 1997
Description of amendment request: The modification involves
replacing the service water (SRW) heat exchangers with new plate and
frame heat exchangers having increased thermal performance capability.
The saltwater (SW) and SRW piping configuration will be modified as
necessary to allow proper fit-up to the new components. A flow control
scheme to throttle saltwater flow to the heat exchangers and the
associated bypass lines will be added.
[[Page 33119]]
Saltwater strainers with an automatic flushing arrangement will be
added upstream of each heat exchanger. The majority of the physical
work associated with this modification is restricted to the SRW pump
room.
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. Would not involve significant increase in the probability or
consequences of an accident previously evaluated.
None of the systems associated with the proposed modification
are accident initiators. The SW and SRW Systems are used to mitigate
the effects of accidents analyzed in the UFSAR [Updated Final Safety
Analysis Report]. The SW and SRW Systems provide cooling to safety-
related equipment following an accident. They support accident
mitigation functions; therefore, the proposed modification does not
increase the probability of an accident previously evaluated.
The proposed modification will increase the heat removal
capacity of the SRW System. The design provided under this activity
ensures that the safety features provided by the SW and SRW are
maintained, and in some instances enhanced; i.e., the availability
of important-to-safety equipment required to mitigate the
radiological consequences of an accident described in the UFSAR is
enhanced by the flexibility and increased thermal margin provided
with this design.
The redundant cooling capacity of the SW and SRW Systems have
not been altered. Furthermore, the proposed activity will not
change, degrade, or prevent actions described or assumed in any
accident described in the UFSAR. The proposed activity will not
alter any assumptions previously made in evaluating the radiological
consequences of any accident described in the UFSAR. Therefore, the
consequences of an accident previously evaluated in the UFSAR have
not increased.
Therefore, the proposed modification does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed activity involves modifying the SW and SRW System
components necessary to support the installation of new SRW heat
exchangers. None of the systems associated with this modification
are identified as accident initiators in the UFSAR. The SW and SRW
Systems are used to mitigate the effects of accidents analyzed in
the UFSAR. None of the functions required of the SRW or SW System
have been changed by this modification. This activity does not
modify any system, structure, or component such that it could become
accident initiator, as opposed to its current role as an accident
mitigator.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The safety design basis for the SW and SRW Systems is the
availability of sufficient cooling capacity to ensure continued
operation of equipment during normal and accident conditions. The
redundant cooling capacity of these systems, assuming a single
failure, is consistent with assumptions used in the accident
analysis.
The design, procurement, installation, and testing of the
equipment associated with the proposed modification are consistent
with the applicable codes and standards governing the original
systems, structures, and components. The design of instruments and
associated cabling ensures that physical and electrical separation
of the two subsystems is maintained. Common-mode failure is not
introduced by this activity. The equipment is qualified for the
service conditions stipulated for that environment. New cable and
raceways for this design will be installed in accordance with
seismic design requirements. The additional electrical load has been
reviewed to ensure the load limits for the vital 1E buses are not
exceeded. The circuits and components related to the control valves
control loops are safety-related, and are similar to those used for
the other safety-related flow control functions. The proposed
modification will not have any adverse effects on the safety-related
functions of the SW and SRW Systems.
For the above reasons, the existing safety bases have not been
altered by the proposed modification. This activity will not reduce
the margin of safety as it exists now. In fact, the margin of safety
has been increased by this activity due to the increase in the
thermal capacity of the dual train design and the increased
availability of safety-related components.
Therefore, this proposed modification does not significantly
reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Alexander W. Dromerick, Acting Director.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: April 23, 1997
Description of amendment request: The proposed changes would revise
surveillances 4.3.2.1.1.a, 4.3.2.1.4.b, 4.3.2.1.6.g, 4.3.2.1.10a,
4.3.2.1.10.b, and 4.7.3.b.3 to provide enhanced descriptions of the
tests being performed and the tested components.
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:
This change clarification does not involve a significant hazards
consideration for the following reasons:
(1) The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The components affected by the proposed changes are not
initiators of any accident previously evaluated. The proposed
changes to specification 4.3.2.1 items affect only the description
of the testing and make no changes in actual operation or testing.
The sample heat exchanger valves isolate on receipt of a Safety
Injection signal and that feature is unaffected by the additional
testing in the proposed change. Therefore, there is no increase in
the probability or consequence of a previously analyzed accident.
(2) The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes to the surveillance frequencies do not
involve physical alterations or additions to plant equipment or
alter the manner in which safety-related systems function or are
normally operated. The additional testing proposed for the sample
heat exchanger valves demonstrates the proper operation of a design
feature but does not operate the valve in any new way. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes to specification 4.3.2.1 clarify existing
testing. The additional testing for the CCW [component cooling
water] surge tank level instrumentation adds two components to the
surveillance documentation. Therefore, there is no reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
[[Page 33120]]
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: Mark Reinhart, Acting.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: April 14, 1997
Description of amendment request: The proposed amendments would
revise TS 3/4.3.8, ``Feedwater/Main Turbine Trip System Actuation
Instrumentation'' by changing the minimum channels required from 3 to
4. This change reflects a modification that is being installed to
correct a design deficiency that could have resulted in a failure to
trip the feedwater pumps and main turbine on high water level due to
the loss of one of the two instrument lines. The modification adds an
auxiliary contact to the trip system logic resulting in an additional
channel. The licensee is also proposing to modify the TS action
statements for inoperable channels to be similar to TS 3.3.1, ``Reactor
Protection System.''
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 because:
The proposed Technical Specification (TS) change will resolve
the common instrument line failure (break) from preventing reactor
high water level trip of Feedwater Pumps and Main Turbine. It will
not change the probability of occurrence of any accidents, because
this instrumentation is not an accident initiator. This
instrumentation resolves a potential concern regarding the results
of an instrument line break in conjunction with a Feedwater
Controller Failure Maximum Demand, which has been postulated and
analyzed separately, but are not required to be analyzed in
combination, as is described in Chapter 15 of the LaSalle UFSAR.
There will not be any increase in probability of feedwater transient
(postulated feedwater controller failure with assumed simultaneous
failure of one high level trip channel of Feedwater/Main Turbine
Trip Actuation Instrumentation), nor an instrument line break. The
design change associated with this TS change will prevent the
failure of the level 8 trip of Feedwater Pumps and Main Turbine due
to loss of common variable water leg of level instrument channels
``B'' and ``C''. Thus there is a slight increase [in] the
reliability of the high level trip by assuring that a single
instrument failure, including a failure of a sensing line, will not
prevent a level 8 trip. The Feedwater/Main Turbine Trip on Reactor
Vessel Water Level-High, Level 8, mitigates the consequences of the
transient, Feedwater Controller Failure Maximum Demand, due to the
main turbine trip with subsequent Turbine Stop Valve closure scram
and Reactor Recirculation Pump Trip. This limits the neutron flux
peak and fuel thermal transients so that no fuel damage occurs. MCPR
remains at or above the operating limit and peak centerline fuel
temperature increase is small. The consequences of an accident will
not increase, because the redundancy of the instrumentation portion
of the Trip Function is somewhat increased.
TS 3.3.8 limiting Condition for Operation (LCO) Actions b and c
are proposed to be changed to be similar to the LCO for TS 3.3.1,
Reactor Protection System Action b.1 to assure trip capability,
while being consistent with the allowed outage times of current TS
3.3.8. Also, the proposed action statements and allowed outage times
are consistent with LCO 3.3.2.2, ``Feedwater and Main Turbine High
Water Level Trip Instrumentation'', of NUREG 1433, Revision 1,
Standard Technical Specifications, General Electric Plants, BWR4,
dated April 1995. The limit on continued plant operation of 72 hours
in current Action c.1, is overly restrictive, since with one
inoperable channel tripped and one Operable channel, the Trip
Function is restored to the same status as current Action b.1 (one
more instrument failure will cause a failure to actuate on high
reactor water level). Therefore, although the proposed Actions are
increasing the allowed outage time for the case with only one
remaining Operable channel, from 72 hours to 7 days, the level of
protection for automatic trip capability is maintained except for a
2 hour period during which trip capability may not exist. In
addition, like current Action b.1, the proposed Actions assure that
the longest time that automatic trip capability failure due to
another instrument failure will exist is 7 days. Therefore, the
potential for failure of the Feedwater/Main Turbine trip on reactor
vessel high water level may be slightly increased, but is not
significant considering the non-safety-related Feedwater Pump and
Main Turbine trips are not and are not required to be single-failure
proof.
Based on the above, the proposed amendments will not increase
the probability or consequences of any accident previously
evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The Feedwater/Main Turbine trip is a non-safety function in the
non-safety-related feedwater system. The high water level trip is an
equipment protective action preventing main steam carry over in the
main steam from damaging the main turbine and preventing high
pressure liquid discharge through the safety relief valve discharge
lines in case of a feedwater transient due to a controller failure
to maximum demand. The trip system is not designed to any applicable
standards or regulatory guides or 10CFR50 Appendix A General Design
Criteria per UFSAR Table 7.1-2. The trip system is not designed nor
required to meet the single failure criteria. This is a non-safety/
non-divisional trip actuation required in Operating Condition 1, Run
Mode, such that high integrity of the trip is maintained. The
feedwater system is not required to mitigate the consequences of
accidents.
The design change associated with this TS change will increase
the reliability of the trip logic. This is accomplished by assuring
that a failure of a sensing line will not prevent or cause a level 8
trip. The failure of Feedwater/Main Turbine channel ``C'' trip
channel will not have any impact on the RCIC system nor Feedwater/
Main Turbine channels ``A'' & ``B'', because the added signal is
isolated by a safety-related relay. The 2 out of 3 logic for the
trip is maintained.
In addition, the changes to the action statements of the
specification do not allow a condition that could cause the
actuation instrumentation to fail in a different manner.
Based on the above, the proposed change will not create the
possibility of a new or different kind [of accident] from any
accident or transient previously evaluated.
(3) Involve a significant reduction in the margin of safety
because:
The proposed TS change will not prevent tripping of Feedwater/
Main Turbine or cause false trips. The existing 2 out of 3 logic
trip is maintained and does not affect existing failure modes or
introduce new failure modes. This change will prevent failure of
level 8 trip of Feedwater Pumps and Main Turbine upon loss of common
variable water leg for Reactor Vessel Water Level-High, Level 8,
instrument channels ``B'' & ``C'' and will slightly increase
reliability of the trip logic. Failure of the non-safety-related
trip logic will not impact any safety-related system, structure, or
component.
The changes to the TS LCO Action statements is consistent with
the existing actions, while minimizing the time that automatic trip
capability is not maintained. The change from 72 hours allowed
operation with one channel Operable and only one channel tripped to
7 days is consistent with the current allowed outage time for only
one channel inoperable and not tripped, so any change to the margin
of safety provided by the current action requirements is minor.
Based on the above, the proposed TS change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
[[Page 33121]]
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: May 1, 1997
Description of amendment request: This request changes Technical
Specification (TS) Surveillance Requirement (SR) 4.9.A.8.b by
clarifying the load value for the emergency diesel generator to be
equal to or greater than the largest single load and revise the
frequency and voltage requirements during performance of the test.
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 because of the
following:
The proposed changes represent a clarification of the intent of
the performance of the largest single emergency load rejection
surveillance for the diesel generator. These changes allow for
simulated testing that will more closely duplicate actual emergency
loading conditions. By removing the specific load value requirement
from the surveillance, the test can be performed using the actual
largest load in the same plant configuration that would exist during
an actual accident scenario. Verification of the steady-state
voltage and frequency within the required time limits provides
confidence that the diesel generator can successfully recover from
this transient. This provides greater assurance that the diesel
generator is capable of performing its intended design function
during an accident and the subsequent recovery. The changes to the
surveillance requirement will not significantly increase the
consequences of an accident previously evaluated.
The diesel generator's design function is to mitigate the
consequences of an accident by providing an independent onsite
source of alternate AC power with the capacity for operation of
systems required to shutdown the reactor and maintain it in a safe
shutdown condition until offsite power is restored. The diesel
generator and its associated subsystems are not assumed in any
safety analysis to initiate any accident sequence for Quad Cities
Station; therefore, the probability of an accident previously
evaluated is not increased by the proposed amendment.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The proposed changes do not create the possibility of a new or
different kind of accident previously evaluated for Quad Cities
Station. The changes revise the largest single emergency load
rejection surveillance test acceptance criteria for the diesel
generator. This load rejection transient for the diesel generator is
bounded by a previously performed accident analysis. This analysis
assumes the loss of one diesel generator due to loss of 125 VDC
control power for the duration of a LOCA combined with a LOOP. The
diesel generator's design function is to mitigate the consequences
of an accident by providing an independent onsite source of
alternate AC power with the capacity for operation of systems
required to shutdown the reactor and maintain it in a safe shutdown
condition until offsite power is restored. Only one diesel generator
is required to perform this function per unit. Performance of the
Surveillance Requirement as proposed provides greater assurance that
the diesel generator is capable of performing its intended design
function during an accident and the subsequent recovery. No
significant changes to existing testing or new modes of facility
operation are proposed by this change. The proposed changes maintain
at least the present level of operability. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
(3) Involve a significant reduction in the margin of safety
because:
The proposed amendment is required to ensure the diesel
generator is tested in accordance with the design basis
requirements. The changes represent a revision to the test
acceptance criteria for performance of the largest single emergency
load rejection surveillance for the diesel generator. This is a
possible transient for the diesel generator that is bounded by a
previously performed accident analysis. The proposed changes do not
adversely affect the capability of the diesel generator to perform
its design function. This function is to mitigate the consequences
of an accident by providing an independent onsite source of
alternate AC power with the capacity for operation of systems
required to shutdown the reactor and maintain it in a safe shutdown
condition until offsite power is restored. Performance of the
Surveillance Requirement as proposed provides greater assurance that
the diesel generator is capable of performing its intended design
function during an accident and the subsequent recovery. Existing
plant safety margins or the reliability of the equipment assumed to
operate in the safety analysis are not changed. The proposed changes
have been evaluated at Quad Cities and found to be acceptable for
use based on system design, safety analysis requirements and
operational performance. Since the changes 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: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: May 27, 1997.
Description of amendment request: The proposed amendments would
delete from the Technical Specifications (TS) of each unit the
specified minimum volume of borated water available to the Standby
Makeup Pump; the minimum volume is already specified in other parts of
the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. This amendment to the Catawba TS maintains the necessary
minimum volume of borated water available to mitigate a design basis
SSS [standby shutdown system] event through a 72 hour period.
Eliminating TS Surveillance 4.7.13.3a.2 does not increase the
probability or consequences of any previously evaluated accident,
since an adequate borated water source for the SMP [standby makeup
pump] is continued to be required by other existing TS.
(2) Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. This amendment to the Catawba TS continues to ensure that
the necessary minimum volume of borated water is available to
mitigate an SSS event. The SSS is required to mitigate certain
previously evaluated design basis fire, security, and other events.
This amendment does not create the possibility of a new or different
kind of accident from any accident previously evaluated. This
amendment changes the TS applicable to an accident mitigating
function and does not impact any accident initiator, either new,
different, or previously evaluated.
(3) Will the change involve a significant reduction in a margin
of safety?
No. This amendment continues to ensure that the necessary
minimum volume of borated water is available to mitigate an SSS
design basis event. The available minimum volume is maintained well
above the design basis requirement. Since the source of borated
water that is available to supply the SMP continues to be controlled
by existing TS (TS 3.7.13.3a.1 and 3.9.10), which both envelope the
current 112,320 gallons, sufficient volume has been and will
continue
[[Page 33122]]
to be present to meet design basis requirements. Therefore, no
reduction in a margin of safety will result from the changes
proposed in this amendment.
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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North
Carolina 28242-0001.
NRC Project Director: Herbert N. Berkow.
Duke Power Company, et al., Docket No. 50-414, Catawba Nuclear Station,
Unit 2, York County, South Carolina
Date of amendment request: May 27, 1997
Description of amendment request: The proposed amendment would
delete from the Technical Specification of Unit 2 requirements
regarding steam generator tube sleeving and repair. These requirements
are not applicable to the Westinghouse Model D5 steam generators used
by Unit 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. This amendment to the Catawba Unit 2 Technical
Specifications will have no impact on operation of the facility
since the change will delete steam generator repair methods that are
not applicable to the Catawba Unit 2 steam generators and have not
been used to repair the Catawba Unit 2 steam generators.
(2) Will the change create the possibility of a new or different
type of accident from any accident previously evaluated?
No. This amendment will delete steam generator repair methods
that are not applicable and have not been used. Therefore, the
proposed changes will not create the possibility of a new or
different accident.
(3) Will the change involve a significant reduction in the
margin of safety?
No. This amendment will delete steam generator repair methods
that are not applicable and have not been used. There will be no
impact on safety margins as a result of these changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North
Carolina 28242.
NRC Project Director: Herbert N. Berkow.
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf Nuclear
Station, Unit 1, Claiborne County, Mississippi.
Date of amendment request: May 7, 1997.
Description of amendment request: The amendment request would
eliminate selected response time testing (RTT) surveillance
requirements (SRs) from the Technical Specifications (TSs) for certain
components of the following systems: reactor protection system (SR
3.3.1.1.15), primary containment and drywell isolation instrumentation
(SR 3.3.6.1.8), and emergency core cooling system (SRs 3.5.1.8 and
3.5.2.7).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. No significant increase in the probability or consequences of
an accident previously evaluated results from this change.
The purpose of the proposed Technical Specification (TS) change
is to eliminate response time testing (RTT) requirements for
selected components in the Reactor Protection System (RPS), Primary
Containment and Drywell Isolation Instrumentation, and Emergency
Core Cooling System (ECCS) actuation instrumentation. The Boiling
Water Reactor Owners' Group (BWROG) has completed an evaluation
which demonstrates that [RTT] is redundant to the other TS-required
testing. These other tests, in conjunction with actions taken in
response to NRC Bulletin 90-01, ``Loss of Fill-Oil in Transmitters
Manufactured by Rosemount,'' and Supplement 1 [to the bulletin], are
sufficient to identify failure modes or degradations in instrument
response time and ensure operation of the associated systems within
acceptable limits. There are no known failure modes that can be
detected by [RTT] that cannot also be detected by the other TS-
required testing. This evaluation was documented in NEDO-32291-A,
``System Analyses for Elimination of Selected Response Time Testing
Requirements,'' October 1995. EOI [The licensee] has confirmed the
applicability of this evaluation to Grand Gulf Nuclear Power Station
(GGNS). In addition, EOI will complete the actions identified in the
NRC staff's Safety Evaluation of NEDO-32291-A.
Elimination of [ECCS] RTT during MODES 4 and 5 [(i.e., cold
shutdown and refueling, respectively)] is acceptable since there are
no design basis accidents in MODES 4 and 5 for which the ECCS High
Pressure Core Spray (HPCS) system is required to initiate within a
specified period of time. The requirement to maintain [ECCS]
OPERABLE during Modes 4 and 5 is preserved in the affected Technical
Specification. The ECCS RTT required by SR 3.5.1.8 (applicable
during MODES 1, 2, and 3, [or power operation, startup, and hot
shutdown, respectively]) is adequate to identify any operability
problems with the ECCS HPCS system. In addition, during MODES 4 and
5, the probability and consequences of accidents are reduced due to
the pressure and temperature limitations of these MODES.
Because of the continued application of other TS-required tests
such as channel calibrations, channel checks, channel functional
tests, and logic system functional tests, the response time of these
systems [listed in the first paragraph] will be maintained within
the acceptance limits assumed in the plant [(GGNS)] safety analyses
and required for successful mitigation of an initiating event. The
proposed changes do not affect the capability of the associated
systems to perform their intended function within their required
response time, nor do the proposed changes themselves affect the
operation of any equipment.
As a result, EOI has concluded that the proposed changes do not
involve a significant increase in the probability or the
consequences of an accident previously evaluated.
2. This change would not create the possibility of a new or
different kind of accident from any [accident] previously evaluated.
The proposed changes only apply to the testing requirements for
the components [in the systems] identified above and do not result
in any physical change to these or other components [in other
systems] or their operation. As a result, no new failure modes are
introduced. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. This change would not involve a significant reduction in a
margin of safety.
The current TS-required response times are based on the minimum
allowable values assumed in the plant [(GGNS)] safety analyses.
These analyses conservatively establish the margin of safety. As
described above, the proposed changes do not affect the capability
of the associated systems to perform their intended function within
the allowable response time used as the basis for the plant safety
analyses. The potential failure modes for the components within the
scope of this request were evaluated for
[[Page 33123]]
impact on instrument response time. This evaluation confirmed that,
with the exception of loss of fill-oil of Rosemount transmitters,
the remaining TS-required testing is sufficient to identify failure
modes or degradations in instrument response times and ensure
operation of the instrumentation within the scope of this request is
within acceptable limits. The actions taken in response to NRC
Bulletin 90-01 and Supplement 1 [to the bulletin] are adequate to
identify loss of fill-oil failures of Rosemount transmitters. As a
result, it has been concluded that plant and system response to an
initiating event will remain in compliance with the assumptions of
the [GGNS] safety analysis. Elimination of RTT for ECCS HPCS system
in MODES 4 and 5 does not reduce the margin of safety since there
are no design basis events in MODES 4 and 5 requiring this system to
respond in [a] specified period of time from onset of the event.
Response time testing required by SR 3.5.1.8 (applicable during
MODES 1, 2, and 3) is adequate to identify any equipment or
operability concerns).
Further, although not explicitly evaluated, the proposed changes
will provide an improvement to plant safety and operation by
reducing the time safety systems are unavailable, reducing the
potential for inadvertent safety system actuation, reducing plant
shutdown risk, limiting radiation exposure to plant personnel [that
would be due to the RTT], and eliminating the diversion of key
personnel resources to conduct unnecessary testing. Therefore, EOI
concluded that this request will result in an overall increase in
the margin of safety.
[Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 24, 1997.
Description of amendment request: The proposed amendment will
modify Technical Specification (TS) 3/4.7.4, Ultimate Heat Sink (UHS),
Table 3.7-3, by incorporating more restrictive dry cooling tower (DCT)
fan requirements, and it will change the wet cooling tower water
consumption in the TS Bases. This proposed amendment seeks to modify
the TS to be consistent with revised design basis calculations.
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: No.
The proposed change modifies the UHS TS by not allowing
operation with less than 12 DCT fans per DCT. This change is
necessary to adequately preserve the assumptions and limits of the
revised UHS design basis calculations. These calculations conclude
that the UHS is capable of dissipating the maximum peak heat load
resulting from the limiting design bases accident (i.e., large break
LOCA [large break loss of coolant accident]). The proposed change
does not directly affect any material condition of the plant that
could directly contribute to causing an accident or that could
contribute to the consequences of an accident. The proposed change
ensures that the mitigating effects of the UHS will be consistent
with the design basis analysis. 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: No.
The proposed change modifies the UHS TS to be consistent with
revised design basis calculations. The UHS TS is being modified to
eliminate operation with less than 12 DCT fans per DCT. The proposed
change will not alter the operation of the plant or the manner in
which the plant is operated. 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: No.
The proposed change modifies the UHS TS by not allowing
operation with less than 12 DCT fans per DCT. The proposed change
preserves the margin of safety by ensuring that the UHS will be
capable of dissipating the maximum design basis accident heat load
with adequate margin. Therefore, the proposed change will not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 24, 1997
Description of amendment request: The proposed amendment will
modify Technical Specifications (TS) 3.1.1.1, 3.1.1.2, 3.10.1 and
Figure 3.1-1 by removing the cycle dependent boron concentration and
boration flow rate from the Action Statements and removing the ``RWSP
at 1720 ppm'' curve from the figure. A change to TS Bases 3/4.1.1.1 and
3/4.1.1.2 has been included to support this change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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: No.
The Shutdown Margin requirements are determined by the reload
analysis performed every cycle. The Cycle 9 reload analysis has
determined that the current Shutdown Margin requirements are
acceptable. The proposed change eliminates the reference to 1720 ppm
in the Action Statement because 1720 is not adequate to ensure that
the Shutdown Margin requirements are met at the beginning of cycle.
The proposed Action Statement will continue to ensure that in the
event the Shutdown Margin requirements are not met, boration will be
immediately initiated to restore the Shutdown Margin to within
limits.
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: No.
The proposed change does not change the design or configuration
of the plant nor does it change how boration systems are operated
during normal or accident conditions. It
[[Page 33124]]
ensures that the Shutdown Margin requirements for accidents already
evaluated are promptly restored in the event that the requirements
are not met.
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: No.
The proposed change has not decreased the amount of Shutdown
Margin required. The current Shutdown Margin requirements have been
validated by the Reload Analysis for Cycle 9 and are adequate to
ensure that the reactor can be made subcritical from all operating
conditions, transients, and design basis events. The proposed change
ensures that the Shutdown Margin requirements are promptly restored
in the event that they are not met. As such, the proposed change
ensures that the current margin of safety is maintained.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 3, 1997
Description of amendment request: The proposed amendment requests a
change to the ACTION Requirements for Technical Specification 3/4.3.2
for the Safety Injection System Sump Recirculation Actuation Signal
(RAS). The proposed change will revise the allowed outage time for a
channel of RAS to be in the tripped condition from ``prior to entry
into the applicable MODE(S) following the next COLD SHUTDOWN'' to the
more restrictive time limit of 48 hours and adds a shutdown
requirement. Additionally, the 3.0.4 exemption is being removed from
the ACTION for the tripped condition. A change to the Technical
Specification Basis Section 3/4.3.2 has also been included.
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: No.
The proposed revision to the TS changes the allowed outage time
that a channel of RAS can be in the tripped condition from a maximum
of approximately 18 months when one channel is inoperable and 92
days when two channels are inoperable to 48 hours. If a channel were
in the tripped condition and a single failure occurred (that of one
other channel of RAS), a premature [refueling water storage pool]
RWSP low level signal would be generated. During a Design Basis
Accident with a containment high pressure condition causing the RWSP
outlet check valves to seat, this single failure would prevent the
contents of the RWSP from being injected into the reactor coolant
system and possibly resulting in failure of both trains of
[Emergency Core Cooling System] ECCS and [Containment Spray] CS.
Additionally, this would cause the [Low Pressure Safety Injection]
LPSI pumps to stop. Reducing the time that a channel of RAS can be
placed in the tripped condition will reduce the probability of this
scenario occurring during a Design Basis Accident. Since the allowed
outage time for a channel of RAS is being limited to 48 hours, this
is considered an off-normal operation and a single failure is not
required to be postulated during a Design Basis Accident in the
accident analysis. Reducing the time the channel can be placed in
the tripped condition and thus, the exposure time to this scenario,
would not be an accident initiator. The proposed change of being
more conservative in the time and condition limits in the TS will
not affect the assumptions, design parameters, or results of any
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: No.
The proposed change does not change the design or configuration
of the plant. The proposed change provides a more conservative
allowed outage time for the channel to be in the tripped condition.
There has been no physical change to plant systems, structures or
components nor will the proposed change reduce the ability of any of
the safety-related equipment required to mitigate Anticipated
Operational Occurrences or accidents. In fact, this change will
potentially increase the ability of safety related equipment to
perform its functions. The configuration required by the proposed
specification is permitted by the existing specification.
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: No.
The proposed change provides a more conservative allowed outage
time for the channel to be in the tripped condition. By reducing the
allowed outage time, the probability is reduced that a single
failure (that of a failure of one channel of RAS with one channel in
the tripped condition) would occur that would cause the suction to
be prematurely supplied by the Safety Injection System Sump,
potentially disabling the [High Pressure Safety Injection] HPSI and
CS pumps, and stopping of the LPSI pumps. Therefore, the only change
to the margin of safety would be an increase. Since the allowed
outage time for a channel of RAS is being limited to 48 hours, this
is considered an off-normal operation and a single failure is not
required to be postulated during a Design Basis Accident in the
accident analysis. The proposed changes do not affect the limiting
conditions for operation or their bases.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: May 9, 1997
Description of amendment request: The proposed amendment would
revise the definitions of Limiting Safety System Setting (LSSS) and
Instrument/Channel Calibration to reference a new program being added
to the Technical Specification (TS) (Section 6.13) for the control of
instrument setpoints.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 33125]]
consideration, which is presented below:
1. The proposed TS amendment will not significantly increase the
probability or consequences of any previously-evaluated accidents.
The proposed changes will not result in any direct hardware
changes. The change only adds a program to the TS for the
establishment and control of instrumentation setpoints that is
consistent with current DAEC [Duane Arnold Energy Center] practice.
The Instrument Setpoint Control Program is based upon a methodology
for the calculation of instrument setpoints that conforms to the
guidelines of Regulatory Guide 1.105, Rev. 2. The methodology
ensures that adequate margin exists between the normal plant
operating conditions and actual instrument setpoints to preclude
spurious plant/equipment trips. As a result, the proposed program
establishes the criteria for changes in instrument setpoints to
ensure that such changes will not result in unnecessary plant
transients. Consequently, the probability of any previously-analyzed
event is not increased by this change.
The role of the instrumentation and their associated setpoints
is in detecting and mitigating plant events and thereby limiting the
consequences of any previously-analyzed event. The LSSS[NTSP] and
corresponding LTPO[AV] have been developed in accordance with the
DAEC Instrument Setpoint Control Program criteria to ensure that the
instrumentation remains capable of mitigating events as described in
the safety analyses and that the results and consequences described
in the safety analyses remain bounding. Therefore, these changes do
not involve a significant increase in the consequences of an
accident previously evaluated.
2. The proposed changes will not create a new or different kind
of accident from those previously evaluated.
The proposed changes will not change the method or manner of
plant operation, in particular, calibration of TS-required
instrumentation. The use of the proposed TS program for the control
of changes to instrument setpoints does not impact safe operation of
the DAEC in that the design and safety analysis limits will continue
to be satisfied. The proposed TS program involves no system
additions or physical modifications, other than setpoint changes.
Any setpoint changes must conform to the criteria set forth in the
TS Instrument Setpoint Control Program. The instrument setpoints are
developed using a methodology that conforms to the guidelines
contained in Regulatory Guide 1.105, Rev. 2 to ensure the affected
instrumentation remains capable of mitigating accidents and
transients. Since operational methods remain unchanged and the
instrument setpoints have been evaluated to maintain the plant
within existing design basis criteria, no new or different type of
accident is created.
3. The proposed change will not result in a significant
reduction in any margin of safety.
The proposed TS program establishes the DAEC Instrument Setpoint
Control Program, which is based upon an NRC-approved methodology.
The program establishes the controls and criteria used to establish
and revise instrument setpoints. The setpoint calculations use the
uncertainties associated with the DAEC instrumentation and actual
DAEC physical data and operating practices to ensure the validity of
the resulting LTPO[AV] and LSSS[NTSP]. The methodology is based upon
combining the uncertainties of the associated channels and takes
into account calibration accuracy, instrument uncertainties, drift,
etc. The use of this methodology for establishing these setpoints
ensures that the design and/or safety analysis limits are not
exceeded in any transient or accident. Therefore, the proposed
change does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: Gail H. Marcus.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: May 9, 1997
Description of amendment request: The proposed amendment would
revise the definition of Limiting Condition for Operation (LCO) to
address the situation when systems, components, etc., are removed from
service or otherwise made inoperable during secondary modes of
operation, without requiring entry into the LCO actions.
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 TS amendment will not significantly increase the
probability or consequences of any previously evaluated accidents.
The proposed change merely adds criteria to the TS that are
consistent with the original design and licensing basis assumptions.
Operation in secondary modes of operation (such as surveillance
testing, torus cooling mode (test line-up) or Residual Heat Removal
system, and use of High Pressure Coolant Injection system or Reactor
Core Isolation Cooling system in test line-up for reactor pressure
control during transients) is assumed in the safety analysis report
(Ref. UFSAR Section 6.3.4.2.1 and 7.3.4.2). Because no changes in
actual equipment operation or testing are being made as part of this
change, the probability of any event which could be induced by such
operation or testing is not increased. Also, the change will ensure
that the time such equipment is removed from service is kept very
short in duration, either through existing TS Allowed Outage Time
(AOT) notes or administratively by procedures. This is consistent
with the assumption that the time in such secondary modes of
operation (i.e., safe test interval) is much shorter than the
allowable repair time (i.e., LCO time). Therefore, the proposed
change will not significantly increase the probability of any
previously evaluated accident.
The uniform application of the new TS criteria will further
ensure that the plant remains within the original design and
licensing basis assumptions for equipment removed from service
during secondary modes of operation. In particular, in the special
case where testing also removes the redundant system, train,
component, etc., from service, these criteria ensure that both
affected systems, trains, etc., are properly controlled. This is
acceptable because the time in such secondary modes of operation is
very short in duration, such that the impact on the overall
availability/reliability is insignificant. Therefore, the
consequences of any previously analyzed accident are not
significantly increased by this change.
2. The proposed changes will not create a new or different kind
of accident from those previously evaluated.
The proposed changes will not add a new or different kind of
accident because the plant will not be operated in a different way.
Operation in secondary modes has been previously evaluated and found
to be acceptable (Ref. General Electric reports APED-5736: Guideline
for Determining Safe Test Intervals and Repair Times for Engineered
Safeguards, and NEDO-10739: Methods for Calculating Safe Test
Intervals and Allowable Repair Times for Engineered Safeguard
Systems). The proposed change merely adds criteria to the TS that
are consistent with the assumptions contained within these
evaluations. Consequently, no new or different accidents are
postulated as a result of this proposed change.
3. The proposed change will not result in a significant
reduction in any margin of safety.
Because the criteria being added to the TS enforce the
assumptions of the evaluations that form the basis of the existing
TS (Ref. TS Bases 4.1, 4.2, and 3.5), the proposed change will not
result in 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 location: Cedar Rapids Public Library,
[[Page 33126]]
00 First Street, SE., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: Gail H. Marcus.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: December 20, 1996
Description of amendment requests: The proposed amendments would
reduce the frequency and scope of reactor coolant pump flywheel
inspections.
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:
We have evaluated the proposed T/S changes and have determined
they do not represent a significant hazards consideration based on
the criteria established in 10 CFR 50.92(c). Operation of Cook
Nuclear Plant in accordance with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
This change will reduce the frequency and scope of the
surveillance testing on the reactor coolant pump flywheels.
Operating power plants have been inspecting their flywheels for over
20 years with no flaws identified which affect flywheel integrity.
Past examinations performed to satisfy T/S 4.4.10.1 have not
revealed any cracking of flywheel plates at Cook Nuclear Plant.
Crack extension over a 60 year service life is negligible.
Structural reliability studies have shown that eliminating
inspections after 10 years of plant life will not significantly
change the probability of failure. Most flaws which could lead to
failure would be detected during preservice inspection or, at worst,
early in plant life, and crack growth over plant life is negligible.
As stated in the SER associated with WCAP-14535, assuming an initial
crack of 10% of the distance from the keyway to the flywheel outer
radius and a maximum fatigue crack growth, ASME margins would be
maintained during the 10-year inspection period. Therefore, the
change in test frequency will not endanger public health or safety.
For these reasons, it is our belief the proposed changes do not
involve a significant increase in the probability or consequences of
a previously evaluated accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The changes will not introduce any new modes of plant operation,
nor will any physical changes to the plant be required. Thus, the
changes will not create the possibility of a new or different kind
of accident from any accident previously analyzed or evaluated.
3. Involve a significant reduction in a margin of safety.
This change will reduce the frequency and scope of the
surveillance testing on the reactor coolant pump flywheels.
Operating power plants have been inspecting their flywheels for over
20 years with no flaws identified which affect flywheel integrity.
Past examinations performed to satisfy T/S 4.4.10.1 have not
revealed any cracking of flywheel plates at Cook Nuclear Plant.
Crack extension over a 60 year service life is negligible.
Structural reliability studies have shown that eliminating
inspections after 10 years of plant life will not significantly
change the probability of failure. Most flaws which could lead to
failure would be detected during preservice inspection or at worst
early in plant life, and crack growth over plant life is negligible.
As stated in the SER associated with WCAP-14535, assuming an initial
crack of 10% of the distance from the keyway to the flywheel outer
radius and a maximum fatigue crack growth, ASME margins would be
maintained during the 10-year inspection period. For these reasons,
it is our belief the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: April 30, 1997
Description of amendment request: The proposed amendment would
remove Technical Specifications (TSs) regarding meteorological
monitoring instrumentation in accordance with NRC Generic Letter (GL)
95-10, ``Relocation of Selected Technical Specification Requirements
Related to Instrumentation.'' Specifically, the amendment would delete
TS 3/4.3.7.3, ``Meteorological Monitoring Instrumentation,'' including
associated TS Tables 3/4.3.7.3-1, and TS Bases 3/4.3.7.3. The TS Index
would be revised to show these deletions. The deletion of TS 3.3.7.3
would also eliminate the requirement that a Special Report to be
submitted to the NRC pursuant to TS 6.9.2 when one or more
meteorological monitoring instrumentation channels is inoperable for
more than 7 days. The licensee states that the deleted requirements
would be relocated to the Updated Safety Analysis Report (USAR), except
that the special reporting requirement would be discontinued as the
licensee would continue to evaluate future inoperability of
meteorological instrumentation for reportability in accordance with 10
CFR 50.72 and 10 CFR 50.73. The licensee will also insert the word
``nominal'' in the relocated tables in the USAR to indicate that the
meteorological instrumentation elevations of 30 and 200 feet are
nominal elevations (this change would be made because, as the licensee
reported in LER 96-14, the actual locations of the air temperature
monitoring instruments are 26.8 feet and 194.8 feet and the actual
locations of the wind indicator (speed and direction) monitoring
instruments are 30.9 feet and 199.4 feet). As stated in GL 95-10, the
NRC staff has determined that meteorological monitoring instrumentation
does not serve such a primary protective function as to warrant
inclusion in the TS in accordance with 10 CFR 50.36 criteria. Thus, in
GL 95-10, the NRC staff established that relocation of the
meteorological instrumentation requirements to the USAR (whereby
changes are controlled by the licensee pursuant to 10 CFR 50.59) is
acceptable.
Basis for proposed no significant hazards consideration
determination:
As required by 10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
1. The operation of Nine Mile Point Unit 2 [NMP2], in accordance
with the proposed amendment, will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The NMP2 meteorological monitoring instrumentation is used to
provide data for use in radioactive dose assessment with respect to
routine or accidental releases of radioactive materials to the
atmosphere. The deletion of the special reporting requirements is an
administrative change. The subject special reporting requirements
serve no nuclear related protective function. The relocation of the
meteorological monitoring instrumentation requirements from the TSs
to the USAR, and the addition of the word nominal to the USAR and
tables, will not increase the probability of an accident since the
specification applies only to monitoring instrumentation. This also
is an administrative change and does not reduce the effectiveness of
the current instrumentation requirements. The meteorological
monitoring instrumentation
[[Page 33127]]
requirements are not precursors to any accident previously
evaluated. According to the NRC Staff (GL 95-10), the meteorological
monitoring instrumentation does not serve to ensure the plant is
operated within the bounds of initial conditions assumed in any
design basis accidents or transients previously evaluated, or that
the plant will be operated to preclude transients or accidents. In
addition, the meteorological monitoring instrumentation does not
function as part of the primary success path of a safety sequence
analysis used to demonstrate that the consequences of these events
are within the appropriate acceptance criteria. Therefore, the
proposed changes do not significantly increase the probability or
consequences of an accident previously evaluated.
2. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed deletion of the special reporting requirements is
an administrative change. The subject special reporting requirements
serve no nuclear related protective function. The proposed change
also removes meteorological monitoring instrumentation
specifications from the NMP2 TSs. This also is an administrative
change and does not reduce the effectiveness of the current
instrumentation requirements. The relocation of the meteorological
instrumentation requirements to the USAR, and the addition of the
word nominal to the USAR and tables, will not create the possibility
of a new or different kind of accident since the specification only
applies to monitoring instrumentation. The NRC Staff has concluded
in GL 95-10 that the provisions of the meteorological monitoring
instrumentation specifications are not related to dominant
contributors to plant risk. The NMP2 meteorological instrumentation
is used to provide data for use in radioactive dose assessment with
respect to routine or accidental releases of radioactive materials
to the atmosphere. Since no physical modification to the plant is
being performed, and no changes to actual plant operations are
required by the change, removal of the specifications from the NMP2
TSs will not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not involve a significant reduction in
a margin of safety.
The proposed deletion of the special reporting requirements is
an administrative change. The subject special reporting requirements
serve no nuclear related protective function. The proposed removal
of the instrumentation requirements from the NMP2 TSs is also an
administrative change and does not reduce the effectiveness of the
current instrumentation requirements. The relocation of the
meteorological instrumentation requirements to the USAR, and the
addition of the word nominal to the USAR and tables, will not
involve a reduction in a margin of safety since the specification
only applies to monitoring instrumentation. The instrumentation will
continue to meet the requirements of Regulatory Guide 1.23, and the
offsite dose calculations will continue to use the actual measured
elevation differences. In GL 95-10, the NRC Staff concluded (1) That
the meteorological monitoring instrumentation does not function as
part of the primary success path of a safety sequence analysis, and
(2) that the meteorological monitoring instrumentation
specifications are not related to dominant contributors to plant
risk. Therefore, the removal of the meteorological monitoring
instrumentation specifications from the NMP2 TSs will not result in
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 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Alexander W. Dromerick, Acting Director.
Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone
Nuclear Power Station, Unit No. 1, New London County, Connecticut
Date of amendment request: May 15, 1997
Description of amendment request: The proposed amendment would
revise Technical Specification Sections 3.1 and 4.1 ``Reactor
Protection System'' and the associated Bases to remove run mode
intermediate range monitor high flux/inoperative with the associated
average power range monitor downscale scram trip function and
incorporate editorial revisions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Millstone Nuclear Power Station, Unit No. 1,
in accordance with the proposed amendment, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
No physical change is being made to any systems or components
that are credited in the safety analysis, therefore there is no
change in the probability or consequences of any accident analyzed
in the UFSAR [Updated Final Safety Analysis Report].
The design basis accident applicable to the startup power region
is the Control Rod Drop Accident (CRDA). The UFSAR does not credit
the RUN Mode IRM [intermediate range monitor] High Flux/Inoperative
with the associated APRM [average power range monitor] downscale
scram Trip Function (IRM RUN Mode SCRAM) in the termination of this
accident. Accident mitigation is provided by the APRM 120% power
scram. Therefore, elimination of the IRM RUN Mode SCRAM function has
no adverse affect on previously evaluated accidents.
The Continuous Control Rod Withdrawal Error (CWE) transient is
terminated by the Rod Block Monitor (RBM) in the RUN Mode. The APRM
Reduced High Flux Scram provides the primary STARTUP Mode protection
in conjunction with the IRMs and limits the consequences of this
transient. Therefore, elimination of the IRM RUN Mode SCRAM function
has no effect on the consequences of this transient.
Clarification of the LCO [limiting condition for operation] RPS
[reactor protection system] Table aligns requirements with Limiting
Safety System Settings. Further revisions to LCO 3.1 Reactor
Protection System Table 3.1.1 and associated TS [technical
specification] bases to clarify APRM Trip Functions do not alter the
required trip functions. Deletion of RUN requirement and associated
Action B for Reduced High Flux fixes an editorial error introduced
in a previous amendment. This trip function is not effective with
the mode switch in the RUN position and removal does not alter the
neutron monitoring requirements credited in the accident analyses.
Adding a new surveillance to verify SRM [source range monitor]/
IRM/APRM overlap will enhance neutron monitoring during startups and
shutdowns and does not have an adverse affect on previously
evaluated accidents.
None of the proposed changes will affect any of the rod blocks
or other precursor events to either the CRDA or CWE. Therefore,
there is no change in the probability of any accident previously
analyzed.
2. The operation of Millstone Nuclear Power Station, Unit No. 1,
in accordance with the proposed amendment, will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes affect only the operations of neutron
monitoring and protective systems (IRM and APRM) which provide
indication and mitigation actions only. Operation of these systems
does not create the possibility for new precursors (such as
reactivity) which would introduce a new or different kind of
accident from any accident previously evaluated.
Additionally, the proposed changes do not affect the ability of
those systems required to mitigate previously evaluated accidents
during the modes they are credited.
3. The operation of Millstone Nuclear Power Station, Unit No. 1,
in accordance with the proposed amendment, will not involve a
significant reduction in a margin of safety.
The only scram function that the UFSAR takes credit for in the
mitigation of the limiting accident (control rod drop accident) is
the APRM 120% power scram which is not
[[Page 33128]]
affected by this change. Only the IRM RUN Mode SCRAM, for which the
UFSAR takes no credit in the termination of any analyzed event, is
removed by this change. Removal of the IRM RUN Mode SCRAM will avoid
the need to operate the plant in a ``half scram'' condition with the
potential for an inadvertent plant transient. For these reasons, the
change does not involve a significant reduction in a margin of
safety.
The Continuous Control Rod Withdrawal Error (CWE) transient is
terminated by the Rod Block Monitor (RBM) in the RUN Mode. When
initiated from the STARTUP Mode, the consequences of a CWE are
limited by the APRM Reduced High Flux scram in conjunction with the
IRM scram function. Therefore eliminating the TS requirement for the
IRM RUN Mode SCRAM will not reduce the margin of safety for this
transient.
Clarification of the LCO RPS Table aligns requirements with
Limiting Safety System Settings. Further revisions to LCO 3.1
Reactor Protection System Table 3.1.1 and associated TS bases to
clarify APRM Trip Functions do not alter the required trip
functions. Deletion of the RUN requirement and associated Action B
for Reduced High Flux corrects an editorial error introduced in a
previous amendment. This trip function is not effective with the
mode switch in the RUN position and removal does not alter the
neutron monitoring requirements credited in the accident analyses.
Adding a new surveillance to verify SRM/IRM/APRM overlap will
enhance neutron monitoring during startups and shutdowns and
consequently does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community--Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: May 20, 1997
Description of amendment request: This submittal supersedes the
January 22, 1996, submittal which was previously noticed on February
28, 1996 (61 FR 7554). The proposed change would relocate the
containment isolation valve (CIV) list, Table 3.6-2, from the Technical
Specifications to the Technical Requirements Manual (TRM). This change
would affect Technical Specification Sections 1.8.1.b, 4.6.1.1.a,
3.6.3.1, 4.6.3.1.1, and 4.6.3.1.2, and Basis Section 3/4.6.3. A note at
the bottom of Table 3.6-2 regarding the CIVs that are subject to
administrative controls is retained in the Technical Specifications by
relocating it to Sections 1.8.1.b and 3.6.3.1. This change is being
performed in accordance with Generic Letter 91-08, which provides
guidance for removal of component lists from the Technical
Specifications.
Additionally, a change to provide relief in the surveillance
requirement in Section 4.6.1.1.a is included. The change allows valves,
blind flanges, and deactivated automatic valves located inside the
containment and are locked, sealed, or otherwise secured in the closed
position to be verified closed prior to entering Mode 4 from Mode 5, if
not performed within the previous 92 days. The current requirements
check the valve position once per 31 days.
TS Bases Section 3/4.6.3 is updated to reflect the removal and
relocation of the CIV list to the TRM. Also, details of the
administrative controls for operating CIVs while in Modes 1 through 4
are added to Bases Section 3/4.6.3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to relocate the containment isolation valve
(CIV) list will not result in any hardware or equipment operating
changes. The proposed change is based on Generic Letter (GL) 91-08
and merely relocates the CIV table and removes all references to the
table. The relocation of the CIV table from the Technical
Specifications does not affect the operability requirements of any
of the listed valves. Technical Specifications will still continue
to require the CIVs to be operable. The LCO [limiting condition for
operation] and surveillance requirements for the valves will remain
in Technical Specifications. The CIV table will be relocated to the
Millstone Unit No. 2 Technical Requirements Manual (TRM), which is
controlled in accordance with 10 CFR 50.59. This change does not
alter the design, function, or operation of the valves involved.
Thus, there is no significant affect on the possibility or
consequences of any previously evaluated accident.
The change to Surveillance Requirement (SR) 4.6.1.1.a will allow
the valves, blind flanges and deactivated automatic valves located
inside the containment that are locked, sealed, or otherwise secured
in the closed position to be verified closed prior to entering Mode
4 from Mode 5, if not performed within the previous 92 days, instead
of the current 31 day requirement. This means that the surveillance
interval could be as long as the entire operating cycle, depending
on whether entry into Mode 5 is required during the cycle. The
change in the surveillance frequency (increase in time from 31 days
to not less than 92 days and only prior to entering Mode 4 from Mode
5) recognizes that these valves are operated under administrative
controls and probability of misalignment is low. This provides
adequate assurance that the containment function assumed in the
accident analysis will be maintained. Therefore, there is no
significant affect on the probability or consequences of any
previously evaluated accident. This proposed change to SR 4.6.1.1.a
is consistent with NUREG-1432 Standard Technical Specifications for
Combustion Engineering Pressurized Water Reactors Revision 1 (SR
3.6.3.4).
The information added to the Bases will provide additional
guidance to ensure the plant is operated correctly. This information
will not result in any new approaches to plant operation. Therefore,
there is not significant affect on the probability or consequences
of any previously evaluated accident.
These proposed changes do not alter the design, function, or
operation of the valves involved. Therefore, there is no significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The change to relocate the CIV list from the Technical
Specifications to the TRM will not impose any different operational
or surveillance requirements, nor will the change remove any such
requirements. Adequate control will be maintained. Furthermore, as
stated above, the proposed change does not alter the design,
function, or operation of the valves involved, and therefore does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The change to SR 4.6.1.1.a reduces the surveillance frequency
for valves, blind flanges and deactivated automatic valves located
inside the containment. It does not alter the design, function, or
operation of the valves. Therefore, it does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The information added to the Bases will provide additional
guidance to ensure the plant is operated correctly. This information
does not alter the design, function, or operation of the valves
involved. Therefore, it does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not reduce the margin of safety since
they have no impact on any safety analysis assumption. The proposed
changes do not decrease the scope
[[Page 33129]]
of equipment currently required to be operable or subject to
surveillance testing, nor do the proposed changes affect any
instrument setpoints or equipment safety functions.
The effectiveness of Technical Specifications will be maintained
since the change will not alter function or operability requirements
for any CIV. In addition, the relocation of the valve list is
consistent with the guidance provided in GL 91-08, and the change to
the surveillance interval is consistent with NUREG-0212 Standard
Technical Specifications for Combustion Engineering Pressurized
Water Reactors Revision 2 (LCO 3.6.1.1) and NUREG-1432 Standard
Technical Specifications for Combustion Engineering Pressurized
Water Reactors Revision 1 (LCO 3.6.3).
The information added to the Bases is consistent with the
guidance provided in GL 91-08 and NUREG-1432 Standard Technical
Specifications for Combustion Engineering Pressurized Water Reactors
Revision 1. The intent of the Technical Specifications will be met
since this information will not result in any new approaches to
plant operation.
Therefore, there is no 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: Learning Resources Center,
Three Rivers Community--Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: May 9, 1997
Description of amendment request: The proposed amendment would
revise the shutdown margin requirements and add Technical Specification
3/4.3.5 to provide the limiting condition for operation (LCO) and
surveillance requirements for the shutdown margin monitors. The
proposed amendment would also make administrative changes and revise
the associated Bases section.
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:
NNECO has reviewed the proposed changes in accordance with 10
CFR 50.92 and has concluded that the change does not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10 CFR 50.92(c) are not
satisfied. The proposed changes do not involve [an] SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed Technical Specification changes will revise the
current shutdown margin requirements for Modes 3, 4 and 5 in Figures
3.1-1, 3.1-2, 3.1-3, 3.1-4 and 3.1-5 and allow for additional
boration of the RCS [reactor coolant system] as directed by
Specification 3.3.5. The new Shutdown Margin requirements are based
on re-analyses of the Boron Dilution Event provided by Westinghouse.
In the re-analyses, assumptions were modified in order to justify
the operability of the Shutdown Margin Monitor for count rates which
are lower than currently allowed. The proposed Shutdown Margin
requirements for Modes 3, 4 and 5 will continue to assure that the
operator has a minimum of 15 minutes from the alarm to loss of
shutdown margin during an assumed Boron Dilution Event.
The proposed change also adds Technical Specification 3/4.3.5 to
provide the LCO and Surveillance Requirements for the Shutdown
Margin Monitors. LCO 3.3.5 refers to the Core Operating Limits
Report (COLR) which will specify the minimum count rate/alarm ratio
requirements in order to consider the Shutdown Margin Monitors
operable. The LCO also directs the additional boration of the RCS in
order to allow the Shutdown Margin Monitors to be considered
operable for lower count rates. Also, a footnote (**) is included in
Specification 3/4.3.5 to make the Specification treatment of the
valves consistent with the Mode 6 and Mode 5-loops drained
requirements.
Due to the addition of Technical Specification 3/4.3.5, the
related Bases information is added as BASES Section 3/4.3.5.
Additionally, the Bases information for the Shutdown Margin Monitors
which is currently in BASES Section 3/4.3.1 is moved to the added
BASES Section 3/4.3.5. This Bases information is also revised to be
consistent with the added Specification 3/4.3.5.
Also, due to the addition of Technical Specification 3/4.3.5,
the guidance related to the Shutdown Margin Monitor in Tables 3.3-1
and 4.3-1 is deleted to avoid redundancy.
Additionally, Section 3/4.1.2 of the Bases is revised so that it
refers to Figure 3.1-4 (Shutdown Margin for Mode 5/filled) instead
of Figure 3.1-5 (Shutdown Margin for Mode 5/drained). This change
will make the Bases consistent with the ACTION statement
requirements of Technical Specifications 3.1.2.2 and 3.1.2.6.
Finally, Reference 12 (NUSCO-152, Addendum 4) is added to the
list of references in Section 6.9.1.6.b. The addition of this
reference is considered administrative and is not related to or
required by the changes proposed for the Shutdown Margin
requirements or Shutdown Margin Monitors.
The new requirements for increased Shutdown Margin (Figures 3.1-
1 to 3.1-5) and additional boration (LCO 3.3.5) continue to assure
that the operator will have a response time of at least 15 minutes
to mitigate the consequences of a Boron Dilution Event. The
implementation of the new requirements does not alter the alignment
of any plant equipment and therefore, the change cannot increase the
probability or consequences of any previously analyzed accident.
The proposed changes will not adversely affect the assumptions
or results of other FSAR [Final Safety Analysis Report] accident
analysis and it is concluded that this change is safe. The changes
do not adversely affect any equipment credited in the safety
analysis.
Based upon the re-analyses of the boron dilution event, revised
plant operating requirements (shutdown margin) are generated to
maintain the required operator action time. Therefore, there is no
effect on the probability of occurrence or consequences of
previously evaluated accidents.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed Shutdown Margin requirements for Modes 3, 4 and 5
(Figures 3.1-1 to 3.1-5 and additional boration as per Specification
3.3.5) will continue to assure that the operator has a minimum of 15
minutes from the alarm to loss of shutdown margin during an assumed
Boron Dilution Event. Additionally, the use of these revised
requirements allows the Shutdown Margin Monitor to be considered
operable for count rates which are lower than currently allowed.
The changes do not introduce any new failure modes or
malfunctions since the changes implement revised, more conservative
plant operating requirements (shutdown margin) which are based on
re-analyses of the Boron Dilution Event. Also, the changes do not
eliminate any existing requirements.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed Shutdown Margin requirements for Modes 3, 4 and 5
(Figures 3.1-1 to 3.1-5 and additional boration as per Specification
3.3.5) will continue to assure that the operator has a minimum of 15
minutes from the alarm to loss of shutdown margin during an assumed
Boron Dilution Event. Additionally, the use of these revised
requirements allows the Shutdown Margin Monitor to be considered
operable for count rates...which are lower than currently allowed.
The re-analyses of the Boron Dilution Event demonstrated that
the required
[[Page 33130]]
operator action time is maintained. As such, the re-analyses will
become the ``analysis of record'' for the Boron Dilution Event in
Modes 3, 4 and 5. The Boron Dilution Event analysis is documented in
FSAR Chapter 15.4.6.
The re-analyses of the Boron Dilution Event and the proposed
revisions to the Technical Specifications do not adversely affect
the results of the current FSAR accident analysis and therefore, it
is concluded that this change is safe. Additionally, the change does
not adversely affect any equipment credited in the safety analysis.
The changes do not have an adverse impact on the protective
boundaries and there is no reduction in the margin of safety as
specified in the Technical Specifications. Thus, this proposed
change does not involve a significant reduction in the margin of
safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed changes do not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: May 14, 1997
Description of amendment request: Technical Specification
Surveillance Requirement 4.8.2.1.c.4 requires that each battery charger
be tested to verify that it can supply a specified current at 125
volts. The proposed amendment would increase the required test voltage.
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:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve [an] SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes to Technical Specification Surveillance
4.8.2.1.c.4 to increase the required test voltage for the battery
chargers from 125 volts to greater than or equal to 132 volts is
consistent with the design criteria of the chargers and performing
battery charger surveillance testing does not significantly increase
the probability of an accident previously evaluated. The proposed
changes to increase the required test voltage for the battery
chargers provides the necessary assurance that the battery chargers
will function as required in previous evaluations and does not
significantly increase the consequence of an accident previously
evaluated.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to Technical Specification Surveillance
4.8.2.1.c.4 to increase the required test voltage for the battery
chargers from 125 volts to greater than or equal to 132 volts does
not change the operation of the battery chargers during normal or
accident evaluations.
Therefore, the proposed revision does not create the possibility
or a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to Technical Specification Surveillance
4.8.2.1.c.4 to increase the required test voltage for the battery
chargers from 125 volts to greater than or equal to 132 volts
provides assurance that the battery chargers are capable of
supplying the largest combined demands of the various steady state
loads, plus the current required to recharge its battery, which has
undergone a duty cycle discharge, to its fully charged condition in
less than 24 hours.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: March 3, 1997 as supplemented by letter
dated May 5, 1997. The May 5, 1997, supplement revised the proposed no
significant hazards consideration entirely
Description of amendment request: The proposed changes to the Hope
Creek (HC) Technical Specifications (TSs) would: (1) Change TS 3/4.3.1,
``Reactor Protection System Instrumentation,'' TS 3/4.3.2, ``Isolation
Actuation Instrumentation,'' and TS 3/4.3.3, ``Emergency Core Cooling
System Actuation Instrumentation'' to include additional information
concerning response time testing; (2) Change TS 4.0.5 to reference
inservice inspection and test requirements; (3) Change TS 3/4.6.1,
``Primary Containment,'' and associated Bases to reflect a design
modification; (4) Change TS 3/4.7.7, ``Main Turbine Bypass System,'' to
specify a new operability requirement; and (5) Change the Bases for TS
3/4.8, ``Electrical Power Systems.''
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 changes for the TS related to response time testing
reflect testing methodologies that were approved by the NRC in
Amendment No. 85 to the Hope Creek TS. These proposed TS revisions
involve: (1) no hardware changes; (2) no significant changes to the
operation of any systems or components in normal or accident
operating conditions; and (3) no changes to existing structures,
systems or components. Therefore, these changes will not increase
the probability of an accident previously evaluated. Since the plant
systems associated with these proposed changes will still be capable
of: (1) meeting all applicable design
[[Page 33131]]
basis requirements; and (2) retain the capability to mitigate the
consequences of accidents described in the HC [Updated Final Safety
Analysis Report] UFSAR, the proposed changes were determined to be
justified. As a result, these changes will not involve a significant
increase in the consequences of an accident previously evaluated.
The proposed changes to Surveillance Requirement 4.0.5 do not
alter the current requirements for the Hope Creek inservice
inspection and inservice testing programs and are considered to be
editorial in nature. These proposed TS revisions involve: (1) no
hardware changes; (2) no significant changes to the operation of any
systems or components in normal or accident operating conditions;
and (3) no changes to existing structures, systems or components.
Therefore, these changes will not increase the probability of an
accident previously evaluated. Since the plant systems associated
with these proposed changes will still be capable of: (1) Meeting
all applicable design basis requirements; and (2) retain the
capability to mitigate the consequences of accidents described in
the HC UFSAR, the proposed changes were determined to be justified.
As a result, these changes will not involve a significant increase
in the consequences of an accident previously evaluated.
The proposed changes to the drywell and suppression chamber
purge system are being made to justify design modifications to that
system. As discussed in NRC Notice of Violation 50-354/96-10-01,
this design modification replaced isolation valves containing
resilient material seals with metal seated valves under 10CFR50.59.
As a result of this modification, a 24 month frequency has been
implemented to perform Type C tests on these new metal seated
valves. PSE&G has concluded that the 24 month frequency is
appropriate for the new valves since: (1) This frequency is imposed
by Surveillance Requirement 4.6.1.2.d, which is applicable to
similar containment isolation valves in Table 3.6.3-1 that penetrate
the primary containment; and (2) concerns raised about severe
environment-induced degradation and frequent use for the previously
installed resilient seal material valves are not applicable to the
replacement metal seat valves. PSE&G has concluded that the valve
modification was an enhancement to the Hope Creek design that did
not impact the isolation capability of the drywell and suppression
chamber purge system. No significant changes were made to the
operation of these valves in normal or accident operating
conditions. As a result, these changes will not increase the
probability of an accident previously evaluated. Since the plant
systems associated with these proposed changes will still be capable
of: (1) Meeting all applicable design basis requirements; and (2)
retain the capability to mitigate the consequences of accidents
described in the HC UFSAR, the proposed changes were determined to
be justified. As a result, these changes will not involve a
significant increase in the consequences of an accident previously
evaluated.
The proposed changes to [Limiting Condition for Operation] LCO
3.7.7 establish consistent and appropriate requirements for main
turbine bypass valve operability requirements. These changes do not
impact the assumptions contained in these UFSAR analyses since they
do not change the manner in which Hope Creek is currently permitted
to operate. Since the ACTION Statement for LCO 3.7.7 already allows
indefinite continued operation below 25% of RATED THERMAL POWER with
an inoperable main turbine bypass valve system, the proposed
modification to the APPLICABILITY statement for this LCO does not
involve: (1) Hardware changes; (2) significant changes to the
operation of any systems or components in normal or accident
operating conditions; or (3) changes to existing structures, systems
or components. Therefore these changes will not increase the
probability of an accident previously evaluated. Since the plant
systems associated with these proposed changes will still be capable
of: (1) meeting all applicable design basis requirements; and (2)
retain the capability to mitigate the consequences of accidents
described in the HC UFSAR, the proposed changes were determined to
be justified. As a result, these changes will not involve a
significant increase in the consequences of an accident previously
evaluated.
The proposed changes to the HC emergency diesel generator (EDG)
TS Bases [Change 5--Bases for TS 3/4.8, ``Electrical Power
Systems''] include information contained in the Safety Evaluation
Report for Technical Specification Amendment No. 75. This
information concerns the bases for the allowed-outage-time (AOT) for
the C and D EDGs. Concerning the revisions to planned C and D EDG
outages, PSE&G believes that implementation of 10CFR50.65
requirements to monitor EDG unavailability will provide an
acceptable and more clearly defined method for maintaining EDG
availability within acceptable limits. As stated in PSE&G's letter
LR-N97167, dated March 21, 1997, Hope Creek will not plan C or D EDG
outages that exceed 72 hours if the total unavailability of the EDG
will be greater than 720 hours on a 12 month rolling basis. The
proposed TS revisions involve: (1) no hardware changes; (2) no
significant changes to the operation of any systems or components in
normal or accident operating conditions; and (3) no changes to
existing structures, systems or components. Therefore these changes
will not increase the probability of an accident previously
evaluated. Since the plant systems associated with these proposed
changes will still be capable of: (1) Meeting all applicable design
basis requirements; and (2) retain the capability to mitigate the
consequences of accidents described in the HC UFSAR, the proposed
changes were determined to be justified. As a result, these changes
will not involve a significant increase in the consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes for the TS related to response time testing
reflect testing methodologies that were approved by the NRC in
Amendment No. 85 to the Hope Creek TS and are being made to clarify
the licensing basis for performing response time testing. The
proposed changes will not adversely impact the operation of any
safety related component or equipment. Since the proposed changes
involve: (1) No hardware changes; (2) no significant changes to the
operation of any systems or components; and (3) no changes to
existing structures, systems or components, there can be no impact
on the occurrence of an accident previously evaluated. 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.
The proposed changes to Surveillance Requirement 4.0.5 do not
alter the current requirements for the Hope Creek inservice
inspection and inservice testing programs and are considered to be
editorial in nature. The proposed changes will not adversely impact
the operation of any safety related component or equipment. Since
the proposed changes involve: (1) No hardware changes; (2) no
changes to the operation of any systems or components; and (3) no
changes to existing structures, systems or components, there can be
no impact on the occurrence of an accident previously evaluated.
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.
The proposed changes to the drywell and suppression chamber
purge system are being made to justify design modifications to that
system. As discussed in NRC Notice of Violation 50-354/96-10-01,
this design modification replaced isolation valves containing
resilient material seals with metal seated valves under 10 CFR
50.59. As a result of this modification, a 24 month frequency has
been implemented to perform Type C tests on these new metal seated
valves. PSE&G has concluded that the 24 month frequency is
appropriate for the new valves since: (1) This frequency is imposed
by Surveillance Requirement 4.6.1.2.d, which is applicable to
similar containment isolation valves in Table 3.6.3-1 that penetrate
the primary containment; and (2) concerns raised about severe
environment-induced degradation and frequent use for the previously
installed resilient seal material valves are not applicable to the
replacement metal seat valves. PSE&G has concluded that the valve
modification was an enhancement to the Hope Creek design that did
not impact the isolation capability of the drywell and suppression
chamber purge system. Since the proposed changes will not adversely
impact the operation of any safety related component or equipment,
there can be no impact on the occurrence of any accident.
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
[[Page 33132]]
new or different kind of accident from any accident previously
evaluated.
The proposed changes to LCO 3.7.7 establish consistent and
appropriate requirements for main turbine bypass valve operability
requirements. These changes do not impact the assumptions contained
in these UFSAR analyses since they do not change the manner in which
Hope Creek is currently permitted to operate. Since the ACTION
Statement for LCO 3.7.7 already allows indefinite continued
operation below 25% of RATED THERMAL POWER with an inoperable main
turbine bypass valve system, the proposed modification to the
APPLICABILITY statement for this LCO does not involve: (1) hardware
changes; (2) significant changes to the operation of any systems or
components in normal or accident operating conditions; or (3)
changes to existing structures, systems or components. The proposed
changes will not adversely impact the operation of any safety
related component or equipment. Since the proposed changes involve:
(1) no significant hardware changes; (2) no significant changes to
the operation of any systems or components; and (3) no changes to
existing structures, systems or components, there can be no impact
on the occurrence of any accident. 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.
The proposed changes to the HC emergency diesel generator (EDG)
TS Bases [Change 5--Bases for TS \3/4\.8, ``Electrical Power
Systems''] include information contained in the Safety Evaluation
Report for Technical Specification Amendment No. 75. This
information concerns the bases for the allowed-outage-time (AOT) for
the C and D EDGs. Concerning the revisions to planned C and D EDG
outages, PSE&G believes that implementation of 10CFR50.65
requirements to monitor EDG unavailability will provide an
acceptable and more clearly defined method for maintaining EDG
availability within acceptable limits. As stated in PSE&G's letter
LR-N97167, dated March 21, 1997, Hope Creek will not plan C or D EDG
outages that exceed 72 hours if the total unavailability of the EDG
will be greater than 720 hours on a 12 month rolling basis. The
proposed changes will not adversely impact the operation of any
safety related component or equipment. Since the proposed changes
involve: (1) No hardware changes; (2) no significant changes to the
operation of any systems or components; and (3) no changes to
existing structures, systems or components, there can be no impact
on the occurrence of any accident. Furthermore, there is no change
in plant testing proposed in this change request which 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 TS related to response time testing
reflect testing methodologies that were approved by the NRC in
Amendment No. 85 to the Hope Creek TS. No changes are being made to
methodologies with this proposal. Therefore, the changes contained
in this request do not result in a significant reduction in a margin
of safety.
The proposed changes to Surveillance Requirement 4.0.5 do not
alter the current requirements for the Hope Creek inservice
inspection and inservice testing programs and are considered to be
editorial in nature. Therefore, the changes contained in this
request do not result in a significant reduction in a margin of
safety.
The proposed changes to the drywell and suppression chamber
purge system are being made to reflect design modifications that
have been installed. This design modification replaced isolation
valves containing resilient material seals with metal seated valves
under 10 CFR 50.59. PSE&G has concluded that the 24 month frequency
is appropriate for the new valves since: (1) this frequency is
imposed by Surveillance Requirement 4.6.1.2.d, which is applicable
to other containment isolation valves in Table 3.6.3-1 that
penetrate the primary containment; and (2) concerns raised about
severe environment-induced degradation and frequent use for the
previously installed resilient seal material valves are not
applicable to the replacement metal seat valves. The valve
modification was an enhancement to the Hope Creek design that did
not impact the isolation capability of the drywell and suppression
chamber purge system, and does not result in a significant reduction
in a margin of safety.
The proposed changes to LCO 3.7.7 establish consistent and
appropriate requirements for main turbine bypass valve operability
requirements. These changes do not impact the assumptions contained
in these UFSAR analyses since they do not change the manner in which
Hope Creek is currently permitted to operate. Since the ACTION
Statement for LCO 3.7.7 already allows indefinite continued
operation below 25% of RATED THERMAL POWER with an inoperable main
turbine bypass valve system, the proposed modification to the
APPLICABILITY statement for this LCO would be editorial in nature.
Therefore, the changes contained in this request do not result in a
significant reduction in a margin of safety.
The HC TS Bases [Change 5--Bases for TS \3/4\.8, ``Electrical
Power Systems''] will be revised to include information contained in
the Safety Evaluation Report for Technical Specification Amendment
No. 75. This information concerns the bases for the allowed-outage-
time (AOT) for the C and D emergency diesel generators (EDGs). PSE&G
believes that implementation of 10 CFR 50.65 requirements to monitor
EDG unavailability limits will provide an acceptable and more
clearly defined method for maintaining EDG availability within
acceptable limits and 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, New Jersey 08070.
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: John F. Stolz.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: May 19, 1997
Description of amendment request: The proposed amendment would change
Technical Specification (TS) 3.7.1.3, ``Ultimate Heat Sink'' to reflect
that continued plant operation depends upon the association of ultimate
heat sink (UHS) temperature and safety system availability. The
requirements of TS 3.7.1.1, ``Safety Auxiliaries Cooling System
(SACS)'', TS 3.7.1.2, ``Station Service Water System (SSWS)'' and TS
3.8.1.1, ``Electrical Power Systems'' would be revised to reflect the
revised TS 3.7.1.3. In addition, the Bases for \3/4\.7.1, ``Service
Water Systems'' would be appropriately revised.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed TS revisions related to SSWS/SACS and the emergency
diesel generators (EDGs) [TS 3.7.1.1, TS 3.7.1.2, and TS 3.8.1.1]
involve no hardware changes and no changes to existing structures,
systems or components. The additional system configuration limits and
changes to the operation of SSWS/SACS/EDGs are being made to ensure
that SSWS/SACS can remove required heat loads during design basis
accidents and transients with the proposed UHS river water temperature
and level limits. The link to the UHS LCO in the proposed SSWS/SACS/EDG
TS ACTION Statements and the proposed revisions to the SACS ACTION
Statement for one inoperable SACS subsystem ensure that the plant is
directed to enter a safe shutdown condition whenever the capability to
[[Page 33133]]
mitigate design basis accidents and transients is lost. Since the SSWS/
SACS/EDGs will still remain capable of meeting all applicable design
basis requirements and retaining the capability to mitigate the
consequences of accidents described in the HC UFSAR, the proposed
changes were determined to be justified. As a result, these changes
will not increase the probability of an accident previously evaluated
nor significantly increase in the consequences of an accident
previously evaluated.
The proposed TS revisions related to UHS [TS 3.7.1.3] involve no
hardware changes and no changes to existing structures, systems or
components. The additional system configuration limits and changes
to the operation of UHS supported systems are being made to ensure
that the UHS can remove required heat loads during design basis
accidents and transients with the proposed UHS river water
temperature and level limits. The proposed UHS TS ACTION Statements
ensure that the plant is directed to enter a safe shutdown condition
whenever the capability to mitigate design basis accidents and
transients is lost. The proposed changes to the UHS TS surveillance
requirements to increase monitoring of the river water temperature
at 82 deg.F adequately ensures that the actions required when river
temperatures exceed 85 deg.F are taken as appropriate. Since the UHS
will still remain capable of meeting all applicable design basis
requirements and retaining the capability to mitigate the
consequences of accidents described in the HC UFSAR, the proposed
changes were determined to be justified. As a result, these changes
will not increase the probability of an accident previously
evaluated nor significantly increase in the consequences of an
accident previously evaluated.
With the approval of the proposed changes to the SSWS/SACS/EDG/
UHS TS, the proposed TS Bases changes are considered to be editorial
in nature. As a result, the proposed Bases changes will not increase
the probability of an accident previously evaluated nor
significantly increase in the consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to the SSWS/SACS/EDG 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 additional
system configuration limits and changes to the operation of SSWS/
SACS/EDGs imposed by the proposed changes ensure that SSWS/SACS and
the UHS can remove required heat loads during design basis accidents
and transients with the proposed UHS river water temperature and
level limits. Furthermore, there is no change in plant testing
proposed in this change request which 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.
The proposed changes to the UHS 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 additional
system configuration limits imposed by the proposed UHS LCO ensure
that supported systems can remove required heat loads during design
basis accidents and transients with the proposed UHS river water
temperature and level limits. Furthermore, there is no change in
plant testing proposed in this change request which could initiate
an event. The proposed changes to the UHS TS surveillance
requirements to increase monitoring of the river water temperature
at 82 deg.F adequately ensures that the actions required when river
temperatures exceed 85 deg.F are taken as appropriate. Therefore,
these changes will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
With the approval of the proposed changes to the SSWS/SACS/EDG
UHS TS, the proposed TS Bases changes are considered to be editorial
in nature. As a result, the proposed Bases 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 TS related to the SSWS/SACS/EDGs
establish consistent and appropriate requirements for SSWS/SACS/EDG
and UHS operability requirements. The additional system
configuration limits and changes to the operation of SSWS/SACS/EDG
are being made to ensure that SSWS/SACS can remove required heat
loads during design basis accidents and transients with the proposed
UHS river water temperature and level limits. The link to the UHS
LCO in the proposed SSWS/SACS/EDG TS ACTION Statements and the
revision to the SACS ACTION Statement for one inoperable SACS
subsystem ensure that the plant is directed to: (1) enter a safe
shutdown condition whenever the capability to mitigate design basis
accidents and transients is lost; or (2) enter a conservatively
short period of continued operation when system redundancy is
reduced. Since the SSWS/SACS/EDG will still remain capable of
meeting all applicable design basis requirements and retaining the
capability to mitigate the consequences of accidents described in
the HC UFSAR, the proposed changes contained in this submittal were
determined to not result in a significant reduction in a margin of
safety.
The proposed changes for the TS related to the UHS ensure
continued capability of the UHS to mitigate the consequences of
design basis accidents and transients. The additional SSWS/SACS
configuration limits and changes to the operating limits of the UHS
ensure that the UHS can remove required heat loads during design
basis accidents and transients with the proposed river water
temperature and level limits. The proposed UHS TS ACTION Statements
ensure that the plant is directed to: (1) enter a safe shutdown
condition whenever the capability to mitigate design basis accidents
and transients is lost; or (2) enter a conservatively short period
of continued operation when supported system redundancy is reduced.
Since the UHS will still remain capable of meeting all applicable
design basis requirements and retaining the capability to mitigate
the consequences of accidents described in the HC UFSAR, the
proposed changes contained were determined to not result in a
significant reduction in a margin of safety.
With the approval of the proposed changes to the SSWS/SACS/UHS
TS, the proposed TS Bases changes are considered to be editorial in
nature. As a result, the proposed bases changes will 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: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: John F. Stolz.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: May 21, 1997
Description of amendment request: The proposed amendment would
revise the Virgil C. Summer Nuclear Station Technical Specifications
(TS), Surveillance Requirements (SRs), to change the methodology for
testing the charcoal adsorbers in (1) the control room normal and
emergency air handling system (TS 3/4.7.6), and (2) the spent fuel pool
ventilation system (TS 3/4.9.11), by reference to the methodology of
ASTM D 3803-1989 from the ANSI STD N509-1980.
The proposed reference testing methodology to ASTM D 3803-1989 for
the control room is at a relative humidity of 70% and 30 degrees C with
methyl iodide penetration of < 2.5%. The proposed reference testing
methodology to ASTM D 3803-1989 for
[[Page 33134]]
the spent fuel pool is at a relative humidity of 95% and 30 degrees C
with a methyl iodide penetration of < 2.5%.
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 change revises the methodology for testing the
charcoal adsorbers in the Control Room Normal and Emergency Air
Handling System and the Spent fuel Pool Ventilation System
(Engineered Safeguards Feature [ESF] air handling units) to the
updated Standard Test Method for Nuclear-Grade Carbon.* * *. The
charcoal adsorbers are not initiators of any analyzed event.* * *
The charcoal adsorbers will be tested to the updated version of the
approved standard, which generally contains more stringent testing
requirements. The change does not affect the operation of the ESF
air handling units. The new testing requirements will continue to
ensure that the ESF air handling units will be capable of performing
their safety function and meeting the assumptions in the safety
analysis [Final Safety Analysis Report (FSAR)]. The change does not
affect the mitigation capabilities of any component or system nor
does it affect the assumptions relative to the mitigation of
accidents or transients. Therefore, the 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 revises the methodology for testing the
charcoal adsorbers in the Control Room Normal and Emergency Air
Handling System and the Spent fuel Pool Ventilation System * * * to
the updated Standard Test Method for Nuclear-Grade Carbon. The
change does not involve a significant change in the design or
operation of the plant. The changes do not involve a physical
alteration of the plant (no new or different type of equipment will
be installed), or new or unusual operator actions. No new or
different accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of this change. Therefore, the 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 margin of
safety?
The proposed change revises the methodology for testing the
charcoal adsorbers in the Control Room Normal and Emergency Air
Handling System and the Spent fuel Pool Ventilation System * * * to
the updated Standard Test Method for Nuclear-Grade Carbon. Testing
of the charcoal adsorbers in the ESF air handling units to the new
standard will continue to ensure the systems perform their design
function. The increase in the allowed penetration percentage does
not affect the accident analysis because testing requirements are
more stringent and the higher allowed percentages continue to be
below the assumptions of the safety analysis [FSAR]. Therefore, the
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Project Director: Gordon Edison, Acting.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: May 27, 1997
Description of amendments request: The proposed amendments would
revise the applicable Modes for Source Range Nuclear Instrumentation
(Technical Specification 3/4.3.1, ``Reactor Trip System
Instrumentation''), provide allowances for an exception to the
requirements for the state of the power supplies for Residual Heat
Removal System discharge to charging pump suction valves following Mode
changes (Technical Specification 3/4.5.2, ``ECCS Subsystems--
Tavg greater than 350 deg.F'' and 3/4.5.3, ``ECCS
Subsystems--Tavg less than 350 deg.F''), and delete cycle-
specific guidance concerning manual emergency engineered safety feature
function input checks.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated in
the FSAR [Final Safety Analysis Report]. The purposes for
repositioning the breakers/disconnects for MOVs [motor-operated
valves] 8706A and 8706B are to ensure that the ECCS [Emergency Core
Cooling System] System is aligned properly such that the assumptions
used in the safety analyses are met and to prevent possible
overpressurization of the charging pump suction line piping. The
likelihood of a severe transient occurring in this time frame is
very small and has to be weighed against the possibility of over
pressurizing the CVCS [Chemical and Volume Control System] charging
pump suction piping. The allowance of a 4 hour time period to
perform the required alignment appropriately weighs this risk.
Changing the applicability of the requirement to have indication
from a Source Range Nuclear Instrument available to agree with the
design of the plant does not change the physical design of the plant
or affect any assumptions used in accident analyses and, therefore,
has no effect on the probability or consequences of an accident
previously evaluated in the FSAR. The allowance of 1 hour to perform
the Source Range Channel Check upon reaching P-6 from Mode 2 is
consistent with the current basis for a source range channel
inoperable. Therefore, these changes do not involve a significant
increase in the consequences of an accident previously evaluated.
(2) The proposed changes to the Technical Specifications do not
increase the possibility of a new or different kind of accident than
any accident already evaluated in the FSAR. No new limiting single
failure or accident scenario has been created or identified due to
the proposed changes. Safety-related systems will continue to
perform as designed. Therefore, the proposed changes do not create
the possibility of a new or different kind of accident from any
previously evaluated.
(3) The proposed changes do not involve a significant reduction
in the margin of safety. The margin of safety is not significantly
reduced due to the proposed changes to the breaker/disconnect
positioning requirements of TS [Technical Specifications] 3/4.5.2
and 3/4.5.3 when transitioning between Modes 3 and 4. The likelihood
of either a severe transient occurring in Mode 3 or the possible
overpressurization of the CVCS charging pump suction line by the RHR
[residual heat removal] system in Mode 4 is very small. Changing the
Applicability of the requirement to have indication from a Source
Range Nuclear Instrument available to agree with the design of the
plant does not change the physical design of the plant or affect any
assumptions used in accident analyses and, therefore, has no effect
on the margin of safety. These proposed changes are technically
consistent with the requirements and standard format of NUREG-1431,
Revision 1. Performing the source range channel check within 1 hour
upon reaching P-6 from Mode 2 does not change the physical design of
the plant or affect any assumptions used in accident analyses and,
therefore, also does not [a]ffect the margin of safety. Thus, 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
[[Page 33135]]
proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Project Director: Herbert N. Berkow.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: May 28, 1997
Description of amendments request: The proposed amendments would
insert a footnote in Technical Specification (TS) Surveillance
Requirement 4.8.1.1.2.e, to clarify that load rejection testing of the
shared emergency diesel generator set on either unit may be used to
satisfy TS 4.8.1.1.2.e surveillance requirements for both units.
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 changes clarify that load rejection testing of the
shared emergency diesel generator set is only required once per five
years, and that testing of the shared EDG [emergency diesel
generator] set on one unit may be used to satisfy SR [Surveillance
Requirement] 4.8.1.1.2.e requirements for both units. These changes
do not affect the probability or consequences of an accident. There
are no changes being made to the emergency diesel generator testing
program. These changes simply clarify the existing test program and
the intent of the test requirements.
Therefore, the proposed TS changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes clarify that load rejection testing of the
shared emergency diesel generator set is only required once per five
years, and that testing of the shared EDG set on one unit may be
used to satisfy SR 4.8.1.1.2.e requirements for both units. No new
testing configuration is being proposed that could create the
possibility of any new or different kind of accident from any
accident previously evaluated. There are no changes being made to
the emergency diesel generator testing program. These changes simply
clarify the existing test program and the intent of the test
requirements.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes clarify that load rejection testing of the
shared emergency diesel generator set is only required once per five
years, and that testing of the shared EDG set on one unit may be
used to satisfy SR 4.8.1.1.2.e requirements for both units. A
similar technical specification change has been previously approved
by the NRC for Hatch Nuclear Plant. The technical specification
bases and the Final Safety Analysis Report have been reviewed.
Clarification of the testing requirements has no effect on the
margin of plant safety since no reduction in the test program is
involved.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Project Director: Herbert N. Berkow.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power
Plant, Unit 1, Lake County, Ohio
Date of amendment request: May 2, 1997.
Description of amendment request: The proposed change would
continue to allow entry into Operational Conditions 1, 2, and 3 with
the inboard main steam isolation valve (MSIV) leakage control subsystem
inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This License Amendment application proposes a revision to the
exception to Limiting Condition for Operation (LCO) 3.0.4 as it
applies to the Technical Specification (TS) for the MSIV Leakage
Control System (LCS). This revision is proposed to permit completion
of activities necessary to implement the most appropriate permanent
resolution for the issues that resulted from the elimination of the
secondary containment bypass leakage path through the Main Steam
Line drains. In addition, the revision clarifies that the exception
only applies to the Inboard MSIV LCS subsystem. The drains will
remain in their current configuration, which seals off the secondary
containment bypass leakage path. The sealed drain path results in a
temporary inoperability of the Inboard MSIV LCS subsystem when the
plant is operated below 50 percent rated thermal power (RTP), due to
condensate build-up in the bottom of the steam lines between the
MSIVs. The requested 3.0.4 exception is necessary to permit plant
startups with this temporary inoperability. The exception to LCO
3.0.4 simply permits use of the existing Action statement (Condition
A of LCO 3.6.1.9) during MODE changes.
The probability of occurrence of a previously evaluated accident
is not affected by the proposed revision of the LCO 3.0.4 exception
since no change to the plant or to the manner in which the plant is
operated is involved. The existing plant configuration will be
maintained, and possible concerns resulting from that configuration
have been analyzed. The extra weight of the water pooled between the
MSIVs was analyzed with respect to piping supports and seismic
considerations and was found to be acceptable, and condensate that
is carried past the outboard MSIVs will be drained to the condenser
by drain connections downstream of the outboard MSIVs before it can
reach the turbine. The temporary inoperability of the Inboard MSIV
LCS subsystem when below 50 percent RTP has no impact on accident
initiation probability, since the MSIV LCS does not serve to prevent
accidents, but is only used in mitigating the consequences of Loss
of Coolant Accidents (LOCAs) that have already occurred.
The consequences of an accident are not affected in that the
Outboard MSIV LCS subsystem will be available to perform the MSIV
LCS function by mitigating the consequences of a LOCA during the
temporary period in which the Inboard MSIV LCS subsystem is
unavailable. Condensate that is carried past the outboard MSIVs will
be drained to the condenser by drain connections downstream of the
outboard MSIVs; therefore, no impairment of the Outboard MSIV LCS
subsystem will result from condensed water. The Required Action and
Completion Time for one inoperable MSIV LCS subsystem remains the
same, and limits plant operation to the previously established 30-
day Allowable Outage Time. The Required Action if both subsystems of
MSIV LCS were to become inoperable also remains the same. The MSIV
function of isolating the Main Steam Lines is also unaffected by the
existing plant
[[Page 33136]]
configuration, since MSIV performance will not be affected by the
existence of accumulated water in the bottom of the steam lines
between the MSIVs during plant operation below 50 percent RTP.
Therefore, if necessary, the Main Steam Lines will be isolated, and
leakage past the MSIVs will be routed for filtration as in the
design-basis radiological analyses, and the safety and radiological
consequences of previously evaluated accidents will remain
unaffected.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change to permit inoperability of the Inboard MSIV
LCS subsystem during periods of startup and power ascension to 50
percent RTP and during shutdown below 50 percent RTP does not create
the possibility of a new or different kind of accident from any
previously evaluated. The Inboard MSIV LCS subsystem is only
credited during a large-break LOCA wherein Reactor Coolant System
depressurization occurs. The temporary unavailability of the Inboard
MSIV LCS subsystem can be mitigated by operation of the Outboard
MSIV LCS subsystem. The amendment to the TS is an administrative
change that does not involve change to the current plant design or
methods of operation. No new plant equipment failure modes or
accident initiators are introduced by the LCO 3.0.4 exception.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The response to a large-break LOCA will not be affected since
the Outboard MSIV LCS subsystem can be assumed to be available
during the limited period of time that the Technical Specifications
permit the Inboard subsystem to be unavailable. Allowing entry into
MODES 1, 2, and 3 while utilizing the existing Condition A and
Required Action A.1 does not reduce the margin of safety since the
Completion Time allowed for that Condition is not increased. The
proposed change will have no adverse impact on the reactor coolant
system pressure boundary nor will other system protective boundaries
or safety limits be affected.
The NRC staff has reviewed the licensees' analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power
Plant, Unit 1, Lake County, Ohio
Date of amendment request: May 2, 1997
Description of amendment request: The proposed change would allow
the leakage rate of one or more main steam lines to be up to 35
standard cubic feet per hour (scfh), as long as the total leakage rate
through all four main steam lines is less than or equal to 100 scfh.
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 involves the deletion of the portion of
Technical Specification Surveillance Requirement (SR) 3.6.1.3.10
that states the increased leakage rate of less than or equal to 35
scfh for an individual main steam line is only acceptable for
Operating Cycle 6, and a deletion of the restriction that a main
steam line leakage rate of less than or equal to 35 scfh is
acceptable for only one main steam line. The overall main steam line
leakage limit of less than or equal to 100 scfh for all four main
steam lines is not being revised.
The MSIV [main steam isolation valve] leakage is not an
initiator of an accident, including the steam line rupture accident.
Therefore, the probability of an accident previously evaluated has
not changed.
The consequences of interest are the radiological dose
consequences following a large-break Loss of Coolant Accident
(LOCA). This is the event which the regulatory guidance requires to
be evaluated using the extremely conservative source term
assumptions of Regulatory Guide 1.3, ``Assumptions Used for
Evaluating the Potential Radiological Consequences of a Loss of
Coolant Accident for Boiling Water Reactors.'' Since the overall
main steam line leakage rate of less than or equal to 100 scfh for
all four main steam lines is not being revised, the radiological
consequences of an accident previously evaluated has not changed.
Therefore, the probability or consequences of an accident
previously evaluated have not significantly increased.
2. The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This proposed change does not physically alter the plant or
systems or equipment in the plant, or the method for operation of
the plant. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The proposed change does not revise the overall combined leakage
rate of less than or equal to 100 scfh for all four main steam lines
that is permitted in the present Specification. It is the combined
main steam line penetration leakage rate that is assumed in the
radiological accident analyses. Thus, although individual steam line
leakage rates may be less than or equal to 35 scfh, as long as
overall leakage of the four main steam lines is maintained at its
current value of less than or equal to 100 scfh, the proposed change
does not reduce the margin of safety.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensees' analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: November 9, 1987, as supplemented March 31,
1988, June 8, 1992 and February 4, 1997
Description of amendment request: The proposed changes would revise
the Technical Specifications (TS) for the North Anna Power Station (NA
1&2). The changes would reformat the operability and surveillance
requirements for the intermediate range (IR) channels to be consistent
with NUREG-0452, Revision 4, ``Standard Technical Specifications (STS)
for Westinghouse Pressurized Water Reactors'' (Fall 1981), which is
applicable to NA 1&2. Also, the proposed changes would revise the
nominal IR high flux trip setpoint. The IR nuclear flux trips provide
backup reactor core protection during reactor startup. There is no
operating condition under which the IR trip provides sole overpower
protection. It is a backup trip only, and no credit is taken for the
trip in the NA 1&2 Updated Final Safety Analysis Report (UFSAR).
Operating experience at NA 1&2 has shown the IR channel response to be
sensitive to core loading patterns, varying core burnups, and control
rod positions, and the variability in the channel response had made it
difficult to maintain the channels in proper calibration. Therefore,
the proposed change would
[[Page 33137]]
elevate the nominal IR high flux trip setpoint from a current
equivalent to 25% of rated thermal power to a current equivalent to 35%
of rated thermal power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[The proposed changes would not:]
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. There is no
adverse impact on the safety analysis (since no credit is taken for
the trips in the existing analyses), and no degradation of the
protection system redundancy or reliability. This latter conclusion
is based on sensitivity studies which show that the effectiveness of
the flux trip system in protecting against the low power reactivity
excursions examined in the FSAR is not sensitive to realistic
variations in the actual flux trip setpoint.
2. Create the probability of a new or different kind of accident
from any accident previously identified, since the severity of the
analyzed accidents is unchanged, and since only a change to a
setpoint and the associated surveillance requirements for the
reactor protection system is involved.
3. Involve a significant reduction in a margin of safety, since
none of the safety analysis input or assumptions are changed, nor
are the probability nor the consequences of any previously analyzed
accidents increased.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Brenda Mozafari (Acting).
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.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: March 31, 1997
Brief description of amendment: The proposed amendment would remove
containment isolation valve 863 from Technical Specification Table 3.6-
1, ``Non-Automatic Containment Isolation Valves Open Continuously or
Intermittently for Plant Operation.''
Date of publication of individual notice in Federal Register: May
15, 1997 (62 FR 26823).
Expiration date of individual notice: June 16, 1997.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: April 25, 1997
Brief description of amendment request: The proposed amendment
changes to revise Technical Specification 3.5.2 to eliminate the flow
path from the residual heat removal system to the reactor coolant
system hot legs that is specified in Limiting Condition for Operation
3.5.2.c.2.
Date of publication of individual notice in Federal Register: May
14, 1997 (62 FR 26574).
Expiration date of individual notice: June 13, 1997.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Public Service Electric & Gas Company, Docket No. 50-311, Salem Nuclear
Generating Station, Unit No. 2, Salem County, New Jersey
Date of amendment request: May 1, 1997
Brief description of amendment request: The proposed amendment
would revise Technical Specification (TS) 3/4.7.7, ``Auxiliary Building
Exhaust Air Filtration System,'' and add a new TS Section 3/4.7.11,
``Switchgear and Penetration Area Ventilation System.'' The change to
TS 3/4.7.7 would allow for an increase in the allowed outage time from
7 to 14 days when one auxiliary building exhaust fan is inoperable. The
new TS 3/4.7.11 addresses the support function this system provides to
other necessary safety support components.
Date of publication of individual notice in Federal Register: May
15, 1997 (62 FR 26826).
Expiration date of individual notice: June 16, 1997.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: May 14, 1997
Brief description of amendment request: Your application proposes
changes to revise Technical Specification Surveillance Requirement
4.7.6.1.d.1 to indicate that the specified acceptable filter
differential pressure (DP) is to be measured across the filter housing
and to change the filter DP acceptance value from less than or equal to
3.5 inches water gauge to less than or equal to 2.70 inches water
gauge.
Date of publication of individual notice in Federal Register: May
29, 1997 (62 FR 29158).
Expiration date of individual notice: June 30, 1997.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
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
[[Page 33138]]
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.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: November 26, 1997
Brief description of amendment: The amendment revises Technical
Specifications Definition 1.M, ``Primary Containment Integrity,'' Note
6 on Table 3.2.A for the high flow main steam line instrumentation,
Table 3.2.D for a typographical error, Table 3.2.F to reflect a change
made in instrument type for the suppression chamber water temperature
instrumentation, Table 3.2.F to reflect modifications made to
suppression chamber bulk and local temperature instrumentation, Bases
Section 3/4.6G to remove an obsolete reference to Group I welds, and
Bases Section 3/4.7.A to remove ``high radiation'' from the description
of Primary Containment Group 1 initiation signals. In addition, this
amendment includes changes made to the Bases Section 3.10, ``Core
Alterations,'' as noted by BECo letter dated March 7, 1997.
Date of issuance: May 28, 1997.
Effective date: May 28, 1997.
Amendment No.: 172.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 12, 1997 (62
FR 6568). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 28, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: June 10, 1996, as supplemented by
letter dated February 17, 1997
Brief description of amendments: The amendments change the
Technical Specifications to reflect the transition from General
Electric Company (GE) to Siemens Power Corporation (SPC) as the fuel
supplier for the Quad Cities Nuclear Power Station, Units 1 and 2. In
addition, as an administrative action by the Commission that only
involves the format of the licenses and does not authorize any
activities outside the scope of the application and supplement, the NRC
has amended the licenses to include an Appendix C that lists additional
license conditions. The additional license condition as a result of the
review of this application reflects the relocation of the contents of
TS 5.4 to the Updated Final Safety Analysis Report.
Date of issuance: May 23, 1997.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 177 and 175.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Licenses, Technical Specifications and Updated Final Safety
Analaysis Report.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44355). The February 17, 1997, 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 May 23, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: August 29, 1995, as supplemented
August 7, 1996, and January 10, 1997
Brief description of amendment: The amendment revises Technical
Specifications to incorporate the commitments made in connection with
Amendment No. 183, which allowed the installation of laser welded
sleeves inside of defective steam generator tubes.
Date of issuance: May 20, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 192.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56365) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 20, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Dates of application for amendment: December 24, 1996, and January 31,
1997
Brief description of amendment: Changes Administrative Controls
Section of the Technical Specifications to implement revised management
responsibilities and titles that reflect the permanently shut down
status of the plant.
Date of issuance: May 22, 1997.
Effective date: Effective May 22, 1997, to be implemented within 60
days of issuance.
Amendment No.: 191.
Operating License No. DPR-61: Amendment revised the Technical
Specifications.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14460) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 22, 1997.
No significant hazards consideration comments received: No.
Local Public Document room location: Russell Library, 123 Broad
Street, Middletown, CT 06457.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of application for amendments: March 10, 1997
Brief description of amendments: These amendments modify Unit No. 1
Technical Specification (TS) 5.2.1 to add ZIRLO as fuel assembly
material
[[Page 33139]]
and add reference to the Nuclear Regulatory Commission approved Topical
Report WCAP-12610, ``Vantage+ Fuel Assembly Reference Core Report,'' to
TS 6.9.1.12 for both units.
Date of issuance: May 23, 1997.
Effective date: Both units, as of date of issuance, to be
implemented within 60 days.
Amendment Nos.: 203 and 84.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17231) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 23, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 5, 1997, as supplemented by letter
dated March 26, 1997
Brief description of amendment: The amendment changes the Appendix
A Technical Specifications for Waterford Steam Electric Station, Unit
3, by revising Technical Specifications 3.1.2.7, 3.1.2.8, 3.5.1, 3.5.4,
3.9.1, and Bases 3/4.1.2. The changes will increase the minimum boron
concentration in the Safety Injection Tanks and the Refueling Water
Storage Pool from 1720 to 2050 ppm.
Date of issuance: May 29, 1997, to be implemented within 60 days.
Effective date: May 29, 1997.
Amendment No.: 129.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 26, 1997, (62 FR
14461) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 29, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: June 28, 1996, as supplemented March
11, 1997
Brief description of amendment: The amendment revises Three Mile
Island, Unit 1, Technical Specifications to permit the use of 10 CFR
50, Appendix J, Option B, Performance-Based Containment Leakage
Testing.
Date of issuance: May 27, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 201.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40019) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 27, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY), Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: August 8, 1996
Brief description of amendments: The amendments allowed the
transition from Mode 4 to Mode 3 with the turbine-driven auxiliary
feedwater pump inoperable and allowed a 72-hour period after the entry
into Mode 3 to complete all necessary operability testing.
Date of issuance: May 27, 1997.
Effective date: May 27, 1997, to be implemented within 30 days.
Amendment Nos.: Unit 1--Amendment No. 87; Unit 2--Amendment No. 74.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44359) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 27, 1997.
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.
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear
Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: March 6, 1997
Brief description of amendment: The amendment revises the Technical
Specifications on allowed outage times for certain protective
instrumentation and also for reactor building access control. The
amendment adopts, in part, guidance from NUREG-0123, ``Standard
Technical Specifications for General Electric Boiling Water Reactors
(BWR/5),'' Revision 3, and NUREG-1433, ``Standard Technical
Specifications General Electric Plants BWR/4,'' Revision 1.
Date of issuance: May 28, 1997.
Effective date: As of the date of issuance, to be implemented
within 90 days.
Amendment No.: 101.
Facility Operating License No. DPR-21: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14462) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 28, 1997.
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 06360 and at the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.
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 31, 1997
Brief description of amendment: The amendment modifies Technical
Specification Surveillance 4.7.1.2.1.b, which requires the testing of
the auxiliary feedwater motor-driven and turbine-driven pumps on
recirculation flow at least once per 92 days. The amendment also makes
changes to the appropriate Bases section.
Date of issuance: May 29, 1997.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 139.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 23, 1997 (62 FR
19832) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 29, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
[[Page 33140]]
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.
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 31, 1997
Brief description of amendment: The amendment separates the
required testing of motor-operated valve thermal overload protection
into two new surveillances.
Date of issuance: May 29, 1997.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 140.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 23, 1997 (62 FR
19833) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 29, 1997.
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 06360, and the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: January 16, 1997, as supplemented on
February 24, 1997
Brief description of amendment: This amendment revises the license
to delete the prohibition on moving a spent fuel assembly shipping cask
into the Fuel Building.
Date of issuance: May 19, 1997.
Effective date: This license amendment is effective as of the date
of issuance (May 19, 1997), but shall be implemented within 30 days of
issuance.
Amendment No.: 196.
Facility Operating License No. NPF-1: The amendment revised the
license.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14467).
No significant hazards consideration comments received: No.
Local Public Document Room location: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of application for amendment: January 28, 1997
Brief description of amendment: This amendment changes the
Permanently Defueled Technical Specifications to delete the requirement
for NRC prior approval to changes in the Certified Fuel Handler's
Training Program.
Date of issuance: May 23, 1997.
Effective date: May 23, 1997.
Amendment No.: 197.
Possession-Only License No. NPF-1: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17241).
No significant hazards consideration comments received: No.
Local Public Document Room location: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
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: April 15, 1997
Brief description of amendments: These amendments revise
Surviellance Requirement 3.8.1.8 of Technical Specifications (TS)
3.8.1, ``AC Sources--Operating,'' for San Onofre Nuclear Generating
Station (SONGS), Units 2 and 3. The TS change will allow the licensee
to credit overlap testing to validate the capability of the alternate
offsite power source.
Date of issuance: June 2, 1997.
Effective date: June 2, 1997.
Amendment Nos.: Unit 2--136; Unit 3--128.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 1, 1997 (62 FR
23811) The Commission's related evaluation of the amendments is
contained in a Safety E valuation dated June 2, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: January 10, 1997, as supplemented
May 2 and May 15, 1997
Brief description of amendment: The amendment modifies the Watts
Bar Nuclear Plant (WBN) Unit 1 Technical Specifications (TS) in order
to implement 10 CFR Part 50, Appendix J, Option B, by referring to
Regulatory Guide 1.163, ``Performance-Based Containment Leakage-Test
Program.''
Date of issuance: May 27, 1997.
Effective date: May 27, 1997.
Amendment No.: 5.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4356) The May 2 and May 15, 1997 letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 27, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 18, 1997
Brief description of amendment: This amendment revises Technical
Specification Surveillance Requirement 4.5.2.c to clarify when a
containment entry visual inspection is required. This change reduces
the visual inspection requirement to at least once daily and is in
accordance with the guidance provided in Generic Letter 93-05, ``Line-
Item Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation.''
Date of issuance: May 28, 1997.
Effective date: May 28, 1997, to be implemented within 30 days of
the date of issuance.
Amendment No.: 105.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 23, 1997 (62 FR
19839) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 28, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
[[Page 33141]]
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Notice of Issuance of Amendments To Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By July 18, 1997, 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
[[Page 33142]]
sources and documents of which the petitioner is aware and on which the
petitioner intends to rely to establish those facts or expert opinion.
Petitioner must provide sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner who fails
to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. 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 the
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
Commonwealth Edison Company, Docket No. STN 50-456, Braidwood Station,
Unit No. 1, Will County, Illinois
Date of application for amendment: Two submittals dated May 23, 1997
Brief description of amendment: The amendment revises Technical
Specification (TS) 4.5.2.b.1 to include the use of ultrasonic testing
(UT) to verify that the emergency core cooling system (ECCS) is
completely filled with water. For the ECCS subsystems with high point
vent valves in direct communication with the operating systems, UT is
acceptable in lieu of physically opening the vents.
Date of Issuance: May 23, 1997.
Effective date: Immediately, to be implemented within 30 days.
Amendment No.: 83.
Facility Operating License No. NPF-72: The amendment revised the
TSs.
Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated May
23, 1997.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
Local Public Document Room location: Wilmington Public Library, 201
S. Kankakee Street, Wilmington, Illinois 60481.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket No. STN 50-454, Byron Station, Unit
No. 1, Ogle County, Illinois
Date of application for amendment: May 24, 1997, as supplemented on May
31, 1997
Brief description of amendment: The amendment revises Technical
Specification 4.5.2.b.1 to include the use of ultrasonic testing (UT)
to verify that the emergency core cooling system (ECCS) is completely
filled with water. For the ECCS subsystems with high point vent valves
in direct communication with the operating systems, UT is acceptable in
lieu of physically opening the vents. This amendment supersedes NOED
No. 97-6-010 for Byron, Unit 1, which was granted on May 23, 1997.
Date of Issuance: June 1, 1997.
Effective date: Immediately, to be implemented within 30 days.
Amendment No.: 90.
Facility Operating License No. NPF-37: The amendment revised the
TS. Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated June
1, 1997.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
Local Public Document Room location: Byron Public Library District,
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
NRC Project Director: Robert A. Capra.
Dated at Rockville, Maryland, this 11th day of June, 1997.
For The Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear
Reactor Regulation.
[FR Doc. 97-15827 Filed 6-17-97; 8:45 am]
BILLING CODE 7590-01-P