[Federal Register Volume 60, Number 60 (Wednesday, March 29, 1995)]
[Notices]
[Pages 16181-16196]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-20329]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating LicensesInvolving
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 March 3, 1995, through March 17, 1995. The
last biweekly notice was published on March 15, 1995.
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
[[Page 16182]] of the facility in accordance with the proposed
amendment would not (1) involve a significant increase in the
probability or consequences of an accident previously evaluated; or (2)
create the possibility of a new or different kind of accident from any
accident previously evaluated; or (3) involve a significant reduction
in a margin of safety. The basis for this proposed determination for
each amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By April 28, 1995, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
[[Page 16183]] telephone number, date petition was mailed, plant name,
and publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: February 24, 1995
Description of amendment request: The proposed change would remove
Section 4.3 from the Technical Specifications (TS) because the primary
system testing following opening is already performed in accordance
with the American Society of Mechanical Engineers Boiler and Pressure
Vessel Code, as implemented in the licensee's inservice inspection
program as required by TS 4.0.1.
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 does not involve a significant hazards consideration
for the following reasons.
1. The requested change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. This requested change will provide consistency between
our Technical Specifications (TS) and 10 CFR 50.55a which requires
testing in accordance with Section XI of the ASME Boiler and
Pressure Vessel Code. The requirements contained in TS Section 4.3
were placed into TS prior to incorporation of Section XI into the
ASME Boiler and Pressure Vessel Code. The NRC and industry have
since recognized the ASME Boiler and Pressure Vessel Code, Section
XI as the appropriate testing program. Adequate assurance of primary
system integrity will be provided since primary system testing will
continue to be controlled and performed in accordance with the rules
for inservice inspections provided by ASME Boiler and Pressure
Vessel Code, Section XI as implemented by our approved In-Service
Inspection (ISI) Program, as required by TS Section 4.0.1.
Therefore, there would be no increase in the probability or
consequences of an accident previously evaluated.
2. The requested change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The requested change deletes the current TS requirements
for primary system testing by recognizing that we will continue to
perform required testing consistent with 10 CFR 50.55a and ASME
Boiler and Pressure Vessel Code, Section XI, as implemented by our
approved ISI Program, as required by TS Section 4.0.1. This
requested change does not involve the addition or modification of
plant equipment, nor does it alter the design or operation of plant
systems. Therefore, the requested change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The requested change does not involve a significant reduction
in a margin of safety. The requested change deletes the current TS
Section 4.3 requirements for primary system testing and maintains
the margin of safety by continuing to perform required testing in
accordance with 10 CFR 50.55a and ASME Boiler and Pressure Vessel
Code, Section XI, as implemented by our approved ISI Program, as
required by TS Section 4.0.1. Therefore, the requested change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman
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: March 3, 1995
Description of amendment request: The proposed amendment would
eliminate the requirement to perform periodic measurement testing of
the response times for selected pressure and differential pressure
sensors. The requirement that reactor trip and engineered safety
feature response time functions be within their specified limit at
least once per 18 months will be verified instead of demonstrated. The
associated bases section for response time requirements will be changed
to allow the sensor response time portion of the channel response time
to use historical records, testing results, or vendor supplied
engineering specifications. No other changes to response time methods
are included in 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. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment does not result in a condition where the
design, material, or construction standards that were
applicable prior to the change are altered nor does it modify
any system interface. The same Reactor Trip System and Engineered
Safety Features Actuation System instrumentation is being used; the
time response allocations/modeling assumptions in the Final Safety
Analysis Report (FSAR) Chapter 15 analyses are still the same; only
the method of verifying time response is changed. The proposed
activity will not change, degrade, or prevent actions or alter any
assumptions previously made in evaluating the radiological
consequences of an accident described in the FSAR. Therefore, there
would be no increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment does not alter the performance of the
pressure and the differential pressure transmitters used in the
plant protection systems. The sensors will still have response time
verified by test before placing the sensor in operational service
and after any maintenance that could affect response time. Changing
the method of periodically verifying instrument response for certain
sensors (assuring equipemt operable) from time response testing to
calibration and channel checks will not create any new accident
initiators or scenarios. Periodic surveillance of these instruments
will detect significant degradation in the sensor response
characteristic. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated. [[Page 16184]]
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed amendment to [sic] does not affect the total system
response time assumed in the safety analysis. The periodic system
response time verification method for selected pressure and
differential pressure sensors is modified to allow use of actual
test data or engineering data. The method of verification still
provides assurance that the total system response is within that
defined in the safety analysis, since calibration tests will detect
any degradation which might significantly affect sensor response
time. Therefore, 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of amendment request: May 20, 1994, as supplemented February
2, 1995
Description of amendment request: The proposed amendment would
permit the licensee to use an alternate repair criteria (ARC),
designated as the F* criteria. Use of the F* criteria would
allow tubes with otherwise pluggable indications, to remain in service
as long as the indications are below the designated minimum distance of
the F* criteria. The F* criteria for Byron and Braidwood
defines a length of 1.7 inches of undegraded expanded tube within the
tubesheet as the minimum distance acceptable for implementing the ARC.
Below the F* length, a circumferential tube defect can exist and
the tube can remain in service. The proposed amendment will change the
plugging limit definition and would exclude plugging steam generator
tubes with indications that satisfy the F* criteria. The F*
criteria maintains the structural integrity of the degraded tube as the
primary pressure boundary and allows the tube to remain in service for
heat transfer and core cooling.
This alternate repair criteria qualification is documented in
Babcock & Wilcox Nuclear Technologies (BWNT) Topical Report BAW-10196 P
Revision 1, ``W-D4 F* Qualification Report,'' which is included as
part of the licensee's submittal. The staff's proposed no significant
hazards consideration determination for the requested change was
published on July 6, 1994 (59 FR 34659). In response to the staff's
request for additional information by letter dated February 2, 1995,
the licensee revised their previous submittal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The supporting qualification report for subject criteria
demonstrates that the presence of the tubesheet will enhance the
tube integrity in the region of the tube-to-tubesheet roll
expansions by precluding tube deformation beyond its initial
expanded outside diameter. The resistance to a tube rupture is
strengthened by the presence of the tubesheet in that region. The
results of hardrolling of the tube into tubesheet provides a
mechanical leak limiting seal between the tube and the tubesheet. A
tube rupture cannot occur because the contact between the tube and
the tubesheet does not permit sufficient movement of tube material.
The type of degradation for which the F* criteria has been
developed (cracking with a circumferential orientation) can
theoretically lead to a postulated tube rupture event provided that
the postulated through-wall circumferential crack exists near the
top of the tubesheet. An evaluation including analysis and testing
has been done to determine the resistive strength of the expanded
tubes within the tubesheet. This evaluation provides the basis for
the acceptance criteria for tube degradation subject to the F*
criteria. The F* length of roll expansion is sufficient to
preclude tube pullout from tube degradation located below the
F* distance, regardless of the extent of the tube degradation.
The Technical Specification leakage rate requirements and accident
analysis assumptions remain unchanged in the unlikely event that
significant leakage from this region does occur. For consistency
with current offsite dose limits, the site allowable leakage limit
during a MSLB has been conservatively calculated to be 12.8 gpm for
Byron and 9.1 gpm for Braidwood, which includes the accident leakage
from IPC in addition to the accident leakage from F* on the
faulted steam generator and the operational leakage limit. The
operational leakage limit of Specification 3.4.6.2.c in each of the
three remaining intact steam generators shall include the
operational leakage from F*. As a requirement for operation
following application IPC, the projected distribution of crack
indications over the operating period must be verified to result in
primary to secondary accident leakage less than the site allowable
leakage limit. Thus, the consequences of a MSLB remain unchanged.
The tube rupture and pullout is fully bounded by the existing
steam generator tube rupture analysis included in the UFSAR. The
leakage testing of the roll expanded tubes indicates that for tube
expansion lengths approximately equal to the * distance, any
postulated primary to secondary leakage from * tubes would be
insignificant. The proposed alternate repair criteria does not
adversely impact any other previously evaluated design basis
accident.
The leakage from an F* tube would be limited by the tube-
to-tubesheet interface since this leak would occur below the
secondary face of the tubesheet. Qualification testing and previous
experience indicate that normal and faulted leakage is well below
Technical Specification and administrative limits creating no
increase in the consequences associated with tube rupture type
leakages. The UFSAR analyzed accident scenarios are still bounding
since the normal and faulted leak rates are well within the normal
operating limit of 150 gallons per day. This conclusion is
consistent with previous F* programs approved and used at other
operating plants.
All of the design and operating characteristics of the steam
generator and connected systems are preserved since the F*
criteria utilizes the ``as rolled'' tube configuration that exists
as part of the original steam generator design. The F* joint
has been analyzed and tested for design, operating, and faulted
condition loadings in accordance with Regulatory Guide 1.121 safety
factors. The potential for a tube rupture is not increased from the
original submittal as demonstrated in the qualification analyses and
testing completed in the BWNT report.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
B. The proposed changes do not create the possibility of a new
or different type of accident from any accident previously
evaluated.
Implementation of the proposed F* criteria does not
introduce any changes to the plant design basis. Use of the criteria
does not provide a mechanism to initiate an accident outside of the
region of the expanded portion of the tube. In the unlikely event
the failed tube severed completely at a point below the F*
region, the remaining F* joint would retain engagement in the
tubesheet due to its length of expanded contact within the tubesheet
bore. This engagement length would prevent any interaction of the
severed tube with neighboring tubes. Any hypothetical accident as a
result of any tube degradation in the expanded region of the tube
would be bounded by the existing tube rupture accident analysis.
Tube bundle structural integrity will be maintained. Tube bundle
[[Page 16185]] leak tightness will be maintained such that any
postulated accident leakage from F* tubes will be negligible
with regard to offsite doses.
Therefore, there is not a potential for creating the possibility
of a new or different type of accident from any accident previously
evaluated.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
The use of the F* criteria has been demonstrated to
maintain the integrity of the tube bundle commensurate with the
requirements of Regulatory Guide 1.121 and the primary to secondary
pressure boundary under normal and postulated accident conditions.
Acceptable tube degradation for the * criteria is any
degradation indication in the tubesheet region, more than the
F* distance from the secondary face of the tubesheet or the top
of the last hardroll contact point whichever is further into the
tubesheet. The safety factors used in the verification of the
strength of the degraded tube are consistent with the safety factors
in the ASME Boiler and Pressure Vessel Code and Regulatory Guide
1.121 used in steam generator design. The * distance has been
verified by various testing to be greater than the length of the
roll expanded tube-to-tubesheet interface required to preclude both
tube pullout and significant leakage during normal and postulated
accident conditions. The protective boundaries of the steam
generator continue to be maintained with the use of the F*
criteria. A tube with the indication of degradation previously
requiring removal from service can be kept in service through the
F* criteria. Since the joint is contained within the tubesheet
bore, there is no additional risk associated with the previously
analyzed tube rupture event. The leak testing acceptance criteria
are based on the primary to secondary leakage limit in the Technical
Specifications and the leakage assumptions used in the UFSAR
accident analyses.
Implementation of the alternate repair criteria will decrease
the number of tubes which must be taken out of service with tube
plugs or repaired by sleeves. Both plugs and sleeves reduce the RCS
flow margin; thus, implementation of the F* criteria will
maintain the margin of flow that would otherwise be reduced in the
event of increased plugging or sleeving.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: November 22, 1994, as supplemented
January 30, March 2, and March 13, 1995.
Description of amendment request: This request was previously
published in the Federal Register on February 15, 1995 (60 FR 8746). It
is being renoticed to provide clarification to the scope of the
original request. The amendments would revise Technical Specification
(TS) 3.8 to establish restricted loading patterns and associated burnup
criteria for placing fuel in the Oconee spent fuel pools. In addition,
the Design Features sections associated with the reactor and fuel
storage would be revised. These changes are necessary to address two
new fuel designs which have increased initial fuel enrichment and
therefore cannot be stored in the spent fuel pools under existing TS or
loaded into the reactor. An administrative change would be made to TS
6.9.1 to include spent fuel pool boron concentration in the Core
Operating Limits Report. Other administrative changes would be made in
the Design Features section to make the specification consistent with
wording in the standard TS. Finally, the two additional supplements to
the original request are referenced herein.
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:
Standard 1. The proposed amendments will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Each accident analysis addressed in the Oconee Final Safety
Analysis Report (FSAR) has been examined with respect to changes in
Cycle 15 parameters to determine the effect of the Cycle 16 reload
and to ensure that the acceptance criteria of the FSAR safety
analyses remain satisfied. The transient evaluation of Cycle 16 is
considered to be bounded by previously accepted analyses. Section 7
of the Reload Report addresses ``Accident and Transient Analysis''
for this core reload.
There is no increase in the probability or consequences of an
accident due to the spent fuel storage restrictions proposed in this
amendment request. It has been shown that the calculated, worst case
keff for this area is [less than or equal to] 0.95 under all
conditions. There is no increase in the probability of a fuel drop
accident in the SFP [spent fuel pool] since the mass of the new
assemblies is not significantly different from the mass of the old
assemblies. The likelihood of other accidents, previously evaluated
and described in the FSAR, is also not affected by the proposed
changes. In fact, it could be postulated that since the increase in
fuel enrichment will allow for extended fuel cycle lengths, there
will be a decrease in fuel movement and the probability of an
accident may actually be reduced. There is also no increase in the
consequences of a fuel rod drop accident in the SFP since the
fission product inventory of individual fuel assemblies will not
change significantly as a result of increasing the initial
enrichment. In addition, no change to safety related systems is
being made. Therefore, the consequences of a fuel rupture accident
remain unchanged. In addition, it has been shown that Keff all
conditions. Therefore, the consequences of a criticality accident in
the SFP remain unchanged as well. The above analysis ensures that
the proposed reload amendment request will 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 analyses performed in support of this reload are in
accordance with the NRC approved methods delineated in Specification
6.9.2. The predicted operating characteristics of Oconee 3 Cycle 16
are similar to previously licensed designs. The Mark B10T and Mark
B11 fuel assembly designs remain mechanically compatible with all
fuel handling equipment. Therefore, no new or different kind of fuel
handling accident is created by the proposed amendment request.
Section 15.11 of the Oconee FSAR states that the refueling boron
concentration is maintained such that a criticality accident during
refueling is not considered credible. The proposed amendment request
continues to assure that a criticality accident in the SFP or during
refueling is not credible. The double contingency principle
discussed in ANSI N-16.1-1975 and the April 1978 NRC letter allows
credit for soluble boron under other abnormal or accident
conditions, since only a single accident need be considered at one
time. Thus, by requiring a minimum boron concentration in the SFP, a
criticality accident caused by violating the SFP storage
restrictions is not considered credible. Therefore, the proposed
amendment request does 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 the margin of safety.
The Oconee 3 Cycle 16 design was performed using the NRC
approved methods given in Specification 6.9.2. The safety limits for
Oconee 3 Cycle 16 are unchanged from previous cycles. The limits and
margins summarized in the Oconee 3 Cycle 16 Reload Report are well
within the allowable limits and requirements, and reflect no
reductions to any margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
[[Page 16186]] satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 14, 1993, as supplemented by
letter dated March 3, 1995.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) by removing the reactor
vessel material specimen withdrawal schedule and by updating the
reactor coolant system pressure-temperature (P-T) curves. The specimen
withdrawal schedule will be relocated to the Updated Final Safety
Analysis Report (UFSAR). The original Notice was published on January
19, 1994 (59 FR 2867).
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:
Although the Reactor Vessel material specimens withdrawal
schedule will be removed from the Technical Specifications, the
Technical Specifications bases will continue to provide background
information on the use of the data obtained from material specimens.
Also, updates to the schedule will continue to be submitted to the
NRC for approval prior to implementation.
Operating the plant in accordance with the new, updated P-T
Curves will assure preserving the structural integrity of the
reactor vessel over the life of the plant. The pressure and
temperature limits were developed in accordance with 10 CFR [Part]
50 Appendix G requirements.
Removing the requirements associated with the previous exemption
to Appendix H (TS 4.4.8.1.2 items a & b) is purely an administrative
change.
Therefore, the proposed changes will not significantly increase
the probability or consequences of any accident previously
evaluated.
Removal of the Reactor Vessel material specimen schedule from
the Technical Specifications has no impact on accidents at the
plant. Updates to the schedule will still be required to be
submitted to the NRC prior to implementation per Section II.B.3 of
Appendix H to 10 CFR Part 50.
Also, updates to the P-T Curves will not create a new or
different type [of] accident. The reactor vessel beltline P-T limits
were revised applying the general guidance of the ASME Code,
Appendix G procedures with the necessary margins of safety for
heatup, cooldown and inservice hydro test conditions.
The change to TS 4.4.8.1.2 items a & b is purely administrative.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Removal of the schedule for Reactor Vessel material specimen
withdrawal from the Technical Specifications does not impact the
margin of safety. The schedule will continue to receive NRC review
and approval prior to implementation of updates to the schedule.
Updates to the P-T Curves are provided to preserve the margin to
[sic] safety to assure that when stressed under operating,
maintenance and testing the boundary behaves in a non-brittle manner
and the probability of rapidly propagating fracture is minimized.
The change to TS 4.4.8.1.2 items a & b is purely administrative.
Therefore, the proposed changes will not result in a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 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
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: February 22, 1995
Description of amendment request: The proposed changes are
administrative in nature in that reference to an ``automatic''
containment air lock tester will be deleted from TS 4.6.1.3. The
automatic airlock tester is no longer being used.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment is administrative in nature in that the
revision will eliminate the wording associated with optional use of
the personnel airlock automatic leakage tester. The requirement for
testing the personnel airlock at a pressure greater than or equal to
Pa for at least 15 minutes remains unchanged. The acceptance
criteria of personnel airlock seal leakage less than 0.01 La is
also unchanged. The automatic leakage tester is not an accident
initiator nor a part of the success path(s) which function to
mitigate accidents evaluated in the plant safety analyses. The
proposal does not involve any changes to the configuration or method
of operation of any plant equipment that is used to mitigate the
consequences of an accident, nor does it alter any assumptions or
conditions in the plant safety analyses. Therefore, operation of the
facility in accordance with the proposed amendment would not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
(2)Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment to remove the reference to the personnel
airlock automatic tester from the technical specifications will not
introduce any new failure modes or system interactions, nor will it
require the installation of any new or modified equipment. The
requirement to leak test the personnel air locks will not be
changed. Thus, operation of the facility in accordance with the
proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3)Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment is administrative in nature in that it
eliminates the reference to the personnel airlock automatic leakage
tester but does not alter the surveillance and acceptance criteria
for such testing. Seal leakage testing is performed in accordance
with an approved plant procedure which allows use of either an
automatic tester or a portable testing cart. The automatic leakage
tester is not used to actuate safety related equipment, provide
interlocks, or perform plant control functions. The conditions
evaluated in the plant accident and transient analyses do not
involve this tester. The proposed change does not alter the basis
for any technical specification that is related to the establishment
of, or the maintenance of, a nuclear safety margin. Therefore,
operation of the facility in accordance with the proposed amendment
does not involve a significant reduction in a margin of safety.
Based on the above discussion and the supporting Evaluation of
Technical Specification changes, FPL has determined that the
proposed license amendment involves no significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this [[Page 16187]] review, it appears that the three standards of
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, et al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: February 27, 1995
Description of amendment request: The proposed amendment will
change Table 3.3-3 and 3.3-4 to accommodate an improved coincidence
logic and relay replacement for the 4.16 kV Loss of Voltage Relays.
Actions required for certain trip units with the number of operable
channels one less than the total number of channels will also be
changed. In addition, the format used to state the time delay for the
4.16 kV Degraded Voltage trip unit will be 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)Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change will result in a better overall posture of
the plant under degraded/loss of voltage conditions. The design
upgrade for the 4.16 kV Loss of Voltage system is more reliable, has
inherently higher accuracy, and is easier to maintain and calibrate
in the field. The coincidence logic will eliminate the spurious
plant trip potential from the existing design. Restating the maximum
time delay for the 4.16 kV Degraded Voltage (coincident with SIAS
[safety injection actuation signal]) protective relays in a ``less
than'' format will assure that the transfer of power to the on-site
sources occurs before the level of voltage becomes injurious to the
equipment under accident conditions, and will ensure that stripping
of the emergency power busses and loading of the EDG (s) [emergency
diesel generators] will occur within the time allowed by original
design criteria. The maximum allowed time delay for this function is
not being increased, and the time delay assumed in the accident
analyses for connecting the emergency bus to the diesel generator
will not be exceeded. Therefore, operation of the facility in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2)Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment does not change the operation, function
or modes of plant operation. The ability of the loss of power and
degraded grid voltage protection scheme to properly transfer from
the off-site to the on-site power sources is being maintained. The
relays in the improved design of the 4.16 kV Loss of Voltage
function are of the type presently being used in identical
applications at both St. Lucie plant units. No new hazards are
created or postulated which may cause an accident different from any
accident previously analyzed. The modifications will result in a
more sensitive protection scheme allowing continuous operation
without unnecessary challenges to the safety systems, and will
continue to provide adequate protection to all the safety equipment.
Therefore, operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
(3)Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The capability of the loss of power and degraded grid voltage
protection scheme is enhanced by the changes being proposed and is
confirmed by the existing surveillance requirements. The planned
modifications to the 4.16 kV Loss of Voltage function will result in
a more sensitive undervoltage detection system and reduce the
possibility of spurious actuation. The maximum time assumed in the
safety analyses for connecting each Emergency Bus to its dedicated
Emergency Diesel Generator is not being changed, and assurance that
separation from a degraded off-site power source will occur before
this time interval is exceeded during accident conditions will be
maintained by the proposed amendment. Accordingly, the margin of
safety is not affected. Therefore, operation of the facility in
accordance with the proposed amendment would not involve a
significant reduction in a margin of safety.
Based on the discussion presented above and on the supporting
Evaluation of Proposed TS [Technical Specifications] Changes, FPL
has concluded that this proposed license amendment involves no
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, et al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: February 27, 1995
Description of amendment request: The proposed amendment will
modify surveillance requirement (SR) 4.9.8.1 and 4.9.8.2 to allow a
reduction in the required minimum shutdown cooling flow rate under
certain conditions during operational MODE 6. In addition, the format
of the SR will be changed to clarify the intent of the stated
surveillances.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Operation of the SDCS [shutdown cooling system] is not an
accident initiator and, therefore, does not significantly increase
the probability of an accident previously evaluated.
The proposed change will allow a plant configuration needed to
perform maintenance activities on LPSI [low-pressure safety
injection]/SDCS headers by isolating one injection flow line for an
operable SDCS train during certain MODE 6 conditions. In the event
of a failure or unavailability of the alternate SDCS train, this
configuration could result in the proposed minimum flow rate.
The proposed change only modifies the minimum required flow
rate, and does not affect the probability of this event. FPL has
evaluated the proposed value of reactor coolant flow and has shown
that the bases for the existing LCO [limiting condition for
operation] will continue to be satisfied. Therefore, there are no
significant increases in the consequences of any event from the
proposed change. No other system interactions are involved related
to previously evaluated accidents, and the proposed change has no
adverse effect on any other system performance.
Therefore, operation of the facility in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed change does not affect the normal operation of the
plant. No new [[Page 16188]] systems are introduced and there is no
adverse effect on any other system configuration or performance. The
change will, however, allow isolation of one SDCS injection flow
path for maintenance activities in MODE 6 under controlled
conditions. The failure of the alternate SDCS train does not create
a new accident and has been further evaluated in the reduced flow
configuration, and shown to meet all the TS bases requirements.
Therefore, operation of the facility in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
(3)Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The safety considerations related to the proposed change are
described in the bases to TS [Technical Specification] 3/4.9.8. FPL
has evaluated the proposed reduction in SDCS flow requirement, under
stated conditions, and has shown that the proposed flow rate meets
all the TS bases requirements involving decay heat removal, boron
dilution, and stratification. Established acceptance criteria
providing margins of safety are not being changed by the proposed
amendment. Therefore, operation of the facility in accordance with
the proposed amendment would not involve a significant reduction in
a margin of safety.
Based on the discussion presented above and on the supporting
Evaluation of Proposed TS Changes, FPL has concluded that this
proposed license amendment involves no significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location:Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket No.
50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County,
Georgia
Date of amendment request: March 14, 1995
Description of amendment request: Georgia Power Company (GPC or the
licensee) has proposed a temporary change to Hatch Unit 2 Technical
Specification (TS) Required Action 3.3.6.1.F.1, and associated Bases.
The proposed change would add a note to the Primary Containment
Isolation Instrumentation actions to permit the drywell and wetwell
purge valves which are isolated by the drywell radiation monitor signal
to be opened with one inoperable drywell radiation monitor. The note
will expire prior to startup from the Hatch Unit 2 refueling/
maintenance outage scheduled in the fall of 1995, at which time the
radiation monitor can be repaired or replaced. Should the unit be
forced into a cold shutdown of sufficient duration (i.e., drywell de-
inerted), the inoperable radiation monitor will be repaired at that
time. The TS containment sections allow these valves to be opened for
inerting, de-inerting, and pressure control. However, with radiation
monitor 2D11-K621B inoperable, the primary containment isolation
instrumentation TS require the valves be closed until the unit achieves
a cold shutdown condition. Without the ability to open these valves
until cold shutdown, pressure control and de-inerting are difficult.
The purpose of the high drywell radiation primary containment
isolation signal is to limit fission product release following a
postulated loss-of-coolant accident (LOCA) with significant fuel
damage. It is one of several signals which isolate the primary
containment vent and purge valves. A high drywell pressure signal will
not only shut down the reactor and generate a LOCA signal, it will also
isolate these valves.
High drywell radiation indicates possible gross failure of the fuel
cladding. The generation of this isolation signal is not credited in
any accident or transient analysis. Chapter 15 of the Hatch Unit 2
Final Safety Analysis Report (FSAR) discusses the radiological
consequences of a postulated large break LOCA with fuel failure to show
conformance to 10 CFR Part 100 and 10 CFR Part 50, Appendix A. This
analysis is not affected.
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 change 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.
Opening the containment purge and vent valves with an inoperable
drywell radiation monitor will not increase the probability of any
previously evaluated accident. The fact that the monitor cannot send
an automatic isolation signal will not significantly affect the
consequences of an accident. The function of the primary containment
isolation signal is to detect and limit release of fission products
following significant fuel damage. The generation of this isolation
signal is not credited in any accident or transient analysis.
Chapter 15 of the Unit 2 FSAR evaluates the radiological
consequences of a postulated design basis LOCA with non-mechanistic
fuel damage. This licensing evaluation shows conformance to the
radiological limits presented in 10 CFR 100 and 10 CFR 50, Appendix
A. The results of this analysis are not affected since the valves
are otherwise operable and receive isolation signals from other
instrumentation.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve the installation of any new
equipment, or the modification of any equipment designed to prevent
or mitigate the consequences of accidents or transients. Therefore,
the change has no effect on any accident initiator, and no new or
different type of accidents are postulated to occur.
3. The proposed amendment does not result in a significant
reduction in the margin of safety.
As discussed in Item 1 above, the assumptions and results of the
licensing evaluations remain unchanged. Therefore, the margin of
safety is not significantly affected.
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:Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: February 28, 1995
Description of amendment request: Technical Specification (TS)
Section 6.5.1.12 would be revised to delete the requirement to render
determinations in writing with regard to whether or not activities
listed in TS Sections 6.5.1.2 and 6.5.1.5 constitute an unreviewed
safety question.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the [[Page 16189]] issue of no significant
hazards consideration, which is presented below:
. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed change removes the requirement to
render determinations in writing with regard to whether or not
opposed changes to the Technical Specifications and investigations
of violations of Technical Specifications constitute an unreviewed
safety question. This change is considered an administrative change
to remove a requirement which is not relevant to these activities
and which is also consistent with the BWR Revised Standard Technical
Specifications (NUREG 1433). Existing requirements to perform
Technical and Independent Safety Reviews of these activities are not
affected. Therefore, the proposed amendment does not significantly
increase the probability of occurrence or the consequences of an
accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated. The
proposed change is considered administrative since it removes a
requirement which is not relevant to the affected activities, and
which is also consistent with the BWR Revised Standard Technical
Specifications Administrative Controls for Review and Audit.
Existing requirements to perform Technical and Independent Safety
Reviews for the affected activities are not changed. Therefore, this
change has no effect on the possibility of creating a new or
different king of accident from any accident previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The proposed change removes a requirement which is not
relevant to the affected activities. Existing Technical
Specification requirements to perform Technical and Independent
Safety Reviews for the affected activities are not changed and
therefore, will continue to ensure that such activities properly
address nuclear safety and safe plant operation. Therefore, it is
concluded that operation of the facility in accordance with the
proposed amendment does not involve a 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:Ocean County Library, Reference
Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Phillip F. McKee
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: February 15, 1995
Description of amendment request: The proposed amendment would
modify (by relocation to the Technical Requirements Manual) Technical
Specification (TS) 3/4.3.3.7, Chemical Detection Systems, and TS 3/
4.8.4.1, Electrical Equipment Protective Devices - Containment
Penetration Conductor Overcurrent Protective Devices, and the
associated Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to Technical Specification 3.3.3.7, Chemical
Detection Systems and 3.8.4.1, Electrical Equipment Protective
Devices-Containment Penetration Conductor Overcurrent Protective
Devices, is of an administrative nature in that the listed Technical
Specifications and Bases will be relocated in entirety to the
Technical Requirements Manual (TRM). Any future changes to the
relocated requirements will be in accordance with 10CFR 50.59 and
approved station procedures. Whether the listed Technical
Specifications and Bases are located in Technical Specifications or
the Technical Requirements Manual has no effect on the probability
or consequences of any accident previously evaluated.
The proposed change does not alter the assumptions previously
made in the listed Technical Specifications. The proposed change
allows the Commission and South Texas more effective use of
personnel resources to control requirements that meet the four
Criteria in the Final Policy Statement. The proposed change will not
change the dose to workers.
Since the probability of a [sic] accident is unaffected by the
administrative relocation of the listed Technical Specifications,
and the doses are not affected and do not exceed acceptance limits,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change to Technical Specification 3.3.3.7, Chemical
Detection Systems and 3.8.4.1, Electrical Equipment Protective
Devices-Containment Penetration Conductor Overcurrent Protective
Devices, is of an administrative nature in that the listed Technical
Specifications and Bases will be relocated in entirety to the
Technical Requirements Manual (TRM). Any future changes to the
relocated requirements will be in accordance with 10CFR 50.59 and
approved station procedures. Whether the listed Technical
Specifications and Bases are located in Technical Specifications or
the Technical Requirements Manual has no effect on any previously
evaluated accident. It does not represent a change in the
configuration or operation of the plant and, therefore, does not
create the possibility of a new or different type of accident from
any accident previously evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
The proposed change to Technical Specification 3.3.3.7, Chemical
Detection Systems and 3.8.4.1, Electrical Equipment Protective
Devices-Containment Penetration Conductor Overcurrent Protective
Devices, is of an administrative nature in that the listed Technical
Specifications and Bases will be relocated in entirety to the
Technical Requirements Manual (TRM). Any future changes to the
relocated requirements will be in accordance with 10CFR 50.59 and
approved station procedures. The margin of safety is not reduced
when the requirements are relocated to a Licensee-controlled
document because the requirements to change a License Basis Document
via the 10CFR 50.59 process ensures the same questions concerning
the margin to safety required for a License Amendment are asked. The
major difference is the time and expense required for the License
Amendments. Therefore, this proposed change does not significantly
reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036
NRC Project Director: William D. Beckner
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: February 15, 1995 [[Page 16190]]
Description of amendment request: The proposed amendment would
modify Technical Specification 4.6.2.3.a.2 (and associated Bases) to
reflect the reactor containment fan cooler flow rate assumed in the
accident analyses and to specify that this flow is provided by the
component cooling water 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. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to Technical Specification 4.6.2.3.a.2 is to
reflect the cooling water temperature assumed in the accident
analyses. The revised Technical Specification surveillance
requirement will change the cooling water flow rate requirement to
each Reactor Containment Fan Cooler from greater than or equal to
550 gallons per minute to greater than or equal to 1800 gallons per
minute.
The proposed change, which will result in an increased
acceptance criteria for the flow to the Reactor Containment Fan
Coolers, is not indicative of accident initiators. The change will
ensure that the surveillance requirement reflects the flow rate
value assumed in the South Texas Project accident analyses and that
the design and operability requirements of equipment important to
safety are ensured.
The accident mitigation features of the plant are not affected
by the proposed change since the change reflects the original
assumptions made in the design of the accident mitigation features
of the South Texas Project. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not create the possibility of a new or
different kind of accident previously evaluated in the Safety
Analysis Report because all the accidents were analyzed with a flow
rate of 1800 gallons per minute to the Reactor Containment Fan
Cooler.
3. Does the proposed change involve a significant reduction in a
margin of safety?
There will be no adverse affects on margins of safety since a
more stringent surveillance requirement will be applied to the
Reactor Containment Fan Cooler. The Technical Specification
operability and surveillance requirements are not reduced but rather
made more restrictive by this proposed change. The change ensures
that the margin of safety originally intended for the Reactor
Containment Fan Coolers is maintained.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location:Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036
NRC Project Director: William D. Beckner
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: February 13, 1995
Description of amendment request: The proposed amendment would
delete the audit frequency requirements from the Duane Arnold Energy
Center Technical Specifications (TS) and add them to the Quality
Assurance Program Description located in the Updated Final Safety
Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed amendment does not involve a change in the
probability or consequences of an accident previously evaluated. No
physical changes will occur as a result of this amendment. The
change is administrative in nature and does not impact the operation
of the plant or the plant's response to any accident. Because it
will allow management the flexibility to adjust the audit
frequencies based upon the performance of the program or
organization being audited, the overall performance of the
organization will be improved.
(2) The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No physical changes will occur as a result of this
amendment. The change is administrative in nature and does not
affect the operation or design of the plant; therefore, it does not
create the possibility of a new or different kind of accident from
any accident previously evaluated. The audits will continue to be
performed to provide assurance of conformance to the applicable
requirements.
(3) The proposed amendment will not reduce the margin of safety.
No physical changes will occur as a result of this amendment. The
change is administrative in nature and does not affect the operation
or design of the plant. Safety limits and limiting safety system
settings are not affected by this proposed amendment. The amendment
removes requirements for frequency of audits from the TS, thus
permitting more effective scheduling of audits based on performance
and the status of the activities audited. This should result in a
more effective audit program that will contribute to an improvement
in the overall performance of the organization.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, 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, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis
& Bockius, 1800 M Street, N. W., Washington, D. C. 20036-5869NRC Acting
Project Director: John N. Hannon
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: February 10, 1995
Description of amendment request: The proposed amendment would
modify Technical Specification 3.3.2.1, ``Control Rod Block
Instrumentation,'' to revise two surveillance requirements and their
associated notes for the Rod Withdrawal Limiter (RWL) mode of the Rod
Pattern Control System. These changes will conform these requirements
to their original bases and eliminate the potential for unnecessary
power reductions.
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 are consistent with the Rod Withdrawal
Error (RWE) analysis presented in Clinton Power Station (CPS)
Updated Safety Analysis Report (USAR) Section 15.4.2. The proposed
changes do not result in any change to plant equipment or operation;
only the plant conditions for which the Rod Withdrawal Limiter (RWL)
function(s) are required to be tested are being revised. The
proposed changes continue to ensure that the RWL is OPERABLE and
tested to ensure that continuous control rod withdrawals remain
within the assumptions of the RWE analyses. The proposed changes
have no impact on the probability of occurrence of a RWE event.
Therefore, the proposed changes do not result in a significant
increase in the probability or consequences of any accident
previously evaluated.
(2) The proposed changes do not result in any changes to plant
equipment or operation; only the plant conditions for which the RWL
[[Page 16191]] function(s) are required to be OPERABLE and tested
are being revised. The proposed changes continue to ensure that the
RWL is OPERABLE and tested to ensure that continuous control rod
withdrawals remain within the assumptions of the RWE analyses. As a
result, no new failure modes are introduced. The proposed changes
are clearly within the limits of plant operation as described in the
USAR and the RWE analyses. Therefore, the proposed changes cannot
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) The proposed changes revise the testing requirements to be
consistent with the testing required prior to Amendment No. 95. The
proposed changes ensure that the RWL is OPERABLE and tested to
ensure that continuous control rod withdrawals remain within the
assumptions of the RWE analyses. The proposed changes are clearly
within the limits of plant operation as described in the USAR and
the RWE analyses. Therefore, the proposed changes do not involve a
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:Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Attorney for licensee: Leah Manning Stetzner, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th St., Decatur,
Illinois 62525.
NRC Acting Project Director: John N. Hannon
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine
YankeeAtomic Power Station, Lincoln County, Maine
Date of amendment request: February 14, 1995
Description of amendment request: The proposed amendment would
change responsibility for audits of the emergency and security plans
and their implementing procedures. Audit responsibility would change
from the licensee's Nuclear Safety Audit and Review (NSAR) Committee
and the Plant Operation Review Committee (PORC), to the respective
emergency and security plans. The proposed amendment is consistent with
the guidance of NRC Generic Letter 93-07, Modification of the Technical
Specification Administrative Control Requirements for Emergency and
Security Plans, dated December 28, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the Standards of 10 CFR 50.92(c). A summary of the licensee's
analysis is presented below:
1. The proposed amendment would not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
The proposed changes do not have a direct effect on the physical
plant or the maintenance of the physical plant, but would improve the
safe operation of the plant by reducing the administrative burden of
PORC and NSAR. This change would allow a better focus of management
resources to the operational safety oversight of plant activities. The
requirement to review, audit, document, control, and submit for
regulatory review, the Emergency Plan and the Security Plan and their
implementing procedures, is defined by regulation and remains
unchanged. The proposed changes will not, of themselves, result in any
reduction in the effectiveness of either the Emergency Plan or the
Security Plan to protect the health and safety of the public. The
proposed changes, therefore, will not increase the probability or
consequences of an accident previously evaluated.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any previously evaluated.
This change is administrative in nature and does not change or
modify the physical plant or maintenance of the physical plant.
Applicable regulations continue to enforce the requirement for review
and audit by individuals not responsible for implementation of the
existing programs. Consequently, independent oversight of the programs
and procedures is not compromised by these proposed changes. The
possibility of a new or different accident from any previously
evaluated as a result of future changes in the implementation of the
Security or Emergency Plans is not created.
3. The proposed amendment would not involve a significant
reduction in a margin of safety.
The proposed changes will revise the administrative
responsibilities of the PORC and NSAR committees allowing a better
focus of resources on operational safety reviews. The requirements
of the applicable Federal and State regulations ensure the continued
effective oversight of the implementation of Security and Emergency
Plans. Consequently, the adoption of the proposed changes would not
involve a significant reduction in a margin of safety.
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: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, Maine 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, Maine 04011
NRC Project Director: Walter R. Butler
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: October 18, 1994, as supplemented
February 21, 1995.
Description of amendment request: The following changes requested
in the October 18, 1994, submittal were published in Federal Register
on November 9, 1994 (59 FR 35876). The proposed amendment would require
three Type A overall Integrated Containment Leakage Tests be conducted
at approximately equal intervals during shutdowns during each 10 year
service period. For the third Type A test for the second 10-year
period, it would be conducted during the thirteenth refueling outage
extending the second 10-year service period to the end of the
thirteenth refueling outage. The amendment would also change the
Containment Leakage Bases by reflecting the conditions of a proposed
exemption to 10CFR50, Appendix J, that would remove the requirement
that the third Type A test for each 10-year period be conducted when
the plant is shutdown for the 10-year plant inservice inspection.
By letter dated February 21, 1995, the licensee withdrew the action
related to conducting the third Type A test for the second 10-year
period during the thirteenth refueling outage and the reference to a
proposed exemption to 10 CFR 50, Appendix J, that would remove the
requrement that the third Type A test for each 10-year period be
conducted when the plant is shutdown for the 10-year plant inservice
inspection. The following basis for the proposed no significant hazards
consideration determination relates to the February 21, 1995, request.
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 16192]] consideration, which is presented below:
The proposed change does not involve a SHC because the change
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Type A tests are performed to ensure that the total leakage from
containment does not exceed the maximum allowable primary
containment leakage rate at the design pressure. This assures
compliance with the dose limits of 10CFR100.
The proposed change to Surveillance Requirement 4.6.1.2.a of the
Millstone Unit No. 2 Technical Specifications will increase the
flexibility for scheduling the Type A tests. It does not modify the
maximum allowable leakage rate at the design containment pressure,
does not impact the design basis of the containment, and does not
make any physical or operational changes to existing plant
structures, systems, or components.
Historically, Type A tests have a relatively low failure rate
where Type B and C testing (local leakage rate tests) could not
detect the leakage path. Most Type A test failures are attributed to
failures of Type B or C components (containment penetrations and
isolation valves). Type B and C components are tested per
Surveillance Requirement 4.6.1.2.d of the Millstone Unit No. 2
Technical Specifications. These tests are required to be conducted
at intervals no greater than 24 months. These local leakage rate
tests provide assurance that containment integrity is maintained.
The Type B and C tests will continue to be performed in accordance
with the requirements of Surveillance Requirement 4.6.1.2.d.
The previous Type A, B, and C tests demonstrate that Millstone
Unit No. 2 has maintained control of containment integrity by
maintaining a conservative margin between the acceptance criterion
and the ``As-Found'' and ``As-Left'' leakage results. Based on this,
the Millstone Unit No. 2 containment is considered to be in sound
condition.
Based on the above, the proposed change to Surveillance
Requirement 4.6.1.2.a of the Millstone Unit No. 2 Technical
Specifications does not involve a significant increase in the
probability or consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change to Surveillance Requirement 4.6.1.2.a of the
Millstone Unit No. 2 Technical Specifications will increase the
flexibility in scheduling the Type A tests. It does not make any
physical or operational changes to existing plant structures,
systems, or components. In addition, the proposed change does not
modify the acceptance criteria for the Type A tests. Maintaining the
leakage through the containment boundary to the atmosphere within a
specific value ensures that the plant complies with the requirements
of 10CFR100. The containment boundary serves as an accident
mitigator; it is not an accident initiator. Therefore, the proposed
change to Surveillance Requirement 4.6.1.2.a does not create the
possibility of a new or different kind of accident from any
previously analyzed.
3. Involve a significant reduction in the margin of safety.
The proposed change to Surveillance Requirement 4.6.1.2.a of the
Millstone Unit No. 2 Technical Specifications will increase the
flexibility for scheduling the Type A tests. It does not modify the
maximum allowable leakage rate at the design containment pressure,
does not impact the design basis of the containment, and does not
make any physical or operational changes to existing plant
structures, systems, or components.
Based on the above, 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: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: February 1, 1995
Description of amendment request: This Technical Specification (TS)
change would modify the applicable operational conditions for the
secondary containment isolation radiation monitors located on the
refueling floor and for the radiation monitor located in the railroad
access shaft. Specifically, for the refueling floor exhaust duct and
wall exhaust duct radiation monitors, the proposed change would modify
the applicable operational condition during specific control rod
testing evolutions which are core alterations and would indicate that
the operability requirement does not apply during shutdown margin
demonstrations. For the railroad access shaft exhaust duct radiation
monitor, the change to the TS would modify the applicable operational
condition to address plant evolutions involving irradiated fuel
transfer within the railroad accessshaft and above the access shaft
with the equipment hatch open.
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:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
a. The proposed change to the applicable operational condition
for the refueling floor process radiation monitors does not affect
the probability of the design basis accidents. The monitors function
in response to an airborne radioactivity concentration in the
unfiltered air from the Zone III exhaust system and provide
isolation signals which limit offsite doses to within regulatory
limits. As such, there is no correlation between monitor operability
and accident probability. The monitors act to mitigate the offsite
effects of airborne contamination producing accidents, they are not
potential accident initiators.
The proposed change does not result in a significant increase in
the consequence of the design basis accidents. The postulated event
associated with control rod related CORE ALTERATIONS which could
result in increased Zone III airborne radioactivity concentrations
is criticality resulting from a single control rod withdrawal,
resulting in release of fission products. The probability of an
unintended criticality from a single control rod withdrawal is low,
and the potential for this criticality to result in fuel failure
under shutdown conditions is even more remote. Withdrawal of a
single control rod is an analyzed evolution during which time
adequate design and operating controls exist to preclude
criticality. However, in the unlikely event criticality should
occur, the potential offsite effects would not be significant.
Localized criticality involving a leaking rod, or criticality
induced fuel failure, are the postulated mechanisms by which an
increase in Zone III airborne radioactivity could be attained.
Neither of these postulated, but very unlikely events, will result
in radioactive release in excess of 10CFR100 limits. Any release
would be monitored by instrumentation in the Reactor Building vent
stack required to be OPERABLE at all times. In addition, Area
Radiation Monitors are installed on the refueling floor to
supplement the refueling floor process radiation monitors by
providing radiological information to plant operators. Operators can
use the vent stack and/or ARM information to manually initiate
secondary containment isolation if radiological conditions warrant
this action. Emergency Operating Procedures direct operator action
in the event of higher than normal radiation readings.
b. The proposed change to the applicable operational condition
for the railroad access shaft process radiation monitor does not
affect the probability of the design basis accidents. The monitor
functions in response to an airborne radioactivity concentration in
the unfiltered air from the Zone III exhaust [[Page 16193]] system
and provides isolation signals which limit offsite doses to within
regulatory limits. As such, there is no correlation between monitor
operability and accident probability. The monitor acts to mitigate
the offsite effects of airborne contamination producing accidents,
it is not a potential accident initiator.
The proposed change does not result in a significant increase in
the consequence of the design basis accidents. The design intent of
the railroad access shaft process radiation monitor is to monitor
radiation in the unfiltered air from the Zone III railroad access
shaft exhaust system, and provide signals which automatically
isolate the Zone III portion of the secondary containment, start the
Standby Gas Treatment System, and start the Recirculation System
(Zone III) on a high radiation condition within the access shaft.
This function is intended to limit the consequences of a fuel
handling accident in the railroad access shaft. The monitor has no
significant capability to react to a CORE ALTERATION related
transient, or one resulting from operations with the potential to
drain the reactor vessel. The design intent of the monitor is
maintained under the proposed change, as the proposed change focuses
monitor operability on conditions when irradiated fuel is in the
railroad access shaft or above it with the railroad access shaft
cover open.
For the above stated reasons, the applicable operational
condition for Refueling Floor Exhaust Duct High Radiation Monitors,
Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft
Exhaust Duct Radiation Monitor can be modified without significantly
increasing the probability or consequences of an accident previously
evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The Refueling Floor Exhaust Duct High Radiation Monitors, Wall
Exhaust Duct Radiation Monitors, and the Railroad Access
ShaftExhaust Duct Radiation Monitor function in response to an
airborne radioactivity concentration in the unfiltered air from the
Zone III exhaust system and provide isolation signals which limit
offsite doses to within regulatory limits. As such, there is no
correlation between monitor operability and the potential for
creating new or different accident scenarios. The monitors act to
mitigate the offsite effects of airborne contamination producing
accidents, they are not potential accident initiators.
For the above stated reasons, the applicable operational
condition for Refueling Floor Exhaust Duct High Radiation Monitors,
Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft
Exhaust Duct Radiation Monitor can be modified without creating the
possibility of a new or different kind of accident from any accident
previously evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
a. The proposed change to the applicable operational condition
for the refueling floor process radiation monitors does not involve
a significant reduction in the margin of safety. The postulated
event associated with control rod related CORE ALTERATIONS which
could result in increased Zone III airborne radioactivity
concentrations is criticality resulting from a single control rod
withdrawal under shutdown conditions. There are multiple barriers to
protect against the postulated event of criticality from a single
rod withdrawal. Technical Specifications, plant operating
procedures, and plant design control the withdrawal of control rods
to minimize the potential for an inadvertent criticality event
during shutdown. In addition, a fuel loading verification is
performed, per procedure, on the as loaded core configuration to
ensure that the fuel is loaded correctly. Each reload core is
designed such that there is at least a 99.9% probability with a 95%
confidence that the core will not be critical as a result of a
single control rod withdrawal. The safety margin associated with a
potential criticality event from a single control rod withdrawal,
under shutdown conditions, is not impacted by the proposed change.
In the unlikely event that control rod manipulations resulted in
reactor criticality, adequate protective measures are provided by
core monitoring instrumentation required to be operable in OPCON 5.
Under this scenario, assuming the inadvertent control rod withdrawal
resulted in a significant reactivity addition, the Reactor
Protection System (RPS) would respond by inserting all control rods
via the Scram function. The RPS monitors for recriticality during
OPCON 5 with SRMs (per Technical Specification Section 3.9.2), and
IRMs. The safety margin associated with RPS response to a
criticality event, under shutdown conditions, is not impacted by the
proposed change.
Assuming that a criticality did occur as a result of a single
control rod withdrawal, any increase in Zone III airborne
radioactivity from a previously failed assembly located in the
vicinity of the withdrawn control rod or a fuel rod failure
associated with the control rod withdrawal would not result in an
offsite dose exceeding regulatory limits. Assuming that criticality
occurs following core loading and verification (i.e. 20
days after shutdown), the offsite dose as a result of the release of
fission products from a single failed fuel rod would be much less
than 1% of the applicable site boundary limits. In addition, the
failure of four complete fuel assemblies (i.e. nearly equal to 300
fuel rods in the bundles surrounding the withdrawn control rod)
would not result in offsite dose exceeding the applicable regulatory
limits. Failure of more than four complete fuel assemblies due to
the withdrawal of a single control rod in OPCON 5 is not considered
credible. In fact, given the initial conditions of this event (i.e.
cold, zero power, subcritical) and the reactivity characteristics of
the fuel (i.e. negative fuel temperature reactivity coefficient) it
is very unlikely that a criticality of this nature would result in
failure of any fuel rods. Although the refueling floor process
radiation monitors would not be OPERABLE, Zone III airborne
radioactivity concentrations can be independently detected with Area
Radiation Monitors (ARMs) which are located on the refueling floor.
These monitors provide control room indication, and would alert
operators to changing radiological conditions on the refueling
floor. In addition to providing personnel notification, the ARMs act
as a supplement to the process radiation monitors in detecting
abnormal migrations of radioactive material in or from the process
streams. Operators can manually initiate secondary containment
isolation based on ARM input. The Emergency Operating Procedures
require the operators to take appropriate actions on higher than
normal radiation readings. Moreover, any airborne radioactivity
leakage from Zone III would be monitored via instrumentation in the
Reactor Building vent stack required to be OPERABLE at all times;
local alarms, remote recording, and main control room and Technical
Support Center alarms are provided. Operators can manually initiate
secondary containment isolation based on exhaust sample readings.
Due to the bounding regulatory limits and the redundant monitoring
and operator response capabilities, the safety margin associated
with the potential for offsite airborne radioactive release, under
shutdown conditions, is not significantly impacted by the proposed
change.
b. The elimination of operability requirements associated with
CORE ALTERATIONS, operations with the potential to drain the reactor
vessel, and other irradiated fuel moves not associated with the
railroad access shaft, do not affect the ability of the railroad
access shaft process radiation monitor to implement its design
function. As such, the current operability requirements for the
monitor which involve evolutions in areas other than the railroad
access shaft do not contribute to the margin of plant safety; thus
eliminating these operability requirements will not reduce the
margin of plant safety.
For the above stated reasons, the applicable operational
condition for Refueling Floor Exhaust Duct High Radiation Monitors,
Wall Exhaust Duct Radiation Monitors, and the Railroad Access Shaft
Exhaust Duct Radiation Monitor can be modified without 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz [[Page 16194]]
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: February 2, 1995
Description of amendment request: This amendment would change the
Technical Specifications for the units to increase the licensed
discharge fuel assembly exposure for SPC 9X9-2 fuel from 40 to 45 GWD/
MTU.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not:
I. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
PP&Ls technical basis for increasing the licensed discharge
exposure limit as proposed is documented in PL-NF-94-005-P-A. The
technical basis includes onsite fuel inspections, fuel design
analyses and evaluations, and an in-reactor fuel assembly extended
exposure demonstration. In response to NRC concerns on fuel failures
at higher exposures, very conservative analyses were performed for
the CRDA [control rod drive assembly] assuming very low failure
thresholds, and offsite dose calculation results were shown to be
well within regulatory limits, even at a failure threshold of 30
cal/gm. The NRC has previously reviewed and approved all of the
above information, and inspection results have met all approved
criteria.
An evaluation of FSAR [Final Safety Analysis Report] design
basis events was performed to determine the impact of the proposed
increase in fuel exposure. The LOCA [loss-of-coolant accident]
analysis performed in support of PP&Ls Power Uprate efforts
incorporated the effects of higher exposure and LHGR [linear heat
generation rate]. From a radiological release perspective, the Power
Uprate evaluations of LOCA, MSLB [main steam line break], CRDA, and
refueling accidents each bound the potential impacts of extended
exposure fuel.
Those reload analyses deemed necessary to confirm that the above
conclusions remain valid will be performed on a cycle-specific
basis.
Based on the above, the proposed action will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
II. Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed action will increase the residence time of fuel
within the Susquehanna reactors. The potential consequences of this
action remain solely with the fuels ability to perform within
specified limits during the increased duty, and were reviewed in I
above. All required evaluations involving fuel impacts have been
previously evaluated.
Based on the above, the proposed action cannot create the
possibility of a new or different kind of accident from any accident
previously evaluated.
III. Involve a significant reduction in a margin of safety.
The proposed action will allow increasing the licensed discharge
fuel assembly exposure limit, resulting in increases in the fuel rod
LHGR and LHGR for APRM [average power range monitor] Setpoints,
which are controlled via the Technical Specifications and the Core
Operating Limits Report.
The discussion in I. above delineates the evaluations performed
to support this action. It concludes that neither the probability
nor the consequences of events previously evaluated will be
affected. Operator performance will not be affected, because the
operators only monitor the ratio of the fuel LHGR to the fuel design
limit. No other potentially impacted safety margins have been
identified.
Based on the above, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location:Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: February 10, 1995
Description of amendment request: The proposed amendment would
modify the Susquehanna Steam Electric Station, Unit 1 and 2 Technical
Specifications (TS) to (1) extend the allowable out-of-service times
(AOTs) for maintenance and repair and the surveillance test intervals
(STIs) between channel functional tests for the following groups of
instruments: reactor protection systems instrumentation (TS 3.3.1),
isolation actuation instrumentation (TS 3.3.2), emergency core cooling
system actuation instrumentation (TS 3.3.3), ATWS (anticipated
transient without scram) recirculation pump trip system instrumentation
(TS 3.3.4.1), end-of-cycle recirculation pump trip system
instrumentation (TS 3.3.4.2), reactor core isolation cooling system
(RCIC) actuation instrumentation (TS 3.3.5), control rod block
instrumentation (TS 3.3.6), radiation monitoring instrumentation (TS
3.3.7.1), and feedwater/main turbine trip system actuation
instrumentation (TS 3.3.90); (2) change the required actions and AOTs
for the instruments listed above to make requirements consistent with
supporting analysis in General Electric topical reports and change
additional actions required to prevent extended AOTs from resulting in
extended loss of instrument function; (3) change the required actions
and AOTs for the instruments listed above for instrumentation
associated with the ADS (automatic depressurization system),
recirculation pump trip, and pump suction lineup for HPCI (high
pressure core injection) and RCIC; (4) change applicability
requirements and required actions for the reactor vessel water level-
low, level 3 function that isolates the RHR (residual heat removal)
system shutdown cooling system so that the function is required to be
operable in operational conditions 3, 4, and 5 to prevent inadvertent
loss of reactor coolant via the RHR shutdown cooling system; (5) remove
notes in Table 3.3.2-1, 3.3.2-2, and 4.3.1-1 related to maintenance on
leak detection temperature detectors and remove the note to TS 3.3.6
for Unit 1 related to a previous relief from TS 3.0.4; and reformat,
renumber, and/or reword existing requirements to incorporate the
changes listed above.
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:
I. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed TS changes increase the AOTs and STIs for actuation
instrumentation intended to detect or mitigate accidents; establish
required actions consistent with NUREG-1433 for some instruments
that are more specific but equivalent to existing required actions;
establish new requirements to prevent inadvertent loss of reactor
coolant via the RHR Shutdown Cooling System during OPERATIONAL
CONDITIONS 3, 4 and 5; and, eliminate notes that were intended to
provide one time only exemptions from certain requirements. The
proposed changes affect only those Technical Specification
requirements that govern operability, required actions and routine
testing of plant instruments that detect or mitigate accidents. The
proposed changes do [[Page 16195]] not affect any equipment or
requirements that are assumed to be initiators of any analyzed
events. Therefore, the proposed changes will not involve an increase
in the probability of occurrence of an accident previously
evaluated.
The proposed changes will not increase the consequences of an
accident previously evaluated because the changes will not involve
any physical changes to plant systems, structures, or components
(SSC), or the manner in which these SSC are operated, maintained,
modified, tested or inspected. The proposed changes will not alter
the operation of equipment assumed to be available for the
mitigation of accidents or transients by the plant safety analysis
or licensing basis. The proposed changes extend the intervals
between required performances of routine instrument testing. The
proposed changes also modify time limits allowed for operation with
inoperable instrument channels in situations when an inoperable
instrument channel would not prevent actuation of the associated
equipment. These changes are based on the demonstrated reliability
of these instruments and are justified by the analysis in References
1 through 8 [See February 10, 1995 application]. The small increases
in the probability that the proposed changes will result in an
equipment actuation failure has been determined in References 1
through 8 [See February 10, 1995 application] to be offset by safety
benefits such as a reduction in the number of inadvertent
actuations, a reduction in wear due to excessive testing, and better
utilization of plant personnel and resources. These changes will not
allow continuous plant operation with plant conditions such that a
single failure will result in a loss of any safety function.
Proposed changes to required actions and completion times for
instrumentation associated with the ADS initiation, Recirculation
Pump Trip, and pump suction lineup for HPCI and RCIC make the
required actions and completion times consistent with NUREG-1433,
Standard Technical Specifications for General Electric Plants, BWR/
4, Revision 0 (Reference 12). These changes are also consistent with
the assumptions used in References 1 through 8 [See February 10,
1995 application]. Therefore, these changes establish or maintain
adequate assurance that components are operable when necessary for
the prevention or mitigation of accidents or transients and that
plant variables are maintained within limits necessary to satisfy
the assumptions for initial conditions in the safety analysis. In
addition, the proposed change provides the benefit of avoiding an
unnecessary shutdown transient when appropriate measures are
available to compensate for the inoperable instrumentation.
Therefore, the proposed changes will not increase the consequences
of an accident previously evaluated.
There is no significant increase in the probability or
consequences of an accident previously evaluated resulting from
changes that reformat, renumber, and/or reword existing requirements
to incorporate the changes above or from the removal of notes that
were intended for one time only use and are no longer applicable.
II. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This proposed change will not involve any physical changes to
plant systems, structures, or components (SSC), or the manner in
which these SSC are operated, maintained, modified, tested, or
inspected. The changes in normal plant operation are consistent with
the current safety analysis assumptions. Therefore, this change will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
III. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed TS changes: increase the AOTs and STIs for
actuation instrumentation intended to detect or mitigate accidents;
establish required actions consistent with NUREG-1433 for some
instruments that are more specific but equivalent to existing
required actions; establish new requirements to prevent inadvertent
loss of reactor coolant via the RHR Shutdown Cooling System during
Operational Conditions 3, 4 and 5; and, eliminate notes that were
intended to provide one time only exemptions from certain
requirements.
There is no significant reduction in the margin of safety
resulting from changes to the minimum surveillance test intervals
(STIs) and allowable out-of-service times (AOTs) for the testing
and/or repair of instrumentation. This conclusion is based on the
demonstrated reliability of these instruments and is justified by
the analysis in References 1 through 8 [See February 10 1995
application]. The small increases in the probability that the
proposed changes will result in an equipment actuation failure has
been determined in References 1 through 8 [See February 10, 1995
application] to be offset by safety benefits such as a reduction in
the number of inadvertent actuations, a reduction in wear due to
excessive testing.
These changes will not allow continuous plant operation with
plant conditions such that a single failure will result in a loss of
any safety function.
There is no significant reduction in the margin of safety
resulting from changes to required actions and completion times for
instrumentation associated with the ADS initiation, Recirculation
Pump Trip, and pump suction lineup for HPCI and RCIC. These changes
make the required actions and completion times consistent with
NUREG-1433, Standard Technical Specifications for General Electric
Plants, BWR/4. These changes are also consistent with the
assumptions used in References 1 through 8 [See February 10, 1995
application]. Therefore, these changes establish or maintain
adequate assurance that components are operable when necessary for
the prevention or mitigation of accidents or transients and that
plant variables are maintained within limits necessary to satisfy
the assumptions for initial conditions in the safety analysis. In
addition, the proposed change provides the benefit of avoiding an
unnecessary shutdown transient when appropriate measures are
available to compensate for the inoperable instrumentation.
Additionally, the proposed required actions ensure that actions to
mitigate loss of single failure tolerance are initiated within 24
hours (12 hours for RPS) in accordance with the results of the
analyses in References 1 through 8 [See February 10, 1995
application] and action to mitigate a loss of instrument function is
initiated within 1 hour. Therefore, these changes will not allow
continuous plant operation with plant conditions such that a single
failure will result in a loss of any safety function. The
Pennsylvania Power & Light Company performed reviews that confirmed
the analyses in References 1 through 8 [See February 10, 1995
application] are applicable to SSES and that there would be no
effect on the identification of excessive instrument setpoint drift
as a result of increasing the minimum interval between instrument
functional tests from monthly to quarterly.
There is no significant reduction in the margin of safety
resulting from changes that reformat, renumber, and/or reword
existing requirements to incorporate the changes above or from the
removal of notes that were intended for one time only use and are no
longer applicable.
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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Public Service Electric and Gas
Company,Delmarva Power and Light Company, and Atlantic City
Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: January 13, 1995
Description of amendment request: The proposed changes concern a
revision to the frequency of calibration for the Local Power Range
Monitor (LPRM) signals from every 6 weeks to every 2000 Megawatt Days
per Standard Ton (MWD/ST).
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: [[Page 16196]]
1. The proposed change does not involve a significant increase
in the probability of consequences of an accident previously
evaluated.
This change does not affect the operation of any equipment. The
change does not affect the fundamental method by which the LPRMs are
calibrated. The increased time between required LPRM calibrations
does not affect either the initiator of any accident previously
evaluated or any equipment required to mitigate the consequences of
an accident, or the isotopic inventory in the fuel. Thus, the change
does not increase either the probability or the radiological
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 previously evaluated.
The proposed change does not introduce a new mode of plant
operation and does not involve the installation of any new equipment
or modifications to the plant. Therefore, it does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The GETAB determination of the Maximum Critical Power Ratio
(MCPR) Safety Limit allows a maximum total nodal uncertainty of the
TIP readings (of which the LPRM Update uncertainty is a part) of
8.7%. The change in LPRM calibration frequency results in an LPRM
Update uncertainty of 4.2% nodal power. This, combined with the
other uncertainties which comprise the total TIP readings
uncertainty, yields a total TIP readings uncertainty of less than
the allowed 8.7%. Thus the change in LPRM calibration frequency will
not affect the MCPR Safety Limit.
The LPRMs are utilized as input to the APRM and RBM systems. The
primary safety function of the APRM system is to initiate a scram
during core-wide neutron flux transients before the actual core-wide
neutron flux level exceeds the safety analysis design basis. This
prevents fuel damage from single operator errors or equipment
malfunctions. The APRMs are calibrated at least twice per week to
the plant heat balance, utilize a radially and axially diverse group
of LPRMs as input and are utilized to detect changes in average, not
local, power changes. Therefore, the effects of decreasing the LPRM
calibration frequency on the APRM system responses will be minimal
due to any individual LPRM drift being practically canceled out (due
to diversity of input) and/or due to the frequent recalibration of
the APRMs to an independent power calculation (the heat balance).
Thus, decreasing the LPRM calibration frequency will not
significantly impact the performance of the APRM system's scram
function, and there is no impact on transient delta-CPRs.
The RBM system is utilized in the mitigation of a Rod Withdrawal
Error (RWE). The RBM system is designed to prevent the operator from
increasing the local power significantly when withdrawing a control
rod. On each selection of a control rod, the average of the
assigned, unbypassed LPRMs is adjusted to equal a 100% reference
signal for each of the two RBM channels. Each RBM channel
automatically limits the local thermal margin changes by limiting
the allowable change in local average neutron flux to the RBM
setpoint. If the local average neutron flux change is greater than
that allowed by the RBM setpoint, within either RBM channel, the rod
withdrawal permissive is removed preventing further movement. Since
the change in local neutron flux is calculated from the change in
the average of the LPRM readings, and calibrated on every rod
selection to the reference signal, offsets in individual LPRM
readings due to calibration differences are effectively eliminated
for a given RBM setpoint. Therefore, the constraints on the
withdrawal of any given rod are unchanged and there will not be any
increase in RWE delta-CPR.
Since the MCPR Safety Limit is unaffected and the delta-CPR
values are unchanged, the cycle CPR limits are unchanged. Therefore,
the change in the frequency of LPRM calibration does not result in a
reduction in a margin of safet