[Federal Register Volume 64, Number 115 (Wednesday, June 16, 1999)]
[Notices]
[Pages 32284-32296]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-15098]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from May 21, 1999, through June 4, 1999. The last
biweekly notice was published on June 2, 1999 (64 FR 29707).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By July 19, 1999, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the
[[Page 32285]]
proceeding, but such an amended petition must satisfy the specificity
requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendments request: May 23, 1997, as revised by letters
dated September 27, 1998, and May 26, 1999.
Description of amendments request: The proposed amendments would
revise Technical Specification (TS) Limiting Condition of Operation
(LCO) 3.4.14 and TS Sections 5.5.9 and 5.6.8 to allow the use of steam
generator (SG) tube sleeves as an alternative to plugging defective SG
tubes. The May 26, 1999, letter completely revised the May 23, 1997,
request for amendments, and this notice supersedes the original Federal
Register notice dated July 30, 1997 (62 FR 40845).
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 TS LCO 3.4.14.d and e will replace the
leakage limits of 1 gallon per minute (gpm) primary to secondary
leakage through all SGs and 720 gallon per day (gpd) through any one
SG with a new limit of 150 gpd through any one SG. This is a more
restrictive change. A TS limit of 150 gpd primary to secondary
Leakage through any one steam generator is significantly less than
the initial conditions assumed in the safety analyses. The 150 gpd
limit is based on operating experience as an indication of one or
more propagating tube leak mechanisms. The Steam Generator Tube
Surveillance Program described in TS Section 5.5.9 ensures that the
structural integrity of the SG tubes is maintained. The leakage rate
limit of 150 gpd for any one SG provides additional assurance
against tube rupture at normal and faulted conditions and provides
additional assurance that cracks will not propagate to burst prior
to detection by leakage monitoring methods and commencement of plant
shutdown. Therefore, this change to TS LCO 3.4.14.e will not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes to TS 5.5.9 will add inservice inspection
requirements for SG tube sleeves. These requirements will ensure
that all installed SG tube sleeves will be inspected prior to
initial operation and routinely thereafter, to assure the capability
of each sleeve to perform its design function during each operating
cycle. The tube sleeves will be the Combustion Engineering, Inc. (CE
or ABB-CE) Leak Tight sleeves, as described in CE report CEN-630-P,
``Repair of \3/4\'' O.D. Steam Generator Tubes Using Leak Tight
Sleeves,'' Revision 02, dated June 1997. (This proprietary report is
provided as Enclosure 4 with this submittal.) The tube sleeve
dimensions, materials and joints are designed to the applicable ASME
[American Society of Mechanical Engineers] Boiler and Pressure
Vessel code requirements. An extensive test program was performed
that demonstrated that the sleeves will fulfill their intended
function as leak tight structural members. Evaluation of sleeved
tubes indicates no detrimental effects on the sleeve-tube assembly
resulting from reactor coolant system flow, coolant chemistries, or
thermal and pressure conditions. Structural analyses of the sleeve-
tube assembly have established its integrity under normal and
accident conditions. Mechanical testing using ASME code stress
allowables was performed to support the analyses. Also, corrosion
tests were performed and revealed no evidence of sleeve or tube
corrosion considered detrimental under anticipated service
conditions. A sleeved tube will exhibit greater hydraulic resistance
and reduced heat transfer capability than an un-sleeved tube.
However, these effects are much less than would be imposed by taking
the tube out of service by plugging. Section 10.0 of CE report CEN-
630-P describes the analyses to determine the hydraulic and heat
transfer effects. Calculations using plant-specific information will
identify sleeve-to-plug equivalency ratios. The proposed changes to
the SG inservice inspection program will assure that sleeved SG
tubes will meet the structural requirements of tubes that are not
defective. The proposed sleeve plugging limit of 35% of nominal wall
will ensure that the sleeves remaining in service will perform their
design function. Also, installation of
[[Page 32286]]
sleeves will not significantly [a]ffect the primary system flow rate
or the heat transfer capability of the SGs. Therefore, this change
to TS section 5.5.9 will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The change to the SG reporting requirements in TS section 5.6.8
will ensure that the number of sleeved SG tubes will be reported to
the NRC along with the number of plugged tubes. This is an
administrative change that has no effect on the operation or
maintenance of the plant and will 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 TS LCO 3.4.14.d and e will replace the
leakage limits of 1 gpm primary to secondary leakage through all SGs
and 720 gpd through any one SG with a new limit of 150 gpd through
any one SG. This is a more restrictive change that will provide
added assurance against steam generator tube ruptures. Since the
current allowable primary to secondary leakage is being reduced,
this change will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes to TS section 5.5.9 for the SG inservice
inspection program will assure that sleeved SG tubes will meet the
structural requirements of tubes that are not defective. Also,
installation of sleeves will not significantly [a]ffect the primary
system flow rate or the heat transfer capability of the SGs.
Therefore, this change to TS section 5.5.9 will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The change to the SG reporting requirements in TS section 5.6.8
will ensure that the number of sleeved SG tubes will be reported to
the NRC along with the number of plugged tubes. This is an
administrative change that has no effect on the operation or
maintenance of the plant and will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed change to TS LCO 3.4.14.d and e will replace the
leakage limits of 1 gpm primary to secondary leakage through all SGs
and 720 gpd through any one SG with a new limit of 150 gpd through
any one SG. This is a more restrictive change that will provide
added assurance against steam generator tube ruptures. Since the
current allowable primary to secondary leakage is being reduced,
this change will not involve a significant reduction in a margin of
safety.
The proposed changes to TS section 5.5.9 for the SG inservice
inspection program will assure that sleeved SG tubes will meet the
structural requirements of tubes that are not defective. Also,
installation of sleeves will not significantly [a]ffect the primary
system flow rate or the heat transfer capability of the SGs.
Therefore, this change to TS section 5.5.9 will not involve a
significant reduction in a margin of safety.
The change to the SG reporting requirements in TS section 5.6.8
will ensure that the number of sleeved SG tubes will be reported to
the NRC along with the number of plugged tubes. This is an
administrative change that has no effect on the operation or
maintenance of the plant and will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 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: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Section Chief: Stephen Dembek.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: May 5, 1999.
Description of amendment request: The proposed amendments would
revise the basis for evaluation of the reactor building ventilation
(VR) system exhaust plenum masonry walls. Specifically, the amendment
would approve the use of different methodology and acceptance criteria
for the reassessment of certain masonry walls subjected to transient
pressurization loads resulting from a high energy line break.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The change involves reassessment of the VR exhaust plenum due to
a transient pressurization during a Main Steam Line Break (MSLB).
Since the transient pressurization is a result of the MSLB, and the
block walls and the dampers are not initiators of any accident, the
probability of an accident previously evaluated is not affected.
This analysis does not affect the total amount of radioactive
release due to the MSLB Outside of the Primary Containment, so the
total offsite dose consequences does not change. A small portion of
the release, which passes the dampers prior to closure, will now be
an elevated release via the plant ventilation stack instead of a
ground level release. The original analysis assumed the entire
release was a ground level release, and thus remains bounding for
the MSLB accident.
The Control Room and Auxiliary Electric Equipment Room (AEER)
dose consequences are impacted only slightly due to the small amount
of steam/air mixture released from the new pressure relief damper.
The steam/air mixture becomes mixed with the air volume in that area
of the Auxiliary Building but was all assumed to be available for
inleakage to the Control Room and AEER. The dose increase for the
Control Room and AEER is less than or equal to 0.05 Rem thyroid and
negligible change to the whole body dose, such that the dose due to
the MSLB accident remains much less than the DBA LOCA dose and
General Design Criteria 19. The MSLB accident dose consequences
remain bounded by the Design Basis Loss of Coolant Accident.
The effects of the steam released by the pressure relief damper
into the Auxiliary Building has been evaluated for environmental
qualification impact on systems, structures and components (SSCs) in
the area of the Auxiliary Building affected for both radiation and
steam/temperature affects. The effect on area temperature is about 4
deg.F and is above initial temperature for not more than 24 hours.
The change in humidity is negligible, and radiation dose impact is
small and bounded by previous calculations.
These consequences assume that the VR exhaust plenum masonry
walls do not rupture based on the design changes being made in
conjunction with the masonry wall reevaluation for each LaSalle Unit
that will prevent the failure of the VR exhaust plenum masonry
walls.
Therefore this proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The MSLB accident is previously analyzed but considered only
instantaneous closure of installed dampers. The reevaluation and
design changes extend the previous accident analysis to assure that
structures previously considered unaffected by the MSLB will
maintain their structural integrity. The block walls are static and
the dampers function in response to an accident, thus the analysis
method and design changes are not accident initiators. Therefore the
change does not create the possibility of a new [or] different kind
of accident from any accident previously evaluated.
The design changes being made in conjunction with the masonry
wall reevaluation for each LaSalle Unit that will prevent the
failure of the VR exhaust plenum masonry walls are as follows:
(1) Installation of a pressure relief damper,
(2) An excess-flow check damper, and
(3) Required masonry wall support improvements in the reactor
building ventilation exhaust plenum for each Unit.
The reevaluation of the masonry walls uses different load
factors and load combinations
[[Page 32287]]
as well as reduced acceptance criteria than previously used for
these walls. The change in the evaluation does not cause the rupture
or failure of the effected masonry walls, since the evaluation shows
the walls remain intact.
The installation of the above design changes, in conjunction
with masonry wall analysis assure that the subject masonry walls
will not rupture or fail. Therefore, SSCs that would be affected by
wall rupture can fulfill their intended function, maintaining the
consequences of previously evaluated accident the same.
The new pressure relief damper and excess-flow check damper are
safety-related and are analyzed to function under the conditions
created by the MSLB. In addition, the dampers and the duct they are
installed in have been analyzed to assure no failure will occur
during an Operating Basis Earthquake (OBE) or Safe Shutdown
Earthquake (SSE).
Based on an analysis of potential failure modes in accordance
with ANSI/ANS-58.9-1981, ``Single Failure Criteria for Light Water
Reactor Safety-Related Fluid Systems,'' Paragraph 4.1, the active
function of the pressure relief damper and excess flow check damper
are considered exempted from consideration of single failure. The
principles governing operation of the dampers are simple and direct
and not subject to change or deterioration with time, similar to the
function of a code safety relief valve and a swing check valve. With
periodic testing of the dampers, continued reliable performance is
assured.
The dampers are designed and set so that the pressures created
by normal ventilation flow changes do not cycle the dampers, and
thus the new dampers do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Administrative controls will be in place prior to implementation
of this change to assure the testing and maintenance is periodically
performed in accordance with vendor recommendations. These dampers
will be included as equipment required to be monitored/maintained,
because the function performed by the dampers is within the scope of
the Maintenance Rule, 10 CFR 50.65.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
Does the change involve a significant reduction in a margin of
safety?
Originally, no masonry walls were evaluated for HELB
pressurization effects, because the walls were considered protected
by the isolation dampers. However, the original design methodology
for masonry did include load combinations including Pa:
Abnormal
1.0D + 1.0L + 1.5Pa
Abnormal/Severe Environment
1.0D + 1.0L + 1.25Pa + 1.25Eo
Abnormal/Extreme Environment
1.0D + 1.0L + 1.0Pa + 1.0Ess,
Where D is Dead Load; L is Live Load; Pa is
pressurization due to HELB; Eo is Loads generated by the
Operating Basis Earthquake (OBE); and Ess is Loads
generated by the Safe Shutdown Earthquake (SSE).
The current reevaluation was required due to determination that
some block walls in the LaSalle Auxiliary Building are affected by a
transient pressurization due to a MSLB. The specific changes from
the original analyses involve the following for loads and load
combinations.
1. Abnormal:
1.0D + 1.0L + 1.0PHELB
2. Abnormal/severe environmental:
1.0D + 1.0L + [(1.1Eo)2 +
1.0PHELB2]\1/2\
3. Abnormal/extreme environmental:
1.0D + 1.0L + [1.0Ess2 +
1.0PHELB2]\1/2\
Where:
(1) PHELB is the short-term differential pressurization
load on the VR plenum masonry walls resulting from non-instantaneous
opening/closure of the protection dampers.
(2) The Load Factor on pressure due to HELB is 1.0 for all cases.
(3) The Loading Combination of pressure and seismic is the Square
Root of the Sum of Squares (SRSS).
LaSalle has selected the proposed load combinations in
consideration of the following:
Isolation, check, and relief dampers protect the walls;
therefore the pressurization effects are not sustained, but are
transient in nature.
The transient pressurization effect (PHELB) is
derived from a conservative detailed analysis of an instantaneous
HELB combined with non-instantaneous damper opening/closure. Due to
the precise nature and conservatism of this HELB analysis, there is
little uncertainty in PHELB .
Therefore a load factor of 1.0 is used for all abnormal load
combinations.
PHELB is a short duration, dynamic load. Accordingly,
the seismic and transient HELB pressurization loads are combined
using the Square Root of Sum of the Squares (SRSS) method because
the peak effects of these dynamic loads are unlikely to occur
simultaneously. This combination method is used in the analysis of
other components such as component supports.
The proposed load combinations accordingly provide a
conservative basis for reassessment of the VR exhaust plenum masonry
wall systems.
In regards to the masonry acceptance criteria, the original
acceptance criteria used for this condition are the National
Concrete Masonry Associations (NCMA) ``Specification for the Design
and Construction of Load Bearing Masonry--1979'' allowable stresses
times a 1.67 factor. These allowable stresses correspond to stress
equal to the modulus of rupture (fr) of the masonry
divided by a factor of safety of 3.35. During reviews to address
masonry wall issues per NRC IE Bulletin 80-11, six walls did not
meet this acceptance criteria. The acceptance criteria used for
these walls was for fr values determined from testing at
Clinton Power Station divided by a factor of safety of 2.5. This
acceptance criteria was accepted by the NRC for LaSalle in
Supplement 5 of NUREG 0519, Safety Evaluation Report related to the
Operation of LaSalle County Station, Units 1 and 2. The VR exhaust
plenum walls will use the same acceptance criteria for the transient
HELB pressurization cases.
The minimum masonry safety factor for the LaSalle Unit 2 walls
affected by the HELB loads range from 2.6 to 3.1 with one wall
having a safety factor of 4.9.
Masonry wall steel support members were originally designed for
this condition elastically to the American Institute of Steel
Construction's (AISC) ``Steel Construction Manual--Seventh Edition''
allowable stresses times a 1.6 factor. In the reassessment of these
members due to the transient HELB pressurization, elasto-plastic
behavior is allowed (with a ductility ratio limit of 10). It is
appropriate to consider them similar to high-energy line break
systems that will maintain their integrity as they absorb the energy
of the incidental pressure excursion.
High-energy line breaks are discussed in Section 3.6 of the
UFSAR. The discussion in this section focuses on the design of pipe
whip restraints, and in Table 3.6-6 acceptance criteria are
provided. This table shows that the energy absorbing portions of the
pipe whip restraint are allowed to go plastic, thereby absorbing
energy. While Table 3.6-6 of the UFSAR deals with energy absorbing
portions of the pipe whip restraints, wide-flange shapes are not
addressed. Wide-flange shapes absorb energy through flexural
deformations.
Guidance on appropriate acceptance criteria for flexural members
is provided in Appendix A to SRP 3.5.3, ``Barrier Design
Procedures.'' This appendix indicates that for tension due to
flexure in structural steel members, a ductility ratio value not to
exceed 10.0 is acceptable. SRP 3.8.4, paragraph III.5 also notes
that some localized points on the structure, the allowable stresses
specified for ``structural steel'' may be exceeded, provided that
integrity of the structure is not affected.
Note that only one of the Unit 2 walls affected by these HELB
loads required the use of the elasto-plastic acceptance criteria for
two structural steel members.
In summary, these alternate criteria for reassessment of the
integrity of the LaSalle Reactor Building Ventilation Exhaust Plenum
masonry walls in conjunction with the design changes adding a
pressure relief damper, an excess flow check damper and masonry wall
support steel changes, assures that the walls will maintain their
integrity during a MSLB. The safety factor is reduced; however, the
walls have sufficient strength and safety margin to maintain
structural integrity and thus perform their intended safety function
during the pressurization transient due to a MSLB accident.
Therefore, these changes do not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois
[[Page 32288]]
Valley Community College, Oglesby, Illinois 61348-9692.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix, County, Michigan
Date of amendment request: May 11, 1999 (Accession No. 9905170189).
Description of amendment request: The proposed amendment would
delete from the Defueled Technical Specifications (DTS) the definition
for site boundary and Figure 5.1-1, Big Rock Point Site Map, and revise
the description of the Big Rock Point site under subsection 5.1. The
amendment also proposes editorial changes associated with the above
proposed revisions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
In accordance with 10 CFR 50.91, Consumers Energy Company has
made a determination that the proposed amendment does not involve
significant hazards considerations. Consumers Energy Company has
concluded that the proposed amendment will 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 proposed change is administrative in nature and has no
[e]ffect on the health and safety of the public. There is no
reduction or elimination of federal regulatory requirements
associated with the proposed amendment. The information being
removed from the Defueled Technical Specifications is unnecessary
since Site Boundary is already defined in 10 CFR Part 20, and the
site map [Defueled Technical Specification Figure 5.1-1] is already
provided in the Updated Final Hazards [Summary] Report. Furthermore,
the proposed changes are consistent with the guidance provide in
NUREG-1625 ['Proposed Standard Technical Specifications for
Permanently Defueled Westinghouse Plants''].
The proposed change does not:
(1) Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed amendment does not change the site boundary as it
currently exists. Deleting the Site Boundary definition and changing
the upper case characters to lower case throughout the DTS and the
Bases where it appears, and deleting the site figure from the DTS
and related references will not increase the probability or
consequences of a new or different kind of accident previously
evaluated. This proposed change is administrative in nature and does
not involve fuel handling or affect or modify any system, structure
or component.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed amendment does not change the site boundary as it
currently exists. Deleting the Site Boundary definition and changing
the upper case characters to lower case throughout the DTS and the
Bases where it appears, and deleting the site figure from the DTS
and related references will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This proposed change is administrative in nature and does not
involve fuel handling or affect or modify any system, structure or
component.
(3) Involve a significant reduction in the margin of safety.
The proposed changes do not involve any physical changes to the
plant or plant procedures. There will be no 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: North Central Michigan
College, 1515 Howard Street, Petosky, MI 49770.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Dr. Michael T. Masnik.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: October 2, 1998, supplemented May 13,
1999.
Description of amendment request: The proposed amendments would
resolve an unreviewed safety question involving use of credit for
reactor building overpressure in the licensing basis for the available
net positive suction head for the reactor building spray pumps and the
low pressure injection pumps. If approved, the appropriate changes
would be incorporated in the Oconee 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.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
The reactor building spray (RBS) and low pressure injection
(LPI) systems are not considered as initiators of any analyzed
event, therefore, this change has no impact on the probability of an
event previously analyzed.
The consequences of a previously analyzed event are dependent on
the initial conditions assumed for the analysis, the availability
and successful functioning of the equipment assumed to operate in
response to the analyzed event, and the set points at which these
actions are initiated. The proposed change permits limited reactor
building overpressure to be credited in the calculation of available
net positive suction head (NPSH) for the RBS and LPI pumps for a
limited period of time during the sump recirculation phase. It is
supported by calculations which demonstrate that adequate reactor
building overpressure will be available to ensure the RBS and LPI
systems will be capable of performing their safety functions. Thus,
the proposed change does not significantly increase the consequences
of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from the accidents previously evaluated?
The proposed change permits limited reactor building
overpressure to be credited in the calculation of available NPSH for
the RBS and LPI pumps for a limited period of time during the sump
recirculation phase. It does not involve a physical alteration of
the plant. The proposed change is supported by calculations which
demonstrate that adequate reactor building overpressure will be
available to ensure the RBS and LPI systems will be capable of
performing their safety functions. This change will not alter the
manner in which the RBS or LPI system is initiated, nor will the
function demands on the RBS or LPI system be changed. Thus, the
proposed change does not create the possibility of a new or
different kind of accident.
3. Involve a significant reduction in a margin of safety?
The proposed change permits limited reactor building
overpressure to be credited in the calculation of available NPSH for
the RBS and LPI pumps for a limited period of time during the sump
recirculation phase. Crediting a slight amount of overpressure does
not result in a significant reduction in the margin of safety,
because conservative analyses demonstrate that adequate reactor
building overpressure will be available to ensure the RBS and LPI
systems will be capable of performing their safety functions. Thus,
the proposed change does not involve a significant reduction in a
margin of safety.
Duke has concluded based on the above information that there are
no significant hazards involved in this LAR.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 32289]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: May 11, 1999.
Description of amendment request: The proposed amendments would:
(a) revise the pressure-temperature (P-T) limits of Technical
Specification (TS) 3.4.3 for heatup, cooldown, and inservice test
limitations for the Reactor Coolant System to a maximum of 33 Effective
Full Power Years; (b) revise TS 3.4.12, Low Pressure Overpressure
Protection System (LTOP), to reflect the revised P-T limits of the Unit
1, 2, and 3 reactor vessels; (c) permit operation during LTOP
conditions with two reactor coolant pumps in operation in a single
loop; and (d) relax the LTOP operating envelope, thereby reducing
potential challenges to the reactor coolant system power operated
relief valves.
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.
A. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
No.
These proposed Technical Specification (TS) changes were
developed utilizing the procedures of ASME XI, Appendix G, in
conjunction with Code Cases N-514, N-588 and N-626, as described in
the Technical Justification. Usage of these procedures provides
compliance with the underlying intent of 10 CFR 50 Appendix G and
provide safety limits and margins of safety that ensure failure of a
reactor vessel will not occur.
The proposed changes do not impact the capability of the reactor
coolant pressure boundary (i.e., no change in operating pressure,
materials, seismic loading, etc.) and therefore do not increase the
potential for the occurrence of a loss of coolant accident (LOCA).
The changes do not modify the reactor coolant system pressure
boundary, nor make any physical changes to the facility design,
material, or construction standards. The probability of any design
basis accident (DBA) is not affected by this change, nor are the
consequences of any DBA affected by this change. The proposed
Pressure-Temperature (P-T) limits, Low Temperature Overpressure
(LTOP) limits and setpoints, and allowable operating reactor coolant
pump combinations are not considered to be an initiator or
contributor to any accident analysis addressed in the Oconee UFSAR.
The proposed changes do not adversely affect the integrity of
the RCS such that its function in the control of radiological
consequences is affected. Radiological off-site exposures from
normal operation and operational transients, and faults of moderate
frequency do not exceed the guidelines of 10 CFR 100. In addition,
the proposed changes do not affect any fission product barrier. The
revised PORV LTOP setpoint is established to protect reactor coolant
pressure boundary. The changes do not degrade or prevent the
response of the PORV or safety-related systems to previously
evaluated accidents. In addition, the changes do not alter any
assumption previously made in the mitigation of the radiological
consequences of an accident previously evaluated.
Therefore, the probability or consequences of an accident
previously evaluated will not be increased by approval of the
requested changes.
B. Create the possibility of a new or different kind of accident
from the accident previously evaluated?
No.
The proposed license amendment revises the Oconee reactor vessel
P-T limits, LTOP limits and setpoints, and allowable operating
reactor coolant pumps combinations. Compliance with 10 CFR 50
Appendix G, includes utilization of ASME XI, Appendix G, as modified
by Code Cases N-514, N-588 and N-626 to meet the underlying intent
of the regulations.
Operation of Oconee in accordance with these proposed Technical
Specifications changes will not create any failure modes not bounded
by previously evaluated accidents. Consequently, approval of these
changes will not create the possibility of a new or different
accident from any accident previously evaluated.
C. Involve a significant reduction in a margin of safety?
No.
The proposed Technical Specification (TS) changes were developed
utilizing the procedures of ASME XI, Appendix G, in conjunction with
Code Cases N-514, N-588 and N-626, as described in the Technical
Justification. Usage of these procedures provides compliance with
the underlying intent of 10 CFR 50 Appendix G and provides safety
limits and margins of safety which ensure failure of a reactor
vessel will not occur.
No plant safety limits, set points, or design parameters are
adversely affected. The fuel, fuel cladding, and Reactor Coolant
System are not impacted. Therefore, there will be no significant
reduction in any margin of safety as a result of approval of the
requested changes.
Duke has concluded based on this information there are no
significant hazards considerations involved in this amendment
request.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: May 17, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications associated with the enabling of the
Oscillation Power Range Monitor (OPRM) instrumentation reactor
protection system (RPS) trip function. The OPRM is designed to detect
the onset of reactor core power oscillations resulting from thermal-
hydraulic instability and suppresses them by initiating a reactor scram
via the RPS trip logic.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change specifies limiting conditions for
operations, required actions and surveillance requirements of the
OPRM system and allows operation in regions of the power to flow map
currently restricted by the requirements of Interim Corrective
Actions (ICAs) and certain limiting conditions of operation of
Technical Specifications (TS) 3.4.1. The OPRM system can
automatically detect and suppress conditions necessary for thermal-
hydraulic (T-H) instability. A T-H instability event has the
potential to challenge the Minimum Critical Power (MCPR) safety
limit. The restrictions of the ICAs and TS 3.4.1 were imposed to
ensure adequate capability to detect and suppress conditions
consistent with the onset of T-H oscillations that may develop into
a T-H instability event. With the installation of the OPRM System,
these restrictions are no longer required.
[[Page 32290]]
The probability of a T-H instability event is most significantly
impacted by power to flow conditions such that only during operation
inside specific regions of the power to flow map, in combination
with power shape and inlet enthalpy conditions, can the occurrence
of an instability event be postulated to occur. Operation in these
regions may increase the probability that operation with conditions
necessary for a T-H instability can occur.
However, when the OPRM is operable with operating limits as
specified in the COLR [Core Operating Limits Report], the OPRM can
automatically detect the imminent onset of local power oscillations
and generate a trip signal. Actuation of an RPS trip will suppress
conditions necessary for T-H instability and decrease the
probability of a T-H instability event. In the event the trip
capability of the OPRM is not maintained, the proposed change
includes actions which limit the period of time before the effected
OPRM channel (or RPS system) must be placed in the trip condition.
If these actions would result in a trip function, an alternate
method to detect and suppress thermal hydraulic oscillations is
required. In either case the duration of this period of time is
limited such that the increase in the probability of a T-H
instability event is not significant. Therefore the proposed change
does not result in a significant increase in the probability of an
accident previously evaluated.
An unmitigated T-H instability event is postulated to cause a
violation of the MCPR safety limit. The proposed change ensures
mitigation of T-H instability events prior to challenging the MCPR
safety limit if initiated from anticipated conditions by detection
of the onset of oscillations and actuation of an RPS trip signal.
The OPRM also provides the capability of an RPS trip being generated
for T-H instability events initiated from unanticipated but
postulated conditions. These mitigating capabilities of the OPRM
system would become available as a result of the proposed change and
have the potential to reduce the consequences of anticipated and
postulated T-H instability events. Therefore, the proposed change
does not significantly increase the consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change specifies limiting conditions for
operations, required actions and surveillance requirements of the
OPRM system and allows operation in regions of the power to flow map
currently restricted by the requirements of ICAs and TS 3.4.1. The
OPRM system uses input signals shared with APRM [Average Power Range
Monitor] and rod block functions to monitor core conditions and
generate an RPS trip when required. Quality requirements for
software design, testing, implementation and module self-testing of
the OPRM system provide assurance that no new equipment malfunctions
due to software errors are created. The design of the OPRM system
also ensures that neither operation nor malfunction of the OPRM
system will adversely impact the operation of other systems and no
accident or equipment malfunction of these other systems could cause
the OPRM system to malfunction or cause a different kind of
accident. Therefore, operation with the OPRM system does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Operation in regions currently restricted by the requirements of
ICAs and TS 3.4.1 is within the nominal operating domain and ranges
of plant systems and components for which postulated equipment and
accidents have been evaluated. Therefore operation within these
regions does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed change which specifies limiting conditions for
operations, required actions and surveillance requirements of the
OPRM system and allows operation in certain regions of the power to
flow [map] does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change specifies limiting conditions for
operations, required actions and surveillance requirements of the
OPRM system and allows operation in regions of the power to flow map
currently restricted by the requirements of ICAs and TS 3.4.1.
The OPRM system monitors small groups of LPRM signals for
indication of local variations of core power consistent with T-H
oscillations and generates an RPS trip when conditions consistent
with the onset of oscillations are detected. An unmitigated T-H
instability event has the potential to result in a challenge to the
MCPR safety limit. The OPRM system provides the capability to
automatically detect and suppress conditions which might result in a
T-H instability event and thereby maintains the margin of safety by
providing automatic protection for the MCPR safety limit while
significantly reducing the burden on the control room operators. In
the event the trip capability of the OPRM is not maintained, the
proposed change includes actions which limit the period of time
before the effected OPRM channel (or RPS system) must be placed in
the trip condition. If these actions would result in a trip
function, an alternate method to detect and suppress thermal
hydraulic oscillations is required. Since, in either case, the
duration of this period of time is limited so that the increase in
the probability of a T-H instability event is not significant.
Operation with the OPRM system does not involve a significant
reduction in a margin of safety.
Operation in regions currently restricted by the requirements of
ICAs and TS 3.4.1 is within the nominal operating domain assumed for
identifying the range of initial conditions considered in the
analysis of anticipated operational occurrences and postulated
accidents. Therefore, operation in these regions does not involve a
significant reduction in the margin of safety.
The proposed change, which specifies limiting conditions for
operations, required actions and surveillance requirements of the
OPRM system and allows operation in certain regions of the power to
flow map, 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: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station
(VCSNS), Unit No. 1, Fairfield County, South Carolina
Date of amendment request: May 17, 1999.
Description of amendment request: The proposed amendment would
change VCSNS Technical Specification 3.7.1.3 ``Condensate Storage
Tank--Limiting Conditions for Operation'' to revise the tank minimum
contained water volume from 172,000 gallons to 179,850 gallons.
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. This request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
FSAR [Final Safety Analysis Report] 10.4.9.1 states that minimum
required usable volume for the Condensate Storage Tank (CST) is
158,570 gallons based on maintaining the plant at HOT STANDBY
conditions for eleven hours. This volume has already been adjusted
for both plant uprate conditions and replacement steam generator
requirements. This change to LCO [Limiting Condition for Operation]
3.7.1.3 will ensure that 160,054 gallons is maintained in the CST,
being available and dedicated to the Emergency Feedwater (EFW)
System. Thus, this change will ensure that the EFW System has an
adequate water supply to perform its design basis function in regard
to maintaining the plant in HOT STANDBY condition.
2. This request does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This change increases the minimum required volume of water in
the CST, thus
[[Page 32291]]
ensuring that the EFW System can perform its required safety
function. The maximum and normal water levels in the CST are not
being changed. Therefore, no new failure modes of the CST, or
flooding concerns are created.
3. This request does not involve a significant reduction in a
margin to safety[.]
This change does not reduce any margin associated with the CST
inventory available to the EFW. In fact, a small gain in margin
(less than 1%) is realized by specifying the minimum required volume
based on the maximum volume available due to nozzle locations and
other physical characteristics of the tank instead of the minimum
required to maintain HOT STANDBY for 11 hours. Additionally, the
requirement for sufficient CST volume to maintain HOT STANDBY for 11
hours is still met and the Service Water System still provides the
long term supply of safety grade cooling water to the EFW System.
The Service Water supply is not affected by this change, and thus
the margin for safety grade cooling water to the EFW System (or
safety grade cooling of the RCS) is not 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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc, Docket No. 50-348 Joseph M.
Farley Nuclear Plant Unit 1, Houston County, Alabama
Date of amendment request: April 30, 1999.
Description of amendment request: The proposed amendment would add
an additional condition to the Farley Nuclear Plant (FNP), Unit 1
license. This condition would allow cycle 16 operation based on a risk-
informed approach to evaluate steam generator tube structural
integrity.
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 significantly increase the
probability or consequences of an accident previously evaluated in
the FSAR [Final Safety Analysis Report]. The probability of tube
burst is slightly increased as a result of this proposed amendment
but is within current industry guidance. Therefore, the probability
of a previously evaluated accident are not significantly increased.
There is no change in the FNP design basis as a result of this
change and, as a result, this change does not involve a significant
increase in the consequences of an accident previously evaluated.
The proposed changes to the TSs [technical specifications] do
not increase the possibility of a new or different kind of accident
than any accident already evaluated in the FSAR. No new limiting
single failure or accident scenario has been created or identified
due to the proposed changes. Safety-related systems will continue to
perform as designed. The proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
The proposed changes do not involve a significant reduction in
the margin of safety. There is no impact in the accident analyses.
These proposed changes are technically consistent with the
requirements of NEI [Nuclear Energy Institute] 97-06, ``Steam
Generator Program Guidelines,'' Draft Regulatory Guide DG 1074,
``Steam Generator Tube Integrity,'' and Regulatory Guide (RG) 1.174,
``An Approach for Using Probabilistic Risk Assessment In Risk-
Informed Decisions on Plant-Specific Changes to the Licensing
Basis.'' Thus the proposed changes do not involve a significant
reduction in the margin of safety.
Accordingly, SNC [Southern Nuclear Operating Company] has
determined that the proposed amendment to the Facility Operating
License NPF-2 does not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama.
NRC Section Chief: Richard L. Emch, Jr.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: May 3, 1999.
Description of amendment request: The proposed changes will modify
the Technical Specifications to ensure the emergency ventilation system
is maintained operable consistent with the assumptions in the
radiological dose consequence reanalysis from a Large Break Loss-of-
Coolant Accident and to clearly identify that the ventilation system is
a shared system between the two units.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. There is no significant change in the probability or
consequences of an accident previously evaluated. There are no
system changes which would increase the probability of occurrence of
an accident. The dose consequences of the accidents have been
reviewed, and in some cases the doses at the EAB [exclusion area
boundary] * * * and the doses to the control room personnel were
found to increase. However, this increase is not significant because
the revised doses remain below the limits of 10 CFR 100 and below
the limits of GDC [General Design Criterion]--19 of Appendix A of 10
CFR 50.
2. No new accident types or equipment malfunction scenarios have
been introduced. Therefore, the possibility of an accident of a
different type than any evaluated previously in the UFSAR [Updated
Final Safety Analysis Report] is not created.
3. There is no significant reduction in the margin of safety, as
the revised dose calculations for all accidents continue to meet the
appropriate GDC-19 limits.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Section Chief: Richard L. Emch, Jr.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: May 6, 1999.
Description of amendment request: The proposed changes will modify
the Technical Specifications, revising the surveillance frequency for
the Reactor Trip System (RTS) and Engineered Safety Features Actuation
System (ESFAS) analog instrumentation
[[Page 32292]]
channels and also revising the allowed outage time and action times for
the RTS and ESFAS analog instrumentation channels and the actuation
logic.
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:
Virginia Electric and Power Company has reviewed the
requirements of 10 CFR 50.92 as they relate to the proposed Reactor
Trip System (RTS) and Engineered Safety Features Actuation System
(ESFAS) Technical Specification changes for the North Anna Units 1
and 2 and determined that a significant hazards consideration is not
involved. In support of this conclusion, the following evaluation is
provided.
Criterion 1--Operation of North Anna Units 1 and 2 in accordance
with the proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The determination that the results of the
proposed changes remain within acceptable criteria was established
in the SER(s) [Safety Evaluation Reports] prepared for WCAP-10271,
WCAP-10271 Supplement 1, WCAP-10271 Supplement 2, WCAP-10271
Supplement 2, Revision 1 and WCAP-14333 issued by letters dated
February 21, 1985, February 22, 1989, April 30, 1998, and July 15,
1998.
Implementation of the proposed changes is expected to result in
an increase in total RTS and ESFAS yearly unavailability. The
proposed changes have been shown to result in a small increase in
the core damage frequency (CDF) due to the combined effects of
increased RTS and ESFAS unavailability and reduced inadvertent
reactor trips.
The values determined by the WOG [Westinghouse Owners Group] and
presented in the WCAP for the increase in CDF were verified by
Brookhaven National Laboratory (BNL) as part of an audit and
sensitivity analyses for the NRC [Nuclear Regulatory Commission]
Staff. Based on the small value of the increase compared to the
range of uncertainty in the CDF, the increase is considered
acceptable. The analysis performed by the WOG and presented in the
WCAP included changes to the surveillance frequencies for the
automatic actuation logic and actuation relays and the reactor trip
and bypass breakers. The overall increase in the CDF, including the
changes to the surveillance frequencies for the automatic actuation
logic and actuation relays and the reactor trip and bypass breakers,
was approximately 6 percent. However, even with this increase, the
overall CDF remains lower than the NRC safety goal of
10-4/reactor year.
Changes to surveillance test frequencies for the RTS and ESFAS
interlocks do not represent a significant reduction in testing. The
currently specified test interval for interlock channels allows the
surveillance requirement to be satisfied by verifying that the
permissive logic is in its required state using the annunciator
status light. The surveillance as currently required only verifies
the status of the permissive logic and does not address verification
of channel setpoint or operability. The setpoint verification and
channel operability is verified after a refueling shutdown. The
definition of the channel check includes comparison of the channel
status with other channels for the same parameter. The requirement
to routinely verify permissive status is a different consideration
than the availability of trip or actuation channels which are
required to change state on the occurrence of an event and for which
the function availability is more dependent on the surveillance
interval. Therefore, the change in the interlock surveillance
requirement to at least once every 18 months does not represent a
significant change in channel surveillance and does not involve a
significant increase in unavailability of the RTS and ESFAS.
For the additional relaxations in WCAP-14333, the WOG evaluated
the impact of the additional relaxation of allowed outage times and
completion times, and action statements on core damage frequency.
The change in core damage frequency is 3.1 percent for those plants
with two out of three logic schemes that have not implemented the
proposed surveillance test interval, allowed outage times, and
completion times evaluated in WCAP-10271 and its supplements. This
analysis calculates a significantly lower increase in core damage
frequency than the WCAP-10271 analysis calculated. This can be
attributed to more realistic maintenance intervals used in the
current analysis and crediting the AMSAC [ATWS (anticipated
transient without scram) mitigating system actuation circuitry]
system as an alternative method of initiating the auxiliary
feedwater pumps. Therefore, the overall increase in CDF is estimated
to be 3.1% for the proposed changes per the generic Westinghouse
analysis.
The NRC performed an independent evaluation of the impact on
core damage frequency (CDF) and large early release fraction (LERF).
The results of the staff's review indicate that the increase in core
damage frequency is small (approximately 3.2%) and the large early
release fraction would increase by only 4 percent for 2 out of 3
logic schemes that have not implemented the proposed surveillance
test interval, allowed outage times, and completion times evaluated
in WCAP-10271 and its supplements. Further, the absolute values for
CDF still remain within NRC safety goals.
Therefore, the proposed changes do not result in a significant
increase in the severity or consequences of an accident previously
evaluated. Implementation of the proposed changes affects the
probability of failure of the RTS and ESFAS but does not alter the
manner in which protection is afforded or the manner in which
limiting criteria are established.
Criterion 2--The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes do not result in a change in the manner in
which the RTS or ESFAS provide plant protection. No change is being
made which alters the functioning of the RTS or ESFAS (other than in
a test mode). Rather the likelihood or probability of the RTS or
ESFAS functioning properly is affected as described above.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident as defined in the Safety Analysis
Report.
The proposed changes do not involve hardware changes. Some
existing instrumentation is designed to be tested in bypass and
current Technical Specifications allow testing in bypass. Testing in
bypass is also recognized by IEEE [Institute of Electrical and
Electronics Engineers] Standards. Therefore, testing in bypass has
been previously approved and implementation of the proposed changes
for testing in bypass does not create the possibility of a new or
different kind of accident from any previously evaluated.
Furthermore since the other proposed changes do not alter the
physical operation or functioning of the RTS or ESFAS the
possibility of a new or different kind of accident from any
previously evaluated has not been created.
Criterion 3--The proposed license amendment does not involve a
significant reduction in a margin of safety.
The proposed changes do not alter the safety limits, limiting
safety system setpoints or limiting conditions for operation. The
RTS and ESFAS analog instrumentation remain operable to mitigate as
assumed in the accident analysis. The impact of reduced testing
other than as addressed above is to allow a longer time interval
over which instrument uncertainties (e.g., drift) may act.
Implementation of the proposed changes is expected to result in
an overall improvement in safety by less frequent testing of the RTS
and ESFAS analog instruments will result in less inadvertent reactor
trips and actuation of Engineered Safety Features components.
This analysis demonstrates that the proposed amendment to The
North Anna Unit 1 and 2 Technical Specifications does not involve a
significant increase in the probability or consequences of a
previously evaluated accident, does not create the possibility of a
new or different kind of accident and does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
[[Page 32293]]
NRC Section Chief: Richard L. Emch Jr.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois and Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: March 22, 1999.
Brief description of amendments: The amendments modify the
technical specifications to permit the use of the Gamma-Metrics Post
Accident Neutron Monitors source range neutron flux detectors in
addition to the Westinghouse source range neutron flux monitors to
satisfy the requirement that two source range neutron flux monitors be
operable during Mode 6 operations (refueling).
Date of issuance: June 2, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 109 & 109, 102 & 102.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 29, 1999 (64 FR
14944). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: December 2, 1996, as
supplemented on May 27, 1999.
Brief description of amendments: The amendments revised Technical
Specification 3/4.4.2 to reduce the number of required Safety/Relief
valves (SRVs). This change supports a modification to remove five of
the currently installed SRVs due to excess capacity and to reduce the
amount of valve maintenance and associated worker radiation dose. The
revised TS requires that 12 of the remaining installed 13 SRVs be
operable.
Date of issuance: June 3, 1999.
Effective date: Immediately, to be implemented prior to startup of
L1C10 for Unit 1 and prior to startup of L2C9 for Unit 2.
Amendment Nos.: 133 & 118.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4343). The May 27, 1999, submittal provided additional clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: March 23, 1999 (NRC-99-0025).
Brief description of amendment: The amendment revises Technical
Specification Surveillance Requirement (SR) 4.4.1.1.1 to require each
recirculation pump discharge valve be demonstrated operable at least
once every 18 months, deletes the ``*'' footnote from the SR, and
revises the footnote itself to read ``Not used.''
Date of issuance: May 25, 1999.
Effective date: May 25, 1999, with full implementation within 90
days.
Amendment No.: 133.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 21, 1999 (64 FR
19555)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 25, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of application for amendments: July 9, 1998, as supplemented
March 31, 1999.
Brief description of amendments: These amendments revised Technical
Specification (TS) 3/4.7.1.1 and associated Bases for both units. This
amendment specifies maximum allowable reactor power level based on the
number of operable main steam safety valves (MSSVs) rather than
requiring reduction in reactor trip setpoint. This change is consistent
with the Nuclear Regulatory Commission's improved Standard Technical
Specifications for Westinghouse plants (NUREG-1431, Revision 1). The
maximum allowable reactor power level with inoperable MSSVs will be
calculated based on the recommendations of Westinghouse Nuclear Safety
Advisory Letter 94-01. The change to the Unit 1 TS 3.7.1.1 also deletes
reference to 2 loop operation since 2 loop operation is not a licensed
[[Page 32294]]
condition for either unit. Unit 1 TS Table 3.7-3 is then renumbered to
be Table 3.7-2.
The March, 31, 1999 letter withdrew a portion of the amendment
which would have removed the values of the orifice diameter of each
MSSV from the TSs. This information will be maintained in the TSs.
Date of issuance: June 3, 1999.
Effective date: Units 1 and 2 as of date of issuance and shall be
implemented within 60 days.
Amendment Nos.: 223 and 99.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43203). The March 31, 1999 letter did not change the initial proposed
no significant hazards consideration determination or expand the
amendment beyond the scope of the initial notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: August 31, 1998.
Brief description of amendment: Changes the Crystal River Unit 3
Technical Specifications to add additional instrumentation variables to
Improved Technical Specification Table 3.3.17-1, Post-Accident
Monitoring Instrumentation.
Date of issuance: June 3, 1999.
Effective date: As of date of issuance, to be implemented prior to
commencing cycle 12 operation.
Amendment No.: 177.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56250).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: November 23, 1998, as
supplemented January 29 and May 7, 1999.
Brief description of amendment: The amendment changes the Improved
Technical Specifications for several reactor protection system and
engineered safeguards actuation system setpoint values, and changes the
surveillance requirement to verify valve position for valves in the
high pressure injection system flowpath.
Date of issuance: May 21, 1999.
Effective date: As of date of issuance, to be implemented prior to
commencing Cycle 12 operation.
Amendment No.: 178.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71966). The supplemental letters dated January 29 and May 7, 1999,
did not change the original proposed no significant hazards
consideration determination, or expand the scope of the amendment
request as originally noticed.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 21, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
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: December 16, 1998.
Description of amendment request: These amendments consist of
changes to the Technical Specifications (TS) in response to Florida
Power & Light's (FPL) application dated December 16, 1998, regarding
facility staff qualifications for multi-discipline supervisor (MDS)
positions at Lucie Units 1 and 2. The amendments revise the
administrative controls in TS Section 6.3, ``Unit Staff
Qualifications,'' by modifying FPL's commitment to ANSI/ANS 3.1-1978,
``Selection and Training of Nuclear Power Plant Personnel,'' to
incorporate specific staff qualifications for the position of MDS.
Date of Issuance: May 25, 1999.
Effective Date: May 25, 1999.
Amendment Nos.: 161 and 102.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the TS.
Date of Initial Notice in Federal Register: February 10, 1999 (64
FR 6698).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 25, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: November 5, 1998, as
supplemented February 18, 1999.
Brief description of amendment: The amendment modifies the safety
limits and surveillances of the LPRM and APRM systems and related Bases
pages to ensure the APRM channels respond within the necessary range
and accuracy and to verify channel operability. In addition, an
unrelated change to the Bases of Specification 2.3 is included to
clarify some ambiguous language.
Date of Issuance: June 2, 1999.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 208.
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69342). The February 18, 1999, supplemental letter provided
clarifying information, was within the scope of the original
application, and did not change the staff's original no significant
hazards consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated June 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-336,
and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New
London County, Connecticut
Date of application for amendment: December 22, 1998, as
supplemented March 19, 1999.
Brief description of amendment: The amendment replaces specific
titles in Section 6.0 of the Technical
[[Page 32295]]
Specifications of all three Millstone units with generic titles.
Date of issuance: June 3, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 105, 235, and 171.
Facility Operating License Nos. DPR-21, DPR-65, and NPF-49:
Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4158). The March 19, 1999 letter provided clarifying information that
did not change the scope of the December 22, 1998, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: April 20, 1999.
Brief description of amendments: The amendments revised the
implementation date for the relocation of the requirements specified in
Technical Specification Sections 3.1.E and 5.1 to the Updated Final
Safety Analyis Report. On December 7, 1998, the NRC had previously
issued license amendments 141 and 132 for Units 1 and 2, respectively,
approving the relocation of aforementioned requirements by June 1,
1999. The proposed amendments would postpone the implementation date to
September 1, 1999.
Date of issuance: June 2, 1999.
Effective date: June 2, 1999, with full implementation within 30
days .
Amendment Nos.: 145 and 136.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 29, 1999 (64 FR
23131) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: January 4, 1999.
Brief description of amendments: These amendments revise the
administrative section of the Technical Specification pertaining to
controlled access to high radiation areas, and the reporting dates for
the annual occupational radiation exposure report and the annual
radioactive effluent release report.
Date of issuance: May 24, 1999.
Effective date: Units 1 and 2, as of date of issuance and shall be
implemented within 30 days.
Amendment Nos.: 135 and 100.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6706) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 24, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County , New York
Date of application for amendment: January 25, 1999.
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) by relocating certain requirements from the TSs to
the Final Safety Analysis Report.
Date of issuance: May 24, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 189.
Facility Operating License No. DPR-64: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 21, 1999 (64 FR
19562).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 24, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit
1, Luzerne County, Pennsylvania
Date of application for amendment: March 12, 1999.
Brief description of amendment: This amendment would change the
allowable values for both the core spray system and the low pressure
coolant injection system reactor steam dome pressure-low functions.
Date of issuance: May 25, 1999.
Effective date: As of date of issuance, and shall be implemented
within 30 days after startup from the Unit 1 eleventh refueling and
inspection outage currently scheduled for spring 2000.
Amendment No.: 181.
Facility Operating License No. NPF-14: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17028).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 25, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: October 27, 1998, as
supplemented by letters in 1999 dated January 11, January 29, February
25, and April 7 (two letters), and May 17.
Brief description of amendment: The amendment revised Technical
Specification 4.4.5.4, Table 4.4-3 and the associated Bases to allow
the repair of the steam generator tubes with the Electrosleeve tube
repair method.
Date of issuance: May 21, 1999.
Effective date: May 21, 1999, to be implemented within 30 days from
the date of issuance. The amendment includes a two cycle operating
limit that requires all steam generator tubes repaired with
Electrosleeves to be removed from service at the end of two operating
cycles following installation of the first Electrosleeve in the steam
generators.
Amendment No.: 132.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66604). The supplemental letters in 1999 dated January 11, January 29,
February 25, and April 7 (two letters)
[[Page 32296]]
provided additional clarifying information that did not expand the
staff's original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 21, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Elmer Ellis Library,
University of Missouri, Columbia Missouri 65201.
Dated at Rockville, Maryland, this 9th day of June 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-15098 Filed 6-15-99; 8:45 am]
BILLING CODE 7590-01-P
