[Federal Register Volume 68, Number 32 (Tuesday, February 18, 2003)]
[Notices]
[Pages 7806-7810]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-3835]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. (as shown in Attachment 1), License Nos. (as shown in
Attachment 1) EA-03-009]
In the Matter of: All Pressurized Water Reactor Licensees; Order
Modifying Licenses (Effective Immediately)
I
The Licensees identified in the Attachment to this Order hold
licenses issued by the U.S. Nuclear Regulatory Commission (NRC or
Commission) authorizing operation of pressurized water reactor (PWR)
nuclear power plants in accordance with the Atomic Energy Act of 1954
and 10 CFR part 50.
II
The reactor pressure vessel (RPV) heads of PWRs have penetrations
for control rod drive mechanisms and instrumentation systems. Nickel-
based alloys (e.g., Alloy 600) are used in the penetration nozzles and
related welds. Primary coolant water and the operating conditions of
PWR plants can cause cracking of these nickel-based alloys through a
process called primary water stress corrosion cracking (PWSCC). The
susceptibility of RPV head penetrations to PWSCC appears to be strongly
linked to the operating time and temperature of the RPV head. Problems
related to PWSCC have therefore increased as plants have operated for
longer periods of time. Inspections of the RPV head nozzles at the
Oconee Nuclear Station, Units 2 and 3 (Oconee), in early 2001
identified circumferential cracking of the nozzles above the J-groove
weld, which joins the nozzle to the RPV head. Circumferential cracking
above the J-groove weld is a safety concern because of the possibility
of a nozzle ejection if the circumferential cracking is not detected
and repaired.
Section XI of the American Society of Mechanical Engineers Boiler
and Pressure Vessel Code (ASME Code), which is incorporated into NRC
regulations by 10 CFR 50.55a, ``Codes and standards,'' currently
specifies that inspections of the RPV head need only include a visual
check for leakage on the insulated surface or surrounding area. These
inspections may not detect small amounts of leakage from an RPV head
penetration with cracks extending through the nozzle or the J-groove
weld. Such leakage can create an environment that leads to
circumferential cracks in RPV head penetration nozzles or corrosion of
the RPV head. In response to the inspection findings at Oconee and
because existing requirements in the ASME Code and NRC regulations do
not adequately address inspections of RPV head penetrations for
degradation due to PWSCC, the NRC issued Bulletin 2001-01,
``Circumferential Cracking of Reactor Pressure Vessel Head Penetration
Nozzles,'' dated August 3, 2001. In response to the Bulletin, PWR
Licensees provided their plans for inspecting RPV head penetrations and
the outside surface of the heads to determine whether any nozzles were
leaking.
In early March 2002, while conducting inspections of reactor vessel
head penetrations prompted by Bulletin 2001-01, the Licensee for the
Davis-Besse Nuclear Power Station (Davis-Besse) identified a cavity in
the reactor vessel head near the top of the dome. The cavity was next
to a leaking nozzle
[[Page 7807]]
with a through-wall axial crack and was in an area of the reactor
vessel head that the Licensee had left covered with boric acid deposits
for several years. On March 18, 2002, the NRC issued Bulletin 2002-01,
``Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure
Boundary Integrity,'' which requested PWR Licensees to provide
information on their reactor vessel head inspection and maintenance
programs, the material condition of their reactor vessel heads, and
their boric acid inspection programs. In their responses, the Licensees
provided information about their boric acid inspection programs and
their inspections and assessments to ensure that their respective plant
did not have reactor vessel head degradation like that identified at
Davis-Besse.
The experience at Davis-Besse and the discovery of leaks and nozzle
cracking at other plants reinforced the need for more effective
inspections of RPV head penetration nozzles. The absence of an
effective inspection regime could, over time, result in unacceptable
circumferential cracks in RPV head penetration nozzles or in the
degradation of the RPV head by corrosion. These degradation mechanisms
increase the probability of a more significant loss of reactor coolant
pressure boundary through ejection of a nozzle or other rupture of the
RPV head. The NRC issued Bulletin 2002-02, ``Reactor Pressure Vessel
Head and Vessel Head Penetration Nozzle Inspection Programs,'' dated
August 9, 2002, requesting that Licensees provide information about
their inspection programs and any plans to supplement existing visual
inspections with additional measures (e.g., volumetric and surface
examinations). Licensees have responded to Bulletin 2002-02 with
descriptions of their inspection plans for at least the first refueling
outage following the issuance of Bulletin 2002-02 or with a schedule to
submit such descriptions before the next refueling outage. Many of the
Licensees' responses to Bulletin 2002-02 did not describe long-term
inspection plans. Instead the Licensees stated that they would follow
guidance being developed by the industry-sponsored Materials
Reliability Program.
Inspections performed at several PWR plants in late 2002 found
leakage and cracks in nozzles or J-groove welds that have required
repairs or prompted the replacement of the RPV head. In addition, as
discussed in NRC Information Notice 2003-02, ``Recent Experience with
Reactor Coolant System Leakage and Boric Acid Corrosion,'' issued
January 16, 2003, leakage has recently occurred at some plants from
connections above the RPV head and has required additional assessments
and inspections to ensure that the leakage has not caused significant
degradation of RPV heads.
III
Based on recent experience, current inspection requirements in the
ASME Code and related NRC regulations do not provide adequate assurance
that reactor coolant pressure boundary integrity will be maintained for
all combinations of construction materials, operating conditions, and
operating histories at PWRs. The long-term resolution of RPV head
penetration inspection requirements is expected to involve changes to
the ASME Code and NRC regulations, specifically 10 CFR 50.55a. Research
being conducted by the NRC and industry is increasing our understanding
of material performance, improving inspection capabilities, and
supporting assessments of the risks to public health and safety
associated with potential degradation of the RPV head and associated
penetration nozzles. These research activities are important to the
long term development of revisions to the ASME Code and NRC
regulations.
The operating history of PWRs supports a general correlation among
certain operating parameters, including the length of time plants have
been in operation, and the likelihood of occurrence of PWSCC of nickel-
based alloys used in RPV head penetration nozzles. Bulletin 2002-02
presented a three-tier categorization of susceptibility to RPV head
penetration nozzle degradation based on reactor operating durations and
temperatures. Licensees' responses to the Bulletin included an estimate
of the effective degradation years (EDY) and the appropriate
categorization of each plant into one of the three susceptibility
categories. Each Licensee proposed an inspection plan for RPV head
penetrations based upon the susceptibility to degradation via PWSCC (as
represented by the value of EDY calculated for the facility). In
addition, recent operating experience has shown that, under certain
conditions, leakage from mechanical and welded connections above the
RPV head can lead to the degradation of the low alloy steel head by
boric acid corrosion.
Revising the ASME Code and subsequently the NRC regulations will
take several years. The Licensees' actions to date in response to the
NRC bulletins have provided reasonable assurance of adequate protection
of public health and safety for the near term operating cycles, but
cannot be relied upon to do so for the entire interim period until NRC
regulations are revised. Additional periodic inspections of RPV heads
and associated penetration nozzles at PWRs, as a function of the unit's
susceptibility to PWSCC and as appropriate to address the discovery of
boron deposits, are necessary to provide reasonable assurance that
plant operations do not pose an undue risk to the public health and
safety. Consequently, it is necessary to establish a minimum set of RPV
head inspection requirements, as a supplement to existing inspection
and other requirements in the ASME Code and NRC regulations, through
the issuance of an Order to PWR Licensees.
It is appropriate and necessary to the protection of public health
and safety to establish a clear regulatory framework, pending the
development of consensus standards and incorporation of revised
inspection requirements into 10 CFR 50.55a, directly or through
reference to a future version of the ASME Code. In order to provide
reasonable assurance of adequate protection of public health and safety
for the interim period, all PWR Licenses identified in the Attachment
to this Order shall be modified to include the inspection requirements
for RPV heads and associated penetration nozzles identified in Section
IV of this Order. The NRC requirements imposed by this Order are based
on the body of evidence available through February 2003. Continuing
research and operating experience may support future changes to the
requirements imposed through this Order. In addition, pursuant to 10
CFR 2.202, I find that in the circumstances described above, the public
health, safety, and interest require that this Order be immediately
effective.
IV
Accordingly, pursuant to sections 103, 104b, 161b, 161i, 161o, 182,
and 186 of the Atomic Energy Act of 1954, as amended, and the
Commission's regulations in 10 CFR 2.202 and 10 CFR part 50, it is
hereby ordered, effective immediately, that all licenses identified in
the attachment to this order are modified as follows:
A. To determine the required inspection(s) for each refueling
outage at their facility, all Licensees shall calculate the
susceptibility category of each reactor vessel head to PWSCC-related
degradation, as represented by a value of EDY for the end of each
operating cycle, using the following equation:
[[Page 7808]]
[GRAPHIC] [TIFF OMITTED] TN18FE03.013
Where:
EDY = total effective degradation years, normalized to a reference
temperature of 600 [deg]F
[Delta]EFPYj = operating time in years at Thead,j
Qi = activation energy for crack initiation (50 kcal/mole)
R = universal gas constant (1.103x10-3 kcal/mole[deg]R)
Thead,j = 100% power head temperature during time period j
([deg]R = [deg]F + 459.67)
Tref = reference temperature (600 [deg]F = 1059.67 [deg]R)
n = number of different head temperatures during plant history
This calculation shall be performed with best estimate values for
each parameter at the end of each operating cycle for the RPV head that
will be in service during the subsequent operating cycle. The
calculated value of EDY shall determine the susceptibility category and
the appropriate inspection for the RPV head during each refueling
outage.
B. All Licensees shall use the following criteria to assign the RPV
head at their facility to the appropriate PWSCC susceptibility
category:
High--(1) Plants with a calculated value of EDY greater than 12, OR (2)
Plants with an RPV head that has experienced cracking in a penetration
nozzle or J-groove weld due to PWSCC.
Moderate--Plants with a calculated value of EDY less than or equal to
12 and greater than or equal to 8 AND no previous inspection findings
requiring classification as High.
Low--Plants with a calculated value of EDY less than 8 AND no previous
inspection findings requiring classification as High.
C. All Licensees shall perform inspections of the RPV head \1\
using the following techniques and frequencies.\2\
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\1\ This Order imposes additional inspection requirements.
Licensees are required to address any findings from these
inspections (i.e., perform analyses and repairs) in accordance with
existing requirements in the ASME Code and 10 CFR 50.55a. The NRC
has issued guidance to address flaw evaluations for RPV head
penetration nozzles (see letter dated November 21, 2001, from J.
Strosnider, NRC, to A. Marion, Nuclear Energy Institute) and will,
as necessary, issue revised guidance pending the updating of the
ASME Code and related NRC regulations.
\2\ The requirements of this Order are generally consistent with
inspection plans that the NRC staff accepted in letters to some
Licensees regarding their responses to Bulletin 2002-02. If the NRC
staff has already accepted a specific variation from the
requirements of this Order (e.g., inspections to less than two (2)
inches above the J-groove weld), the Licensee may continue with the
previously accepted inspection plan for the next refueling outage
after issuance of this Order, provided that in its response to this
Order the Licensee identifies all discrepancies between the
requirements of this Order and the previously accepted inspection
plan. Licensees proposing to deviate from the requirements of this
Order for subsequent refueling outages shall seek relaxation of this
Order pursuant to the procedure specified at the end of this
Section.
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(1) For those plants in the High category, RPV head and head
penetration nozzle inspections shall be performed using the following
techniques every refueling outage.\3\
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\3\ For repaired RPV head penetration nozzles that establish a
new pressure boundary, the ultrasonic testing inspection shall
include the weld and at least one (1) inch above the weld in the
nozzle base material. For RPV head penetration nozzles or J-groove
welds repaired using a weld overlay, the overlay shall be examined
by either ultrasonic, eddy current, or dye penetrant testing in
addition to the examinations required by (1)(b)(i) or (1)(b)(ii).
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(a) Bare metal visual examination of 100% of the RPV head surface
(including 360[deg] around each RPV head penetration nozzle), AND
(b) Either:
(i) Ultrasonic testing of each RPV head penetration nozzle (i.e.,
nozzle base material) from two (2) inches above the J-groove weld to
the bottom of the nozzle and an assessment to determine if leakage has
occurred into the interference fit zone, OR
(ii) Eddy current testing or dye penetrant testing of the wetted
surface of each J-Groove weld and RPV head penetration nozzle base
material to at least two (2) inches above the J-groove weld.
(2) For those plants in the Moderate category, RPV head and head
penetration inspections shall be performed such that at least the
requirements of 2(a) or 2(b) are performed each refueling outage. In
addition the requirements of 2(a) and 2(b) shall each be performed at
least once over the course of every two (2) refueling outages.
(a) Bare metal visual examination of 100% of the RPV head surface
(including 360[deg] around each RPV head penetration nozzle).
(b) Either:
(i) Ultrasonic testing of each RPV head penetration nozzle (i.e.,
nozzle base material) from two (2) inches above the J-groove weld to
the bottom of the nozzle and an assessment to determine if leakage has
occurred into the interference fit zone, OR
(ii) Eddy current testing or dye penetrant testing of the wetted
surface of each J-Groove weld and RPV head penetration nozzle base
material to at least two (2) inches above the J-groove weld.
(3) For those plants in the Low category, RPV head and head
penetration nozzle inspections shall be performed as follows. An
inspection meeting the requirements of 3(a) must be completed at least
every third refueling outage or every five (5) years, whichever occurs
first. If an inspection meeting the requirements of 3(a) was not
performed during the refueling outage immediately preceding the
issuance of this Order, the Licensee must complete an inspection
meeting the requirements of 3(a) within the first two (2) refueling
outages following issuance of this Order. The requirements of 3(b) must
be completed at least once over the course of five (5) years after the
issuance of this Order and thereafter at least every four (4) refueling
outages or every seven (7) years, whichever occurs first.
(a) Bare metal visual examination of 100% of the RPV head surface
(including 360[deg] around each RPV head penetration nozzle).
(b) Either:
(i) Ultrasonic testing of each RPV head penetration nozzle (i.e.,
nozzle base material) from two (2) inches above the J-groove weld to
the bottom of the nozzle and an assessment to determine if leakage has
occurred into the interference fit zone, or
(ii) Eddy current testing or dye penetrant testing of the wetted
surface of each J-Groove weld and RPV head penetration nozzle base
material to at least two (2) inches above the J-groove weld.
D. During each refueling outage, visual inspections shall be
performed to identify potential boric acid leaks from pressure-
retaining components above the RPV head. For any plant with boron
deposits on the surface of the RPV head or related insulation,
discovered either during the inspections required by this Order or
otherwise and regardless of the source of the deposit, before returning
the plant to operation the Licensee shall perform inspections of the
affected RPV head surface and penetrations appropriate to the
conditions found to verify the integrity of the affected area and
penetrations.
[[Page 7809]]
E. For each inspection required in Paragraph C, the Licensee shall
submit a report detailing the inspection results within sixty (60) days
after returning the plant to operation.\4\ For each inspection required
in Paragraph D, the Licensee shall submit a report detailing the
inspection results within sixty (60) days after returning the plant to
operation if a leak or boron deposit was found during the inspection.
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\4\ This reporting requirement supercedes the 30-day reports
requested by NRC Bulletin 2002-02.
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F. In the response required by Section V of this Order, all
Licensees shall notify the Commission if: (1) They are unable to comply
with any of the requirements of Section IV, or (2) compliance with any
of the requirements of Section IV is unnecessary. Licensees proposing
to deviate from the requirements of this Order shall seek relaxation of
this Order pursuant to the procedure specified below.
The Director, Office of Nuclear Reactor Regulation, may, in
writing, relax or rescind any of the above conditions upon
demonstration by the Licensee of good cause. A request for relaxation
regarding inspection of specific nozzles shall also address the
following criteria:
(1) The proposed alternative(s) for inspection of specific nozzles
will provide an acceptable level of quality and safety, or
(2) Compliance with this Order for specific nozzles would result in
hardship or unusual difficulty without a compensating increase in the
level of quality and safety.
Requests for relaxation associated with specific penetration
nozzles will be evaluated by the NRC staff using its procedure for
evaluating proposed alternatives to the ASME Code in accordance with 10
CFR 50.55a(a)(3).
V
In accordance with 10 CFR 2.202, the Licensee must, and any other
person adversely affected by this Order may, submit an answer to this
Order, and may request a hearing on this Order, within twenty (20) days
of the date of this Order. Where good cause is shown, consideration
will be given to extending the time to request a hearing. A request for
extension of time in which to submit an answer or request a hearing
must be made in writing to the Director, Office of Nuclear Reactor
Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555,
and include a statement of good cause for the extension. The answer may
consent to this Order. Unless the answer consents to this Order, the
answer shall, in writing and under oath or affirmation, specifically
set forth the matters of fact and law on which the Licensee or other
person adversely affected relies and the reasons as to why the Order
should not have been issued. Any answer or request for a hearing shall
be submitted to the Secretary, Office of the Secretary of the
Commission, U.S. Nuclear Regulatory Commission, ATTN: Rulemakings and
Adjudications Staff, Washington, DC 20555. Copies shall also be sent to
the Director, Office of Nuclear Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington, DC 20555; to the Assistant General
Counsel for Materials Litigation and Enforcement at the same address;
to the Regional Administrator for NRC Region I, II, III, or IV, as
appropriate for the specific plant; and to the Licensee if the answer
or hearing request is by a person other than the Licensee. Because of
possible disruptions in delivery of mail to United States Government
offices, it is requested that answers and requests for hearing be
transmitted to the Secretary of the Commission either by means of
facsimile transmission to 301-415-1101 or by e-mail to
[email protected] and also to the Assistant General Counsel for
Materials Litigation and Enforcement either by means of facsimile
transmission to 301-415-3725 or by e-mail to [email protected]. If
a person other than the Licensee requests a hearing, that person shall
set forth with particularity the manner in which his interest is
adversely affected by this Order and shall address the criteria set
forth in 10 CFR 2.714(d).\5\
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\5\ The version of Title 10 of the Code of Federal Regulations,
published January 1, 2002, inadvertently omitted the last sentence
of 10 CFR 2.714 (d) and paragraphs (d)(1) and (d)(2) regarding
petitions to intervene and contentions. For the complete, corrected
text of 10 CFR 2.714 (d), please see 67 FR 20884, April 29, 2002.
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If a hearing is requested by the Licensee or a person whose
interest is adversely affected, the Commission will issue an Order
designating the time and place of any hearing. If a hearing is held,
the issue to be considered at such hearing shall be whether this Order
should be sustained.
Pursuant to 10 CFR 2.202(c)(2)(i), the Licensee may, in addition to
demanding a hearing at the time the answer is filed or sooner, move the
presiding officer to set aside the immediate effectiveness of the Order
on the ground that the Order, including the need for immediate
effectiveness, is not based on adequate evidence but on mere suspicion,
unfounded allegations, or error.
In the absence of any request for hearing, or written approval of
an extension of time in which to request a hearing, the provisions
specified in Section IV above shall be final twenty (20) days from the
date of this Order without further order or proceedings. If an
extension of time for requesting a hearing has been approved, the
provisions specified in Section IV shall be final when the extension
expires if a hearing request has not been received. An answer or
request for hearing shall not stay the immediate effectiveness of this
order.
Dated this 11th day of February, 2003.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
Attachment to Order:
Facilities
Beaver Valley Power Station, Units 1 and 2
Docket Nos. 50-334 and 50-412
License Nos. DPR-66 and NPF-73
Calvert Cliffs Nuclear Power Plant,
Units 1 and 2
Docket Nos. 50-317 and 50-318
License Nos. DPR-53 and DPR-69
R.E. Ginna Nuclear Power Plant
Docket No. 50-244
License No. DPR-18
Indian Point Nuclear Generating Station,
Units 2 and 3
Docket Nos. 50-247 and 50-286
License Nos. DPR-26 and DPR-64
Millstone Power Station, Units 2 and 3
Docket Nos. 50-336 and 50-423
License Nos. DPR-65 and NPF-49
Salem Nuclear Generating Station,
Units 1 and 2
Docket Nos. 50-272 and 50-311
License Nos. DPR-70 and DPR-75
Seabrook Station, Unit 1
Docket No. 50-443
License No. NPF-86
Three Mile Island Nuclear Station, Unit 1
Docket No. 50-289
License No. DPR-50
Catawba Nuclear Station, Units 1 and 2
Docket Nos. 50-413 and 50-414
License Nos. NPF-35 and NPF-52
Crystal River Nuclear Power Plant
Docket No. 50-302
License No. DPR-72
Joseph M. Farley Nuclear Plant,
Units 1 and 2
Docket Nos. 50-348 and 50-364
License Nos. NPF-2 and NPF-8
Shearon Harris Nuclear Power Plant, Unit 1
Docket No. 50-400
License No. NPF-63
William B. McGuire Nuclear Station,
Units 1 and 2
Docket Nos. 50-369 and 50-370
License Nos. NPF-9 and NPF-17
North Anna Power Station, Units 1 and 2
Docket Nos. 50-338 and 50-339
License Nos. NPF-4 and NPF-7
[[Page 7810]]
Surry Power Station, Units 1 and 2
Docket Nos. 50-280 and 50-281
License Nos. DPR-32 and DPR-37
Oconee Nuclear Station, Units 1, 2 and 3
Docket Nos. 50-269, 50-270 and 50-287
License Nos. DPR-38, DPR-47 and DPR-55
H.B. Robinson Steam Electric Plant, Unit 2
Docket No. 50-261
License No. DPR-23
St. Lucie Nuclear Plant, Units 1 and 2
Docket Nos. 50-335 and 50-389
License Nos. DPR-67 and NPF-16
Turkey Point Nuclear Generating Station,
Units 3 and 4
Docket Nos. 50-250 and 50-251
License Nos. DPR-31 and DPR-41
Sequoyah Nuclear Plant, Units 1 and 2
Docket Nos. 50-327 and 50-328
License Nos. DPR-77 and DPR-79
Watts Bar Nuclear Plant, Unit 1
Docket No. 50-390
License No. NPF-90
Virgil C. Summer Nuclear Station, Unit 1
Docket No. 50-395
License No. NPF-12
Vogtle Electric Generating Plant,
Units 1 and 2
Docket Nos. 50-424 and 50-425
License Nos. NPF-68 and NPF-81
Braidwood Station, Units 1 and 2
Docket Nos. STN 50-456 and STN 50-457
License Nos. NPF-72 and NPF-77
Byron Station, Units 1 and 2
Docket Nos. STN 50-454 and STN 50-455
License Nos. NPF-37 and NPF-66
Donald C. Cook Nuclear Plant, Units 1 and 2
Docket Nos. 50-315 and 50-316
License Nos. DPR-58 and DPR-74
Davis-Besse Nuclear Power Station, Unit 1
Docket No. 50-346
License No. NPF-3
Kewaunee Nuclear Power Plant
Docket No. 50-305
License No. DPR-43
Palisades Plant
Docket No. 50-255
License No. DPR-20
Point Beach Nuclear Plant, Units 1 and 2
Docket Nos. 50-266 and 50-301
License Nos. DPR-24 and DPR-27
Prairie Island Nuclear Generating Plant, Units 1 and 2
Docket Nos. 50-282 and 50-306
License Nos. DPR-42 and DPR-60
Arkansas Nuclear One, Units 1 and 2
Docket Nos. 50-313 and 50-368
License Nos. DPR-51 and NPF-6
Callaway Plant, Unit 1
Docket No. 50-483
License No. NPF-30
Comanche Peak Steam Electric Station,
Units 1 and 2
Docket Nos. 50-445 and 50-446
License Nos. NPF-87 and NPF-89
Diablo Canyon Nuclear Power Plant,
Units 1 and 2
Docket Nos. 50-275 and 50-323
License Nos. DPR-80 and DPR-82
Fort Calhoun Station, Unit 1
Docket No. 50-285
License No. DPR-40
Palo Verde Nuclear Generating Station,
Units 1, 2 and 3
Docket Nos. STN 50-528, STN 50-529 and
STN 50-530
License Nos. NPF-41, NPF-51 and NPF-74
San Onofre Nuclear Station, Units 2 and 3
Docket Nos. 50-361 and 50-362
License Nos. NPF-10 and NPF-15
South Texas Project Electric Generating Station, Units 1 and 2
Docket Nos. 50-498 and 50-499
License Nos. NPF-76 and NPF-80
Waterford Steam Electric Generating Station, Unit 3
Docket No. 50-382
License No. NPF-38
Wolf Creek Generating Station, Unit 1
Docket No. 50-482
License No. NPF-42
[FR Doc. 03-3835 Filed 2-14-03; 8:45 am]
BILLING CODE 7590-01-P