[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]


=======================================================================
-----------------------------------------------------------------------

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\
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    (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\
---------------------------------------------------------------------------

    \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).
---------------------------------------------------------------------------

    (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.
---------------------------------------------------------------------------

    \4\ This reporting requirement supercedes the 30-day reports 
requested by NRC Bulletin 2002-02.
---------------------------------------------------------------------------

    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\
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    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