LIC-05-0043, Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors

Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors

ML051050581
Person / Time
Site:	Fort Calhoun
Issue date:	04/15/2005
From:	Phelps R Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EA-03-009, LIC-05-0043
Download: ML051050581 (21)

Text

Omaha Public Power District 444 South 16th Street Mall Omaha NE 68102-2247 April 15,2005 LIC-05-0043 U.S. Nuclear Regulatory Commission ATTN.: Document Control Desk Washington, D.C. 20555-0001

References:

1. Docket No. 50-285

2. Letter from Ralph L. Phelps (OPPD) to Document Control Desk (NRC) dated April 7, 2005, Fort Calhoun Station Unit No. 1, Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors (LIC-05-0040)

3. Letter from R. William Borchardt (NRC) to Ross Ridenoure (OPPD) dated February 20, 2004, Issuance of First Revised NRC Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors (NRC-04-0022)

(ML040220181)

SUBJECT:

Fort Calhoun Station Unit No. 1, Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors In Reference 2, the Omaha Public Power District (OPPD) provided information in support of a relaxation request with respect to Reference 3. NRC personnel in an April 4, 2005 phone conversation asked that the information in Reference 2 be provided in advance of forthcoming analysis information to facilitate their review. In Reference 2, OPPD specified plans to submit additional analysis information from Westinghouse and Dominion Engineering, and a status report of nozzle examinations completed to date, as a supplement to support review and approval of Reference 2. This letter forwards the required analysis information, consisting of the hoop stress distribution analysis for the Incore Instrumentation (ICI) and the Control Element Drive Mechanism (CEDM) penetration nozzles, as well as the fracture mechanics crack growth analysis for the CEDM penetration nozzles. A status report of the nozzle penetration inspections will be submitted on or about May 1,2005.

OPPD requests that the NRC complete its review and approval of the subject relaxation request by May 6,2005.

Employment with Equal Opportunity 4171

U. S. Nuclear Regulatory Commission LIC-05-0043 Page 2 The following commitment is made in this submittal:

1. If the NRC staff finds that the crack-growth formula in industry report MRP-55 is unacceptable, OPPD shall revise its analysis that justifies relaxation of Reference 3 within 30 days after the NRC informs OPPD of an NRC-approved crack growth formula. If OPPDs revised analysis shows that the crack growth acceptance criteria are exceeded prior to the end of the current operating cycle, this relaxation is rescinded and OPPD shall, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, submit to the NRC written justification for continued operation. If the revised analysis shows that the crack growth acceptance criteria are exceeded during the subsequent operating cycle, OPPD shall, within 30 days, submit the revised analysis for NRC review. If the revised analysis shows that the crack growth acceptance criteria are not exceeded during either the current operating cycle or the subsequent operating cycle, OPPD shall, within 30 days, submit a letter to the NRC confirming that its analysis has been revised. Any future crack-growth analyses performed for this and future cycles for RPV head penetrations must be based on an acceptable crack growth rate formula.

If you have any questions or require additional information, please contact Thomas R. Byrne at (402) 533-7368.

Ralph L. Phelps Division Manager Nuclear Engineering RLP/TRB/trb Attachment 1 - Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors

LIC-05-0043 Attachment I Page 1 Attachment 1 Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors

LIC-05-0043 Page 2 Attachment 1 Supplemental Information for Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors Executive Summary Section 1.O - Introduction - This section provides an overview of the process used to assess the hoop stresses for the crack growth calculations for the areas in which relaxation is requested.

Section 2.0 - Hoop Stress Distribution Above the Root of the CEDM/ICI Penetration Nozzle J-Groove Weld - A discussion of the hoop stresses found above all of the J-groove welds is provided and graphic results are provided in Figures 1 - 9. Specific information on the proposed relaxation areas (Control Element Drive Mechanism (CEDM) penetration nozzles 22 - 41) is provided and graphic results for these areas are provided in Figures 4 - 7.

Section 3.0 - Crack Growth Calculation for CEDM Penetration Nozzle Area Above the Root of the J-Groove Weld - Crack growth calculations for the proposed relaxation areas in CEDM penetration nozzles 22 - 41 (CEDM nozzle angles 37.3" and 41.7") are presented with the basis of why these crack growths are conservative. Figure 10 graphically summarizes these calculations.

Section 4.0 Hoop Stress Distribution Below the Toe of the CEDM/ICI Penetration Nozzle J-Groove Weld - This section completes the hoop stress assessments by presenting the results in areas below the toe of all of the J-groove welds, as shown in Figures 11 - 19.

Section 5.0 Potential Lack of 100% Eddy Current Coverage of the ICI Penetration Nozzle J-Groove Welds - This section provides the justification for the maintenance of structural integrity of the In-Core Instrumentation (ICI) penetration nozzles with a 10" unexamined area of the ICI J-groove welds.

Section 6.0 References

LIC-05-0043 Page 3 1.O Introduction The objective f this analysis was to obtain accurate stresses in th CEDM and ICI penetration nozzles and their immediate vicinity for FCS. To do so requires a three-dimensional finite element analysis (Reference 1) that considers all the pertinent loadings on the penetrations. Four CEDM locations with nozzle angles of O", 24.6", 37.3", 41.7" and one ICI nozzle with nozzle angle of 54.4" were analyzed. The analyses were used to provide information for the flaw tolerance evaluation and/or determine the adequacy of the inspection coverage. The methodology used in the finite element analysis (Reference 1) is consistent with that performed to support relaxation requests from Reference 7 for other nuclear power plants.

A three-dimensional finite element model comprised of isoparametric brick and wedge elements was used to obtain the stresses and deflections. Taking advantage of the symmetry of the reactor pressure vessel (RPV) head, only half of the CEDM/ICI penetration nozzles were modeled. In the model, the lower portion of the CEDM/ICI penetration nozzle, the adjacent section of the RPV head, and the joining weld were modeled. The vessel to penetration nozzle weld was simulated with two weld passes. The penetration nozzle, weld metal, cladding and the W V head shell were modeled in accordance with the relevant material properties.

The most important loading conditions which exist on the penetration include internal pressure and thermal expansion effects typical of steady state operation.

The RPV head temperature that OPPD used in the analysis is 588°F (Reference 8). In addition, residual stresses due to the welding of the penetrations to the RPV head were considered.

The hoop stress in the penetration nozzle resulting from the steady state operation loadings and welding residual stresses is much higher than the axial stress.

Therefore, only the hoop stress is assessed for crack growth calculations as part of this relaxation request. This is consistent with the field findings, where the cracks discovered are generally oriented axially. Typically, in-service cracks will orient themselves perpendicular to the largest stress component. Also it should be noted that the highest tensile hoop stress is at the uphill side and downhill side locations rather than midway around the penetration, where it is less limiting. This is consistent with finding axial cracks only in the vicinity of the uphill side and downhill side locations. It is these steady state hoop stresses that are used in this analysis to predict crack propagation in the penetration nozzles.

2.0 Hoop Stress Distribution Above the Root of the CEDM/ICI Penetration Nozzle J-Groove Weld Figures 1 - 9 show the hoop stress distributions for the regions that are within 2 inches from the top of the root of the J-groove weld on the uphill side for the FCS

LIC-05-0043 Page 4 CEDM and ICI RPV upper head penetrations. The stress distributions shown are for the inside and outside surface of the FWV upper head penetrations. The stress distributions shown in Figures 1 - 9 are typical of those observed in the upper RPV head penetration nozzles for other nuclear power plants. The stresses are highest in the vicinity of the J-groove weld and decrease rapidly as the distance above the root of the J-groove weld increases.

For the CEDM penetration nozzles numbers 22 - 41 (Figures 4 - 7) inspection coverage is less than 100%. The area where inspection coverage is not obtained is from 1 1/4 inches to 2 inches above the root of the J-groove on the uphill side above the thermal sleeve centering tabs. This area has a maximum hoop stress of 15 ksi. There is nearly universal agreement that high stresses, on the order of the material yield strength, are necessary to initiate Primary Water Stress Corrosion Cracking (PWSCC). There is no known case of stress corrosion cracking of Alloy 600 below the yield stress (Reference 2). Typical yield strengths for wrought Alloy 600 RPV head penetration nozzles are in the range of 37 ksi to 65 ksi. Weld metal yield strengths are generally higher. The yield strength of the CEDM head penetration nozzles for FCS varies from 37 ksi (nozzles 1 - 10 and 15 - 41) to 56 ksi (nozzles 11 - 14) (Reference 3). However, it has been determined that a stress level of 20 ksi is the point below which PWSCC initiation is extremely unlikely (Reference 2). Since the maximum hoop stress is only 15 ksi in the region where inspection coverage is less than loo%, PWSCC initiation in the region not being inspected is extremely unlikely. The area that cannot be inspected for the CEDM Inside Diameter (ID) scans is approximately 1.2% of the total area being inspected.

As shown in Figures 1 - 7, the hoop stresses are highest in the vicinity of the J-groove weld. It is more likely for an indication to exist and be detected in the inspection area of the penetration from 1.25 inches above the J-groove weld to 2.0 inches below the J-groove weld, which includes the high stress region in the vicinity of the J-weld. It is not likely that an indication will exist in the low stress region with a maximum hoop stress of only 15 ksi, which is the area not able to be inspected. The stresses in the area proposed to not be inspected are less than or equal to the stresses in the areas inspected.

LIC-05-0043 Page 5 Figure 1 Hoop Stress in 0' CEDM Nozzle vs. Distance from Top of Weld, Uphill and Downhill 70,000 60,000 50,000 40,000

-a f 30,000 e!

3i 20,000 10,000 0

-10,ooc 0.0 0.5 1.0 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)

LIC-05-0043 Attachment I Page 6 Figure 2 Hoop Stress in 24.6' CEDM Nozzle vs. Distance from Top of Weld, Uphill 70,000 60,000 50,000 40,000

-u) a U

I 30,000 I

t 20,000 10,000 0

-10,000 0.0 0.5 1 .o 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)

Figure 3 Hoop Stress in 24.6' CEDM Nozzle vs. Distance from Top of Weld, Downhill 70,000 60,000 50,000 40,000

.-a

-a-m 30,000 t

I 20,000 10,000 0

-10,000 0.0 1 .o 2.0 3.0 4.0 5.0 Distance Up from Top of The Weld (in.)

LIC-05-0043 Page 7 Figure 4 Hoop Stress in 37.3' CEDM Nozzle vs. Distance from Top of Weld, Uphill 80 000 I . . . _ ~ --"_,-".._-__~..._...._.-

......................... ......................... i I

......................... 1 ........................

I I the Rdot of Uphill Weld

-10 000 ... ................................................ '. .................................................. .....................

-20,0008 " " I ' " ' ~ ' ' I ' ' ~ " " I " '

0.0 0.5 1.o 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)

Figure 5 Hoop Stress in 37.3' CEDM Nozzle vs. Distance from Top of Weld, Downhill 70,000 I r 0.0 1.o 2.0 3.0 4.0 5.0 Distance Up from Top of The Weld (in.)

LIC-05-0043 Page 8 Figure 6 Hoop Stress in 41.7' CEDM Nozzle vs. Distance from Top of Weld, Uphill 0.0 0.5 1.0 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)

Figure 7 Hoop Stress in 41.7' CEDM Nozzle vs. Distance from Top of Weld, Downhill 0.0 1 .o 2.0 3.0 4.0 5.0 6.0 Distance Up from Top of The Weld (in.)

LIC-05-0043 Page 9 Figure 8 Hoop Stress in 54.4' ICI Nozzle VS. Distance from Top of Weld, Uphill 50,000 40,000 30,000

-a0 20,000 e!

ai 10,000 0

-1 0,000 0.0 0.5 1 .o 1.5 2.0 2.5 Distance Up from Top of The Weld (in.)

Figure 9 Hoop Stress in 54.4' ICI Nozzle vs. Distance from Top of Weld, Downhill 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 Distance Up from Top of The Weld (in.)

LIC-05-0043 Page 10 3 .O Crack Growth Calculation for CEDM Penetration Nozzle Area Above the Root of the J-Groove Weld A PWSCC crack growth calculation has been performed in the region above the root of the J-groove weld that is proposed to not be inspected. The purpose of the calculation is to determine the maximum flaw size for an axial inside surface flaw that would grow to 75% of the wall thickness in a single fuel cycle (18 months or 1.5 Effective Full Power Years (EFPY)). One fuel cycle was chosen since OPPD will be installing a new RPV head during the Fall 2006 Refueling Outage (RFO).

The methodology used in the crack growth calculation is consistent with the NRC flaw evaluation guidelines for the upper RPV head penetrations (Reference 4).

The PWSCC crack growth rate used in the NRC flaw evaluation guidelines is the same as that recommended in Reference 5. Assuming an aspect ratio of 6 (Reference 6), the crack growth results are shown in Figure 10 and summarized in Table 1 for both the downhill and uphill side of the two outermost rows of CEDM nozzle locations at FCS (CEDM nozzle penetrations 22 - 41).

Table 1 Minimum Flaw Size to Reach 75% of Wall Thickness in One Fuel Cycle for CEDM Nozzle Penetrations (Aspect Ratio = 6)

Minimum Flaw Size (% Through-wall)

CEDM Nozzle Angle (") Downhill Uphill 37.3 68.5 69.1 41.7 68.2 68.6 Based on the results given in Table 1, for an inside diameter axial surface flaw, a minimum initial flaw depth of 0.26 inch (68% part-through wall) is required to reach 75% of the wall thickness in one fuel cycle. For an aspect ratio of 6, the minimum initial flaw length is 1.56 inches long. Due to the low probability of PWSCC initiation in the low stress region that is more than 1.25 inches above the root of the J-groove weld on the uphill side, the existence of a 68% part-through wall inside diameter axial surface flaw with an aspect ratio of 6 in that region is extremely unlikely.

In addition, there is inherent conservatism in the above crack growth results.

From Table 5-3 of Reference 5 , the mean crack growth amplitude (a) for the Huntington Alloy 600 heats used in the development of the recommended PWSCC crack growth rate are summarized in Table 2.

LIC-05-0043 Page 11 Table 2 Mean MRP-55 Crack Growth Amplitude (a)for Huntington Material Test Data NX8101 I Huntindon I 1 . 3 7 ~ 0-l2 1

NX8664 I Huntington I 1.29xio-l2 NX64:20G I Huntinnton I 7.21~10"'

NX9240 Huntington 4 . 9 7 1~0-I' NX8 168G HuntinHon 1.93x10-I3 The recommended crack growth amplitude, a, from the NRC flaw evaluation guidelines (Reference 4) is 2.67x1O-l2,which is a factor of 1.9 higher than the highest growth crack amplitude for any of the Huntington material heats in Table

2. Since Huntington is the material supplier for the CEDM penetration nozzles for FCS, the PWSCC crack growth rate used in the above crack growth calculation and the resulting crack growth predictions are conservative.

Due to the low probability of PWSCC initiation or the existence of a 68% part-through inside diameter surface axial flaw in the proposed low stress regions not to be inspected, OPPD believes that compliance with the requirements in Reference 7 for the two outermost rows of CEDM penetration nozzles would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety per the information supplied in Reference 9.

LIC-05-0043 Page 12 Figure 10 PWSCC Flaw Growth Prediction at 1.25 Inch or More Above the Root of Uphill Side J-weld 1 .o 0.9 f

1

0.8 1

g 5

0.7 5

0.E 0.:

0 2 3 4 5 6 7 Time (Effective Full Power Years) 4.0 Hoop Stress Distribution Below the Toe of the CEDM/ICI Penetration Nozzle J-Groove Weld Figures 11 - 19 show the hoop stress distributions for the regions that are below the toe of the J-groove weld for the RPV upper head penetrations. The stress distributions shown are for the inside and outside diameter surface of the RPV upper head penetrations. The stress distributions shown in Figures 11 - 19 are typical of those observed in the upper head penetration nozzles for other nuclear power plants. The stresses are highest in the vicinity of the J-groove weld and decrease rapidly as the distance below the toe of the J-groove weld increases.

Based on Figures 11 - 19, the hoop stress for all the penetration nozzles is less than 20 ksi at a distance of 1 inch or more below the toe of the downhill side J-groove weld.

LIC-05-0043 Page 13 Figure 11 Hoop Stress in 0' CEDM Nozzle vs. Distance from Bottom of Weld, Uphill and Downhill 60,000 50,000 40.000 30,000 10,000 0

-1 0.000

-20,000 0.0 0.5 1 .o 1.5 2.0 2.5 Distance from Bottom of Weld (in.)

LIC-05-0043 Page 14 Figure 12 Hoop Stress in 24.6' CEDM Nozzle vs. Distance from Bottom of Weld, Uphill 0.0 0.5 1 .o 1.5 2.0 2.5 3.0 3.5 4.0 4.5 Distance from Bottom of Weld (in.)

Figure 13 Hoop Stress in 24.6' CEDM Nozzle vs. Distance from Bottom of Weld, Downhill 60.000 50,000

\

40,000 30,000

-a; 20,000 e

L 10,000 0

-10,000

-20,000 0.0 0.5 1 .o 1.5 2.0 2.5 Distance from Bottom of Weld (in.)

LIC-05-0043 Page 15 0.0 0.5 1.o 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 Distance from Bottom of Weld (in.)

Figure 15 Hoop Stress in 37.3' CEDM Nozzle vs. Distance from Bottom of Weld, Downhill 80,000 70,000 60,000 50,000 40,000 111 g

v) 20,000 10,000 0

-10.000

-20.000

-30,000 0.0 0.5 1.o 1.5 2.0 2.5 Distance from Bottom of Weld (in.)

LIC-05-0043 Page 16 Figure 16 Hoop Stress in 41.7' CEDM Nozzle vs. Distance from Bottom of Weld, Uphill 0.0 1.o 2.0 3.0 4.0 5.0 6.0 Distance from Bottom of Weld (in.)

Figure 17 Hoop Stress in 41.7' CEDM Nozzle vs. Distance from Bottom of Weld, Downhill 0.0 0.5 1.o 1.5 2.0 2.5 Distance from Bottom of Weld (in.)

LIC-05-0043 Page 17 Figure 18 Hoop Stress in 54.4' ICI Nozzle vs. Distance from Bottom of Weld, Uphill 60,000 I . . _ . -. -I_

I .

_- , - ~ . .

50,000 40,000 30,000

-.- 20,000

-a u) 10,000 z! 0

-10,000

-20,000

-30,000

-40,000 0.0 2.0 4.0 6.0 8.0 10.0 12.0 Distance from Bottom of Weld (in.)

Figure 19 Hoop Stress in 54.4' ICI Nozzle vs. Distance from Bottom of Weld, Downhill 0.0 0.5 1.o 1.5 2.0 Distance from Bottom of Weld (in.)

LIC-05-0043 Page 18 5.0 Potential Lack of 100% Eddv Current Coverage of the ICI Penetration Nozzle J-Groove Welds FCS has six ICI nozzles oriented around the outer periphery of the RPV head at the steepest angle (54.4") of any of the 48 nozzles. This steep angle and the original design information on the weld configuration indicated that up to 10" segments of the required inspection region at the downhill side (toes) of the J-groove welds on the ICI nozzles could not be inspected without great hardship, per the information supplied in Reference 9. Not inspecting these 10" segments would still result in 97.2% inspection coverage of the total ICI J-groove weld area.

OPPD has recently used a micro-camera to look more closely at the downhill area of several J-groove welds and has observed additional weld passes made on these welds that were beyond the original design, resulting in a flatter aspect to the J-weld in this steep downhill region. Therefore, there appears to be less potential for probe lift off than indicated on the drawings. Consequently, more inspection coverage of the 10" segments by eddy current testing may be available than initially estimated. OPPD plans to examine as much of the 360" area as possible for all CEDM nozzle penetrations.

OPPD has completed the 2005 RFO bare metal visual inspection on all the RPV head penetrations and found no evidence of any leakage. During the 2005 RFO, OPPD has also completed eddy current testing of the J-groove welds of all the CEDM nozzles and, based on analysis complete at the time this letter was signed (39 out of 41 CEDM nozzles), has found no indications of cracking. No indications of cracking in the CEDM J-groove welds is significant because the stresses in these welds are higher than the stresses in the J-groove welds of the ICI nozzles. The stresses at the toe of the J-groove weld for the ICI nozzles are lower primarily because the offset of weld geometry precludes a longer weld and because the toe of the weld is further away from the high stress area of the J-groove weld. Therefore, cracking would be expected to occur first in the CEDM J-groove welds, and would be less likely in the ICI J-groove welds.

Finally, no incidence of cracking in the ICI J-groove welds resulting in leakage has been found in any other Combustion Engineering (CE) plants to date. This lack of cracking at other CE plants corroborates the stress analyses performed at FCS.

In conclusion, lack of inspection coverage for 2.8% of the total area for the J-groove welds of the ICI nozzles is not significant. OPPD considers that obtaining inspection data for these areas would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety per the information supplied in Reference 9. The completed inspection coverage area is sufficient to maintain structural integrity over the single cycle of operation prior to RFV head replacement during the Fall 2006 RFO. Information regarding the completed inspection coverage for the ICI nozzles will be submitted to the NRC

LIC-05-0043 Page 19 as part of the status report of the nozzle penetration inspections on or about May 1,2005.

6.0 References

1. Dominion Engineering ~ c Calculation

. Number C-8718 Fort Calhoun CEDM and ICI Nozzle Stress Analysis, April 2005, Rev. 0.

2. Materials Reliability Program: Generic Evaluation of Examination Coverage Requirements for Reactor Pressure Vessel Head Penetration Nozzles (MRP-99, EPRI, Palo Alto, CA, 2003. 1009129.

3. CE NPSD-903-P, CEOG Task 730, CEOG Program to Address Alloy 600 Cracking of CEDM Penetrations, Subtask 1, Nozzle Evaluation, February 1993.

4. USNRC Letter, R. Barrett to A. Marion, Flaw Evaluation Guidelines, April 11,2003.

5. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick Wall Alloy 600 Material (MRP-55) Revision 1, EPRI, Palo Alto, CA, November 2002. 1006695.

6. ASME Code Section XI, 1998 Edition with 2000 Addenda, Article L-3210.

7. Letter from R. William Borchardt (NRC) to Ross Ridenoure (OPPD) dated February 20, 2004, Issuance of First Revised NRC Order (EA-03-009)

Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors (NRC-04-0022) (ML040220181).

8. EPRI MRP-48, PWR Materials Reliability Program Response to NRC Bulletin 2001-01, August 2001.

9. Letter from Ralph L. Phelps (OPPD) to Document Control Desk (NRC) dated April 7, 2005, Fort Calhoun Station Unit No. 1, Relaxation Request for First Revised Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors (LIC-05-0040).