ML101470453
| ML101470453 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 07/23/2010 |
| From: | Markley M Plant Licensing Branch IV |
| To: | Entergy Operations |
| Kalyanam N, NRR/DORL/LP4, 415-1480 | |
| References | |
| TAC ME2411, W3-ISI-016 | |
| Download: ML101470453 (10) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 23, 2010 Vice President, Operations Entergy Operations, Inc.
Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093
SUBJECT:
WATERFORD STEAM ELECTRIC STATION, UNIT 3 - REQUEST FOR ALTERNATIVE W3-ISI-016, INSPECTION OF REACTOR PRESSURE VESSEL HEAD CONTROL ELEMENT DRIVE MECHANISM NOZZLES DURING THIRD 10-YEAR INSERVICE INSPECTION INTERVAL (TAC NO. ME2411)
Dear Sir or Madam:
By letter dated October 19, 2009, as supplemented by letter dated November 2, 2009, Entergy Operations, Inc. (Entergy, the licensee), submitted Request for Alternative W3-ISI-016 for Waterford Steam Electric Station, Unit 3 (Waterford 3). In its submittal, the licensee proposed an alternative to the inspection requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Case N-729-1, "Alternative Examination Requirements for PWR [Pressurized-Water Reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1,n as required and conditioned by paragraph 50.55a(g)(6)(ii)(D) of Title 10 of the Code of Federal Regulations (10 CFR) for the fall 2009 (RF 16) refueling outage during the third 1O-year inservice inspection (lSI) interval at Waterford 3.
Specifically, the licensee stated that, due to the guide-cone threaded connection at the bottom of the control element drive mechanism (CEDM) nozzle and limitations due to nozzle configuration and inspection probe design, inspection of the nozzle, to the extent required by Code Case N-729-1, would result in a hardship or unusual difficulty without a compensating increase in the level of quality or safety. Entergy has stated that, alternately, it plans to inspect the reactor pressure vessel head (RPVH) CEDM penetration nozzles using the ultrasonic testing (UT) method to the extent possible. Additionally, Entergy will perform a bare-metal visual inspection of the RPVH surface and a demonstrated volumetric leak-path assessment.
Also, licensee stated that the prior finite element and fracture mechanics crack growth analysis shows that a postulated through-wall crack in the region of nozzle below the J-groove weld toe that cannot be examined by UT will not propagate to the toe of the J-groove weld for one remaining operating cycle.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's request and determined that the licensee's proposed alternative inspection provides a reasonable assurance of structural integrity and leak-tightness of the CEDM nozzles and that the inspection of the nozzle to the extent required by Code Case N-729-1 would result in a hardship or unusual difficulty without a compensating increase in the level of quality or safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes the licensee's proposed alternative at
- 2 Waterford 3, up to the commencement of the 17th refueling cycle in spring of 2011 when the RPVH will be replaced.
Due to the immediate need of this relief for alternative request, the staff previously provided verbal authorization for the use of this alternative for Waterford 3, on November 4, 2009. The staffs verbal authorization is documented in electronic mail correspondence dated November 4, 2009.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
The NRC staffs safety evaluation is enclosed.
Sincerely, Michael 1. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-382
Enclosure:
Safety Evaluation cc w/encl.: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE W3-ISI-016, INSPECTION OF REACTOR PRESSURE VESSEL HEAD CONTROL ELEMENT DRIVE MECHANISM NOZZLES DURING THIRD 10-YEAR INSERVICE INSPECTION INTERVAL WATERFORD STEAM ELECTRIC STATION, UNIT 3 ENTERGY OPERATIONS, INC.
DOCKET NO. 50-382
1.0 INTRODUCTION
By letter dated October 19, 2009 (Reference 1). as supplemented by letter dated November 2, 2009 (Reference 2), Entergy Operations, Inc. (the licensee), submitted relief request W3-ISI-016 for U.S. Nuclear Regulatory Commission (NRC) review and approval. The request pertains to inservice inspection (lSI) of reactor pressure vessel head (RPVH) control element drive mechanism (CEDM) nozzles at Waterford Steam Electric Station, Unit 3 (Waterford 3). The licensee requests relief from the inspection requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Case N-729-1, "Alternative Examination Requirements for PWR [Pressurized-Water Reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial Penetration Welds,Section XI, Division 1," as required and conditioned by paragraph 50.55a(g)(6)(ii)(D) of Title 10 of the Code of Federal Regulations (10 CFR). Specifically, the licensee stated that, due to the guide-cone threaded connection at the bottom of the CEDM nozzle, inspection of the nozzle, to the extent required by Code Case N-729-1, would result in a hardship without a compensating increase in the level of quality or safety.
The NRC staff previously provided verbal authorization for the use of this alternative for Waterford 3, on November 4, 2009. The staffs verbal authorization is documented in electronic mail correspondence dated November 4,2009 (Reference 3).
Enclosure
- 2
2.0 REGULATORY EVALUATION
The regulations in 10 CFR 50.55a(g)(6)(ii)(0) require augmented lSI of reactor vessel head penetration nozzles of pressurized-water reactors (PWRs) in accordance with ASME Code Case N-729-1, subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(0), paragraphs (2) through (6). Paragraph (3) states in part:
Instead of the specified 'examination method' requirements for volumetric and surface examinations in Note 6 of Table 1 of Code Case N-729-1, the licensee shall perform volumetric and/or surface examination of essentially 100 percent of the required volume or equivalent surfaces of the nozzle tube, as identified by Figure 2 of ASME Code Case N-729-1."
The regulations in 10 CFR 50.55a(a)(3) state, in part, that alternatives to the requirements of 10 CFR 50.55a(g) may be used when authorized by the NRC, if the applicant demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The licensee's relief request, which defines an alternative examination volume or surface for each nozzle, has been submitted on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The Code of record for Waterford 3 for the third 10-year lSI interval is the 2001 Edition through 2003 Addenda of Section XI of the ASME Code. The licensee requests authorization of the proposed alternative for the fall 2009 refueling outage and states that during refueling outage 17, in spring 2011, the RPVH will be replaced.
3.0 TECHNICAL EVALUATION
3.1 Components Affected The components affected are 91 ASME Code Class 1 RPVH penetration CEOM nozzles, Component Numbers 02-T-01 through 02-T-91, and their associated welds, identified by item number 84.20 in Table 1 of ASME Code Case N-729-1.
3.2 Code Requirements The regulations in 10 CFR 50.55a(g)(6)(ii)(D)(1) state, in part, that licensees of existing operating PWRs shall augment their existing lSI program with ASME Code Case N-729-1, subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(D), paragraphs (2) through (6).
The licensee is to perform a volumetric and/or surface examination of essentially 100 percent of the required volume or equivalent surfaces of the nozzle tube, as identified by Figure 2 of ASME Code Case N-729-1. Figure 2 identifies the required volume of tube to be inspected as a distance "a" above the highest point of the root of the J-groove weld to a distance "a" below the lowest point of the toe of the J-groove weld. The distance "a" is equal to 1.5 inches (38 millimeters (mm)) for incidence angle, e, S 30 degrees and for all nozzles ~ 4.5 inches (115 mm) outside diameter (00) or 1 inch (25 mm) for incidence angle, e, ~ 30 degrees; or to
- 3 the end of the tube, whichever is less. If a surface examination is being substituted for a volumetric examination on a portion of a penetrating nozzle that is below the toe of the J-groove weld, the surface examination shall be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically.
3.3 Licensee's Proposed Alternative and Basis The licensee proposes to perform an Electric Power Research Institute (EPRI)-qualified ultrasonic testing (UT) volumetric examination, in accordance with the requirements of 10 CFR 50.55a(g)(6)(ii)(D)(4), from the inside diameter (ID) of each CEDM nozzle from the applicable distance "a", as defined in Figure 2 of Code Case N-729-1, above the highest point on the root of the J-groove weld down to a point approximately 1.544 inches above the bottom of the nozzle. The 1.544-inch distance is determined by the length of the threaded portion at the bottom of the nozzle that is used to attach the CEDM guide-cone plus a length resulting from the chamfer region and the size of the ultrasonic transducer. The licensee will also perform a bare-metal visual (BMV) inspection of the RPVH surface and a demonstrated volumetric leak path assessment.
The licensee has performed a finite element (FE) and fracture mechanics (FM) crack growth analysis to show that a postulated through-wall crack in the region of nozzle below the J-groove weld toe that cannot be examined by UT will not propagate to the toe of the J-groove weld within one operating cycle. This analysis is documented in Engineering Report M-EP-2003-004, Rev. 0, "Fracture Mechanics Analysis for the Assessment of the Potential for Primary Water Stress Corrosion Crack (PWSCC) Growth in the Uninspected Regions of the Control Element Drive Mechanism (CEDM) Nozzles at Waterford Steam Electric Station, Unit 3" (Reference 4).
As an alternative to UT examination of the CEDM nozzles, compliance with the requirements of 10 CFR 50.55a(g)(6)(ii)(D)(3) can be achieved with surface examination of the nozzle ID and OD wetted surfaces. In a letter dated August 13, 2004 (Reference 5), the licensee stated that manually performing a surface examination on the outside surface would increase personnel radiation exposure significantly, estimated to be 27 to 45 person-roentgen equivalent man (rem) for examination of the 91 CEDM nozzles.
The licensee states that the combination of EPRI-qualified UT examinations to the extent possible with an engineering evaluation that includes FE stress analysis and FM crack growth evaluations will determine whether sufficient crack propagation length exists between the tip of a postulated crack in the region that cannot be inspected with UT and the toe of the J-groove weld to facilitate one refueling cycle of crack growth without the crack intersecting the toe of the J-groove weld, thus assuring that the integrity of the pressure boundary until the RPVH can be replaced during the next refueling outage.
4.0 NRC STAFF EVALUATION The susceptibility of RPVH penetration nozzles in PWRs to primary water stress-corrosion cracking (PWSCC) is a safety concern. The nozzles are nickel-based alloys and are welded using nickel-based weld metal alloy to reactor pressure vessel upper head. Primary-water coolant, high-tensile stresses, and elevated operating temperatures of PWRs can result in
- 4 PWSCC cracking of the nickel-based alloys used in the CEDM nozzles, and the J-groove weld and butter material. The subject CEDM nozzles at Waterford 3 meet the conditions for PWSCC and, therefore, may be susceptible to cracking in the nozzles and associated welds which could result in nozzle ejection and/or leakage of boric acid causing corrosion of the low-alloy steel head.
The licensee has identified physical limitations which prevent full volumetric inspections of the nozzle below the J-groove weld. These limitations include guide-cone threaded connection with a welded set screw and two tack welds, and a 45-degree chamfer at the cone-nozzle interface.
This configuration results in a zone at the bottom of each nozzle which cannot be volumetrically inspected to the extent required by Code Case N-729-1. To overcome these physical limitations, the cones would have to be removed, redesigned, and reinstalled at a significant occupational radiological exposure to the personnel. The licensee has the option of performing surface inspections of the ID and OD wetted surfaces of each nozzle to meet the current requirements for the volumetric inspection. However, the licensee has estimated that surface examination would result in a high radiation exposure to the workers, between 27 and 45 person-rem for examination of all 91 nozzles (Reference 5). The NRC staff concurs that removal and reinstallation of the guide cones, or performing a manual surface examination of the nozzle, would result in exposure of personnel to significant radiation dose. Therefore, the staff concludes that compliance with the requirements of 10 CFR 50.55a(g)(6)(ii)(D) would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.
As an alternative, the licensee has proposed to perform a UT examination of the subject nozzles to the required distance "a" per Code Case N-729-1 above the root of the J-groove weld, and to the maximum extent possible below the J-groove weld. The examination will be performed by EPRI-qualified personnel and techniques in accordance with the requirements of 10 CFR 50.55a(g)(6)(ii)(D)(4). In addition, a demonstrated UT leak-path assessment and BMV examination of the RPVH surface will be performed. The licensee has previously carried out analysis of the expected propagation of a postulated through-wall axial crack that could exist within the unexamined zone below the J-groove weld (Reference 4). This analysis consisted of two parts: an FE analysis of the J-groove weld residual stress fields and an FM evaluation of the propagation distance of a postulated through-wall crack within the unexamined zone. The FE analysis used the as-built geometry and sizes that had been determined from previous UT nozzle examinations at a sister plant. In response to the NRC staff's request for additional information (RAI) dated October 28, 2009 (Reference 6), concerning whether reanalysis using as-built data from Waterford 3 is necessary, the licensee stated that the Waterford 3 as-built nozzle configurations, determined by UT examination of each nozzle during refueling outage 12, are bounded by the previous analysis and no revision to the analysis is necessary. This conclusion and the FM evaluation of crack growth within one refueling cycle were accepted by the NRC staff in a safety evaluation dated March 22, 2005 (Reference 7).
The licensee has predicted the propagation distance of a hypothetical through-wall crack located in the uninspected zone at the bottom of the nozzle with its tip at the upper boundary of the unexamined zone and extending down to a point where cracking is no longer expected. The analysis was performed for four different RPVH penetration nozzle angles: 0, 7.8, 29.1, and 49.7 degrees. In its letter dated November 2, 2009, in response to the NRC staff's RAI, the
- 5 licensee stated that each of these groups represented the bounding condition for the nozzles higher on the head (Le., analysis for the 29.1 degree group bounds the intermediate nozzles between 7.8 and 29.1 degrees). In Table 1 below, the column labeled "Predicted Propagation in One Cycle (inch)" lists the values for the predicted propagation distance for the appropriate bounding case of the incidence angle. These values were provided on page 12 of the licensee's RAI response, submitted by letter dated October 24, 2003 (Reference 8), of the previous relief request, submitted by letter dated September 15, 2003 (Reference 9). The NRC staff accepted the analysis of this relief request in a safety evaluation dated November 12, 2003 (Reference 10), conditioned on the acceptance of the crack growth equation and constants that were used. The staff notes that the crack growth equation and constants that were used in the licensee's analysis are the same as those for RPVH penetration nozzles given in ASME Code,Section XI, Appendix 0, 2004 Edition; therefore, the staff concludes that their use is acceptable.
In response the NRC staff's RAI dated October 28, 2009, the licensee provided data (Table 1 of the RAI response) for the as-measured freespan length, the length of the CEDM nozzle below the toe of the J-groove weld that can be examined by UT, of each Waterford 3 CEDM nozzle that was examined with UT during the April 2005 outage. The staff determined the minimum freespan length for each of the same four bounding head angles in Table 1 and tabulated these values in the column labeled "Minimum Length Examined by UT in 2005 (inch)" as follows:
Table 1 Nozzle Group (Head Angle Degrees)
Predicted Propagation in One Cycle (inch)
Minimum Length Examined by UT in 2005 (inch) 0 0.265 1.24 7.8 0.25 1.08 29.1 0.16 0.88 49.7 0.16 0.44 The licensee's flaw-growth analysis evaluated the distance that a hypothetical crack in the uninspected zone at the bottom of the nozzle could grow in one cycle of operation. If the calculated crack propagation distance in one cycle of operation is less than the distance below the J-groove weld toe which is inspected using UT, then the nozzle pressure boundary integrity is ensured for at least one cycle of operation. For each of the bounding nozzle group angles in Table 1, the predicted propagation distance is less than the minimum length which can be examined by Code compliant UT. For each of the nozzle groups, the predicted propagation distance in one cycle of operation is less than one-half of the distance below the J-groove weld toe than can be examined by UT. Furthermore, a crack would have to propagate past the toe of the J-groove weld to the weld root before the pressure boundary integrity would be violated.
Therefore, the NRC staff concludes that there is sufficient margin to ensure pressure boundary integrity of the Waterford 3 RPVH CEDM nozzles for one remaining operating cycle and that the licensee's proposed alternative provides reasonable assurance of structural integrity and leak tightness of the CEDM nozzles.
In summary, the NRC staff concludes that the licensee's proposed alternative inspection provides a reasonable assurance of structural integrity and leak-tightness of the CEDM nozzles,
- 6 and that compliance with the requirements of 10 CFR 50.55a(g)(6)(ii)(D)(3) would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety..
5.0 CONCLUSION
As set forth above, the NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity and leak-tightness of the CEDM nozzles, and that complying with the requirements of ASME Code Case N-729-1, as required and conditioned by 10 CFR 50.55a(g)(6)(ii)(D), would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii) and authorizes the licensee's proposed alternative at Waterford 3, up to the commencement of the 17th refueling cycle in spring of 2011 when the RPVH will be replaced.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
6.0 REFERENCES
- 1.
Murillo, R. J., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request for Alternative W3-ISI-016, Inspection of Reactor Pressure Vessel Head Control Element Drive Mechanism Nozzles during Third Ten-Year Inservice Inspection Interval," dated October 19, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092940243).
- 2.
Murillo, R. J., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Response to NRC Request for Additional Information for Alternative W3-ISI-016, Inspection of Reactor Pressure Vessel Head Control Element Drive Mechanism Nozzles during Third Ten-Year Inservice Inspection Interval," dated November 2,2009 (ADAMS Accession No. ML093080128).
- 3.
Kalyanam, K., U.S. Nuclear Regulatory Commission, electronic mail to Robert J. Murillo and Jim Pollack, Entergy Operations, Inc., "Verbal Authorization of Request for Alternative - W3-ISI-016," dated November 4, 2009 (ADAMS Accession No. ML093080814).
- 4.
Entergy Operations, Inc., "Engineering Report M-EP-2003-004 Rev. 0, Fracture Mechanics Analysis for the Assessment of the Potential for Primary Water Stress Corrosion Crack (PWSCC) Growth in the Uninspected Regions of the Control Element Drive Mechanism (CEDM) Nozzles at Waterford Steam Electric Station, Unit 3," to Reference 8 (below) (ADAMS Accession No. ML032790354).
- 5.
Burford, F. G., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Response to NRC Request for Additional Information Pertaining to Waterford 3 Relaxation Request #4 to NRC Order EA-03-009 for the Control Element Drive Mechanism Nozzles," dated August 13, 2004 (ADAMS Accession No. ML042320558).
- 7
- 6.
Kalyanam, K., U.S. Nuclear Regulatory Commission, electronic mail to Robert J. Murillo and Jim Pollack, Entergy Operations, Inc., "Formal RAI for W3-ISI-016," dated October 28, 2009 (ADAMS Accession No. ML093010525).
- 7.
Berkow, H. N., U.S. Nuclear Regulatory Commission, letter to Joseph E. Venable, Entergy Operations, Inc., "Safety Evaluation for Waterford Steam Electric Station, Unit 3 - Relaxation Request from U.S. Nuclear Regulatory Commission (NRC) First Revised Order EA-03-009 for Control Rod Drive Mechanism (CEDM) Nozzles," dated March 22,2005 (ADAMS Accession No. ML050820683).
- 8.
Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Response to a Request for Additional Information Pertaining to Relaxation Request to NRC Order EA-03-009 for the Control Element Drive Mechanism Nozzles," dated October 24, 2003 (ADAMS Accession No. ML033110181).
- 9.
Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Relaxation Request to NRC Order EA-03-009 for the Control Element Drive Mechanism Nozzles," dated September 15, 2003 (ADAIV1S Accession No ML032790346).
- 10.
Berkow, H. N., U.S. Nuclear Regulatory Commission, letter to Michael A Krupa, Entergy Operations, Inc., "Waterford Steam Electric Station Unit 3 (Waterford 3) - Relaxation Request from U.S. Nuclear Regulatory Commission (NRC) Order EA-03-009 for the Control Element Drive Mechanism Penetration Nozzles (TAC No. MB9644)," dated November 12, 2003 (ADAMS Accession No. ML033160192).
Principal Contributor: J. Wallace Date: July 23, 2010
- SE memo dated OFFICE NRR/LPL4/PM NRR/LPL4/LA NRRIDCIICPNB NRR/LPL4/BC NAME NKalyanam
..IBurkhardt TLupold*
MMarkley (CFLyon for)
DATE 7/23/10 7/23/10 5/21/10 7/23/10