ML092080131
| ML092080131 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 07/31/2009 |
| From: | Markley M Plant Licensing Branch IV |
| To: | Edington R Arizona Public Service Co |
| Hall, J R, NRR/DORL/LPL4, 301-415-4032 | |
| References | |
| TAC ME0664, TAC ME0665, TAC ME0666 | |
| Download: ML092080131 (12) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 July 31, 2009 Mr. Randall K. Edington Executive Vice President Nuclear/
Chief Nuclear Officer Mail Station 7602 Arizona Public Service Company P. o. Box 52034 Phoenix, AZ 85072-2034
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 RELIEF REQUEST NO. 41, REQUEST TO USE APPENDIX I OF ASME CODE CASE N-729-1 FOR REACTOR PRESSURE VESSEL HEAD INSPECTIONS (TAC NOS. ME0664, ME0665, AND ME0666)
Dear Mr. Edington:
By letter dated February 17,2009, Arizona Public Service Company (APS, the licensee) submitted Relief Request No. 41 (RR-41), requesting relief from the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(6)(ii)(D), for the third 10-year inservice inspection interval for the Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3.
The submittal requests U.S. Nuclear Regulatory Commission (NRC) approval for use of the analysis procedures of Appendix I of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code Case N-729-1, for the examination of reactor pressure vessel upper head penetrations for the control element drive mechanisms (CEDMs). NRC prior approval for the use of the Appendix I methods is required by 10 CFR 50.55a(g)(6)(ii)(D)(6).
APS requested approval of RR-4'1 by May 7, 2009, in order to support the start-up of the Unit 3 reactor following completion of the 2009 refueling outage. The NRC staff reviewed the licensee's relief request and found it acceptable, but the staff did not have sufficient time to document its safety evaluation by the requested approval date. Therefore, on May 6, 2009, the staff verbally authorized the implementation of the proposed alternative. This letter transmits the staff's written safety evaluation and the basis for authorizing the proposed alternative. On the basis of the information submitted by APS, the staff concluded that the proposed alternative provides reasonable assurance of structural integrity of the CEDM nozzles. Furthermore, performance of the required examinations for all CEDM nozzles would present a significant hardship due to radiological exposure without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the use of the proposed alternative is authorized for PVNGS, Units 1, 2, and 3, for the third 10-year inservice inspection interval for each unit, or until that unit's reactor vessel head is replaced, whichever occurs first.
All other requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D)(6) not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
R. Edington
- 2 A copy of the related Safety Evaluation is enclosed. If you have any questions, please contact Randy Hall at (301) 415-4032 or via email at randy.hall@nrc.gov.
Sincerely, Michael 1. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN-528, STN-529, and STN 50-530
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 TO USE APPENDIX I OF ASME CODE CASE N-729-1 INSERVICE INSPECTION RELIEF REQUEST NO. 41 ARIZONA PUBLIC SERVICE COMPANY, ET AL.
PALO VERDE NUCLEAR GENERATING STATION, UNITS 1. 2. AND 3 DOCKET NOS. STN 50-528. STN 50-529. AND STN 50-530
1.0 INTRODUCTION
By letter dated February 17,2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090580199), Arizona Public Service Company (APS, the licensee),
requested U.S. Nuclear Regulatory Commission (NRC) authorization for Relief Request No. 41 (RR-41), "Request to Use Appendix I of ASME Code Case N-729-1." The licensee is proposing to perform an alternative VOlumetric and/or equivalent surface examination for each control element drive mechanism (CEDM) penetration nozzle tube. The proposed examinations of the CEDM nozzle tubes would be an alternative to the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a, paragraph (g)(6)(ii)(D), for the third 10-year inservice inspection interval for Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3.
APS requested approval of RR-4'1 by May 7, 2009, in order to support the start-up of the Unit 3 reactor following completion of the 2009 refueling outage. The NRC staff reviewed the licensee's relief request and found it acceptable, but the staff did not have sufficient time to document its safety evaluation by the requested approval date. Therefore, on May 6, 2009, the staff verbally authorized the implementation of the proposed alternative. This safety evaluation documents the NRC staff's basis for authorizing the proposed alternative.
2.0 REGULATORY EVALUATION
The regulations in 10 CFR 50.55a(g)(6)(ii)(D), "Reactor vessel head inspections," require augmented inservice inspection of reactor vessel head penetration nozzles of pressurized water reactors in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, Enclosure
- 2 Section XI, Division 1," subject to the conditions specified in paragraphs (2) through (6) of 10 CFR 50.55a(g)(6)(ii)(D). 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 extent of the examination of the nozzle tube is determined by the incidence angle, e, and the distance "a" above and below the J-groove weld, as defined in Figure 2 of the Code Case.
The Code Case specifies that:
a = 1.5 in. (38 mm) for Incidence Angle, e, S 30 deg and for all nozzles ~ 4.5 in.
(115 mm) 00, or 1 in. (25 mm) for Incidence Angle, e, ~ 30 deg; or to the end of the tube, whichever is less.
The Code Case recognizes that impediments, such as physical obstructions, threads on the nozzle end, or an ultrasonic examination corner shadow zone, could prevent examination of the complete zone for one or more nozzles, and provides an analysis procedure in Appendix I to Code Case N-729-1 to demonstrate the adequacy of an alternate examination volume for each such nozzle. However, paragraph (6) of 10 CFR 50.55a(g)(6)(ii)(D) states that Appendix I of ASME Code Case N-729-1 shall not be implemented without prior NRC approval.
Paragraph (a)(3) of 10 CFR 50.55a states, in part, that alternatives to the requirements of paragraph (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 relief request from the licensee to define an alternative examination volume or surface for each nozzle has been submitted on the basis that the proposed alternative would provide an acceptable level of quality and safety. The Code of record for the third 1O-year inservice inspection interval for PVNGS, Units 1,2, and 3 is the 2001 Edition through 2003 Addenda of Section XI of the ASME Code.
The licensee requests relief for the third 1O-year inservice inspection interval of each unit, which began on July 8, 2008, March 18, 2007, and January 11, 2008, for Units 1, 2, and 3, respectively; or until each unit's reactor vessel head has been replaced, whichever occurs first.
3.0 TECHNICAL EVALUATION
3.1 Affected Systems and Components The affected items are all of the CEDM nozzle penetrations for PVNGS, Units 1, 2, and 3, designated as Item Number B4.20, "UNS N06600 nozzles and UNS N06082 or UNS W86182 partial-penetration welds in head," in Table 1 of Code Case N-729-1.
- 3 3.2 Proposed Alternative The licensee proposes to perform ultrasonic examination (UT) of each reactor pressure vessel head CEDM penetration nozzle (i.e., nozzle base material) for a distance equal to "a" above the J-groove weld on the uphill slope, as defined by Figure 2 of Code Case N-729-1, and to the minimum required inspection distances below the J-groove weld on the downhill slope, as identified in Tables 1, 2, and 3, including footnote 1 on Table 1, of the relief request. Licensee procedures require examination of each CEDM nozzle as far down below the J-groove weld toe as practical. If inspection limitations prevent UT examination to the minimum distances identified in the Tables, then a surface examination of the remaining wetted surface will be conducted on the affected nozzle(s).
Table 1 PVNGS Unit 1 CEDM Nozzle Minimum Required Inspection Coverage Nozzle Angle (0)
Penetration No.
Applicability Minimum Inspection Coverage Required Below the J-Groove Weld Toe on the Downhill Side (in)
EFPY for the Upper Crack Tip to Reach the J-Groove Weld Toe 0
1 0.45 1.7 7.5 2-21 0.45 1.7 28.0 22-45 0.45 1.8 35.7 46-83, 85, 90-92, 94-97 0.40 1.7 51.5 86,88-89 0.35 1.9 35.7 841 0.24 No Propagation Predicted 35.7 931 0.16 No Propagation Predicted 51.5 871 0.32 No Propagation Predicted 1 Nozzles 84, 87 and 93 actual inspection coverage included a 0.04-inch instrument uncertainty. nlis information was provided in a revised analysis of as-built nozzles, licensee's letter dated July 1, 2004, which included Westinghouse Letter PAFM-04-39, Revision 1, ADAMS Accession No. ML041950341.
- 4 Table 2 PVNGS Unit 2 CEDM Nozzle Minimum Required Inspection Coverage Nozzle Angle (0)
Penetration No.
Applicability Minimum Inspection Coverage Required Below the J-Groove Weld Toe on the Downhill Side (in)
EFPY for the Upper Crack Tip to Reach the J-Groove Weld Toe 0
1 0.45 1.7 7.5 2-21 0.45 1.7 28.0 22-45 0.45 1.8 35.7 46-55, 57-85, 90-97 0.40 1.7 35.7 562 0.36 No Propagation Predicted 51.5 86-89 0.35 1.9 2 Nozzle 56 actual inspection coverage included a 0.04" instrument uncertainty. This information was provided in a revised analysis of the as-built nozzle in licensee's letter dated November 24,2004, ADAMS Accession No. ML043420167.
Table 3 PVNGS Unit 3 CEDM Nozzle Minimum Required Inspection Coverage Nozzle Angle (0)
Penetration No.
Applicability Minimum Inspection Coverage Required Below the J-Groove Weld Toe on the Downhill Side (in)
EFPY for the Upper Crack Tip to Reach the J-Groove Weld Toe 0
1-29 0.40 1.7 31.5 30-81 0.35 2.0 47.6 82-85 0.30 2.4 49.5 90-97 0.30 3.4 51.5 86-89 0.20 2.4 3.3 Basis for Request The licensee states that the design of the funnel attachment to the CEDM nozzles (Le., inside threaded connection with weld plug), makes it unable to comply with the requirements to perform a volumetric examination of the CEDM penetration nozzle (Le., nozzle base material) region A-B-C-D, as defined in Figure 2 of ASME Code Case N-729-1, as required by 10 CFR 50.55a(g)(6)(ii)(D). Specifically, the length of the nozzle having the distance "a" below the J-groove weld toe cannot be successfully examined with UT due to multiple ultrasonic signals reflected back from the threaded surfaces. The licensee states that flaws outside of the proposed alternate examination zone will not propagate to the toe of the CEDM nozzle J-groove weld prior to the next examination; therefore, crack propagation will not lead to a safety concern or an unacceptable probability of leakage in the time interval before the next refueling outage.
- 5 3.4
NRC Staff Evaluation
The phenomenon of concern is primary water stress-corrosion cracking (PWSCC). PWSCC typically initiates in the areas of the highest tensile stress in susceptible materials, such as UNS N06600 material, and propagates in a controlled fashion in response to environmental conditions of time, temperature, stress, and a corrosive environment. The areas of CEDM penetrations that have the highest residual stress are the areas adjacent to the J-groove attachment weld. The CEDM nozzles at PVNGS meet the conditions for potential PWSCC, thus may be susceptible to cracking and leaking of primary water if propagating cracks intersect the J-groove weld.
The regulations in 10 CFR 50.55a(g)(6)(ii)(D)(3) state:
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 extent of examination is defined from the distance "a" above the highest point at the root of the J-groove weld to a distance "a" below the lowest point at the toe of the J-groove weld.
The licensee states that the distance equal to "a" above the J-groove weld on the uphill slope, as defined by Figure 2 of Code Case N-729-1, will be examined using UT but, due to the design of the funnel attachment to the CEDM nozzles, UT examination to the distance "a" below the lowest point of the toe of the J-groove weld for all of the CEDM nozzles cannot be achieved.
The licensee proposes to examine an alternate examination zone below the toe of the J-groove weld, shown in the columns "Minimum Inspection Coverage Required Below the J-Groove Weld Toe on the Downhill Side (in)" in Tables 1-3, above. This alternate examination zone will be examined using UT to the extent possible, and only in those cases where the minimum inspection coverage in the tables cannot be achieved using UT will the remaining wetted surfaces be examined by a surface examination. The licensee states that flaws outside of this alternate examination zone will not propagate to the toe of the CEDM nozzle J-groove weld prior to the next examination, thus crack propagation will not lead to a safety concern or an unacceptable probability of leakage in the time interval before the next refueling outage.
Therefore, examination of this alternate examination zone below the toe of the J-groove weld would provide an acceptable level of quality and safety, and should be authorized under 10 CFR 50.55a(a)(3)(i).
The NRC staff notes that the requirements of Code Case N-729-1 require either VOlumetric and/or surface examination of the required zone. The licensee is requesting that an alternate, reduced examination zone be defined based on the region that can be examined using UT alone, and does not propose using surface examination of the remaining examination zone required by the Code Case. The staff finds that a reduced UT examination zone of penetration nozzles below the J-groove weld as the sole basis for relief does not support a determination that the proposed alternative would provide an acceptable level of quality and safety.
- 6 The licensee previously performed UT examinations of these CEDM nozzles at PVNGS, in response to NRC Bulletin (BL) 2001-01, "Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles," dated August 3, 2001; BL 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity," dated March 18, 2002; and BL 2002-02, "Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle Inspection Programs," dated August 9,2002; and the First Revised Order, EA-03-009 (Order). In a previous relief request dated March 19,2004 (ADAMS Accession No. ML04168293), the licensee stated that experience gained from those examinations has shown that scanning becomes ineffective from slightly above the top of the nozzle's chamfer face to the bottom of the nozzle, and that the licensee had also assessed the other examination options in the Order.
These options are similar to those in Code Case N-729-1. Option IV.C(5)(b)(ii) of the Order, which allows surface examination of the wetted surface in question, would result in significant radiological exposure to personnel if surface examination were conducted on surfaces of all CEDM nozzles. The high exposures, at least 30 times the dose of the proposed alternative, would be the result of the manual processes needed to perform surface examinations on the outside diameter of the CEDM nozzles. In a supplemental licensee letter dated April 16, 2004 (ADAMS Accession No. ML041680367), the licensee provided a more in-depth analysis of the dose estimate. The total dose to perform the manual surface examination of all CEDM nozzles is estimated by the licensee to be 48.5 man-rem.
The NRC staff concludes that, consistent with previous relaxation requests from the requirements of the Order, the combination of radiological dose considerations, and employment of deterministic fracture mechanics methodology and crack growth analysis may provide a basis to show that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Therefore, the staff has reviewed the licensee's request in accordance with 10 CFR 50.55a(a)(3)(ii).
The NRC staff has examined the design of the CEDM nozzles and concludes that that the funnel threaded connection with a weld plug, and the as-welded condition of nozzle partial penetration J-groove welds prevents UT examination to the distance "a" below the lowest point of the toe of the J-groove weld for all CEDM nozzles. The staff further concludes that a manual surface examination of the zone between that zone which is accessible with LIT examination and the distance "a" below the J-groove weld toe for the wetted surface of all of the CEDM nozzles would present a significant hardship due to radiological exposure.
The licensee evaluated each CEDM nozzle to determine the distance below the J-groove weld which is amenable to UT examination. A deterministic fracture mechanics analysis was performed for each of the nozzle geometries to predict the time before a crack outside the UT accessible zone would take to propagate to the toe of the J-groove weld. The methodology and calculations for Units 1 and 2 are detailed in Westinghouse report WCAP-15817, Revision 1, "Structural Integrity Evaluation of Reactor Upper Head Penetrations to Support Continued Operation: Palo Verde Units 1 and 2," dated October 2003 (ADAMS Accession No. ML041800350), and licensee letters dated July 1, 2004 (ADAMS Accession No. ML041950341),
and November 24,2004 (ADAMS Accession No. ML043420167). The methodology and calculations for Unit 3 are detailed in WCAP-16044, "Structural Integrity Evaluation of Reactor Upper Head Penetrations to Support Continued Operation: Palo Verde Unit 3," dated June 2003 (ADAMS Accession No. ML043020287).
- 7 The NRC staff has performed an engineering evaluation of the analysis methodology employed in these WCAP reports and supporting licensee letters. The analysis methodology of the predicted time for a hypothetical crack outside the examination zone to propagate to the toe of the J-groove weld is comprised of two parts: a finite element (FE) analysis of the stress distribution near the J-groove weld and an analysis of propagation of a hypothetical crack outside the examination region.
The NRC staff has reviewed the parameters incorporated into the FE model contained in the WCAP reports. The FE model considered the lower portion of the CEDM penetration nozzle, the adjacent section of the vessel closure head, and the joining weld, and performed calculations for the different nozzle-to-head incidence angles. The residual stress of the vessel to penetration nozzle J-groove weld was simulated with two welding passes, the material parameters of the penetration nozzle, the weld metal and cladding were modeled as Alloy 600 and the vessel head shell as carbon steel. The only loads used in the analysis were the steady state operating loads since external loads, such as seismic, were expected to be captured by the full thickness of the reactor vessel head. Except as noted in Tables 1 and 2, the as-designed weld size was used in the FE model. The staff concludes that the methodology employed is appropriate, and that the resulting stress distributions are reasonable and qualitatively consistent with the analysis performed in Electric Power Research Institute (EPRI) report, "Materials Reliability Program Generic Evaluation of Examination Coverage Requirements for Reactor Pressure Vessel Head Penetration Nozzles, Revision 1 (MRP-95R1 )," dated September 2004 (ADAMS Accession No. ML043200634).
The NRC staff notes that the analysis for Unit 1 nozzles 84, 87, and 93 (Note 1 in Table 1) as well as Unit 2 nozzle 56 (Note 2 in Table 2) used the as-built weld dimensions of each individual nozzle in the FE calculations rather than the as-designed weld dimensions. Since the actual dimensions of the weld were obtained from previous UT examinations, use of the as-built dimensions in the FE calculations is expected to represent the actual stress state more accurately than that calculated using the as-designed dimensions. Thus, the staff concludes that their use is acceptable.
The second part of the analysis, the crack growth calculations, was performed assuming that a hypothetical through wall axial flaw existed with its upper extremity at the alternate examination zone boundary and its lower extremity outside the alternate examination zone at the point where the FE model determined that the stresses became compressive. The average of the inside and outside hoop stresses were applied over the entire length of the assumed crack and the stress intensity factor for an axial through wall crack in a cylinder was used. The crack growth model equation used had an identical form to that in EPRI report, "Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Material (MRP-55)," dated July 18, 2002 (ADAMS Accession No. ML023010498), as well as that found in ASME Code,Section XI, Appendix 0, 2004 Edition.
The coefficients used in the equation of the WCAP analysis were the same as those in MRP-55 and Appendix 0 except the crack growth amplitude coefficient, 0, used in the proprietary WCAP reports was more conservative than that in either MRP-55 or Appendix O. Therefore, the staff concludes that the use of the WCAP crack growth analysis is acceptable.
- 8 The NRC staff further notes that the use and methodology of both of these WCAP reports, as well as the calculations detailed in supplemental licensee letters, have been accepted previously in a safety evaluation dated May 5, 2004 (ADAMS Accession No. ML041260228), of the relaxation request for similar inspection requirements in the Order for PVNGS, Units 1, 2, and 3.
The data in Tables 1, 2, and 3 in Section 3.2 of this Safety Evaluation in the columns "Minimum Inspection Coverage Required Below the J-Groove Weld Toe on the Downhill Side (in)" specify the distance below the toe of the J-groove weld to be inspected by UT examination or, where UT examination cannot examine the entire specified distance below the J-groove weld, by a combination of UT and surface examination. The WCAP reports and supporting documentation noted in the tables predict the time of operation before a crack tip located at the boundary of the alternate examination zone below the J-groove weld can propagate to the toe of the J-groove weld. These values are shown in columns "EFPY for the Upper Crack Tip to Reach the J-Groove Weld Toe." The entries "No Propagation Predicted" represent an infinite lifetime prediction. The minimum time found in the tables for all three units is 1.7 effective full power years (EFPY). Since the minimum time prediction is longer than the expected time between inspections of approximately 1.4 EFPY, the adequacy of the alternate examination volume or surface below the J-groove weld has been demonstrated. The NRC staff concludes that the use of the data in Tables 1, 2, and 3 of this Safety Evaluation to define an alternate examination zone below the lowest point at the toe of the J-groove weld provides a reasonable assurance of the structural integrity of the CEDM nozzles.
The NRC staff further concludes that a manual surface examination of the zone between that which is accessible with UT examination and the distance "a" below the J-groove weld toe, as defined in Figure 2 of Code Case N-729-1, for all CEDM nozzles would present a significant hardship due to radiological exposure without a compensating increase in the level of quality and safety. Based on the above, the NRC staff concludes that the proposed request for relief is acceptable.
4.0 CONCLUSION
The analyses submitted in support of APS Relief Request No. 41, "Request to Use Appendix I of ASME Code Case N-729-1," to define an alternate examination zone below the J-groove weld toe provide reasonable assurance of the structural integrity of the CEDM nozzles.
Furthermore, a manual surface examination of the zone between the area accessible with UT examination and the distance "a" below the J-groove weld toe, as defined in Figure 2 of Code Case N-729-1, for all CEDM nozzles, would present a significant hardship due to radiological exposure without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the use of the proposed alternative to define an alternate examination zone below the J-groove weld toe is authorized for the PVNGS, Units 1, 2, and 3, for the third 1O-year inservice inspection interval for each unit, or until that unit's reactor vessel head is replaced, whichever occurs first.
- 9 All other requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D)(6) not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Jay Wallace Date: Jul y 31, 2009
R. Edington
- 2 A copy of the related Safety Evaluation is enclosed. If you have any questions, please contact Randy Hall at (301) 415-4032 or via email at randy.hall@nrc.gov.
Sincerely, IRN Michael 1. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN-528, STN-529, and STN 50-530
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:
PUBLIC RidsNrrPMPaloVerde Resource LPLIV rtf RidsNrrLAJBurkhardt Resource RidsAcrsAcnw_MailCTR Resource RidsOgcRp Resource RidsNrrDciCpnb Resource RidsRgn4MailCenter Resource RidsNrrDorlDpr Resource JWaliace, NRRtDCltCPNB RidsNrrDorlLpl4 Resource LTrocine, EDO RIV ADAMS Accession No.: ML092080131
(*) Concurrence via SE
(**) See previous concurrence OFFICE NRRlLPL4/PM NRRlLPL4/LA DCI/CPNB/BC NRRlLPL4/BC NAME JRHall**
JBurkhardt**
TChan (*)
MMarkley DATE 7/24/09 7/29/09 6/09/09 7/31/09 OFFICIAL AGENCY RECORD