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{{#Wiki_filter:' En | {{#Wiki_filter:' En tergy Entergy Nuclear Operations, Inc.600 Rocky Hill Road Plymouth, MA 02360 Pilgrim Nuclear Power Station LETTER NUMBER 2.14.028 March 12, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 | ||
==SUBJECT:== | ==SUBJECT:== | ||
Request for Alternative | Request for Alternative | ||
-Alternative Examination Requirements for | -Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 | ||
==REFERENCES:== | ==REFERENCES:== | ||
: 1. ASME Boiler and Pressure Vessel Code, Code Case N-702,"Alternative Requirements for Boiling Water Reactor (BWR) | : 1. ASME Boiler and Pressure Vessel Code, Code Case N-702,"Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radii and Nozzle-to-Vessel Shell Welds, Section Xl, Division 1", dated February 20, 2004 2. BWRVIP-241, "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to Shell Welds and Nozzle Blend Radii", dated October 2010 3. NRC Letter to Entergy, "Relief Request PRR-20, Alternative Examination Requirements for Nozzle-to-Shell and Inner Radii Welds Using ASME Code Case N-702 and BWRVIP-108 | ||
-Pilgrim | -Pilgrim Nuclear Power Station (TAC NO. ME3290)", dated August 25, 2010 (1.10.035) | ||
dated August 25, 2010 (1.10.035) | : 4. NRC Letter to BWRVIP, "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP)-241 Report, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to Shell Welds and Nozzle Blend Radii (TAC NO. ME6328)", dated April 19, 2013 | ||
: 4. NRC Letter to BWRVIP, "Final Safety Evaluations of the Boiling | |||
dated | |||
==Dear Sir or Madam:== | ==Dear Sir or Madam:== | ||
Pursuant to 10 CFR 50.55a(a)(3)(i), | Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy Nuclear Operations, Inc. (Entergy) requests NRC authorization to implement alternative examination requirements based on the American Society of Mechanical Engineers (ASME) Code Case N-702 (Reference | ||
Entergy Nuclear Operations, Inc. (Entergy) requests | : 1) and Boiling Water Reactor Vessel Inspection Program (BWRVIP) -241 (Reference | ||
: 1) and Boiling | : 2) as documented in the enclosed Pilgrim Nuclear Power Station (PNPS) Relief Request (PRR) -24, "Alternative Examination Requirements for PNPS Nozzle-to Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241". | ||
-241 (Reference | |||
: 2) as documented in | |||
This template format is similar to the submittal the Nuclear Regulatory Commission (NRC) Staff has approved in a Safety Evaluation (Reference | This template format is similar to the submittal the Nuclear Regulatory Commission (NRC) Staff has approved in a Safety Evaluation (Reference | ||
: 3) for other PNPS Nozzle-to-Vessel Welds and Nozzle Inner Radii in PRR-20. | : 3) for other PNPS Nozzle-to-Vessel Welds and Nozzle Inner Radii in PRR-20. | ||
LETTER NUMBER 2.14. | LETTER NUMBER 2.14.028 Page 2 of 2 In accordance with 10 CFR 50.55a(a)(3)(i), the proposed alternative to the referenced requirements may be approved by the NRC provided an acceptable level of quality and safety are maintained. | ||
the proposed alternative to the referenced requirements may be approved by the NRC provided an acceptable level of quality and | Entergy believes the proposed alternative provides an acceptable level of quality and safety as well as a significant reduction in worker radiation exposure. | ||
Entergy believes the proposed alternative provides an acceptable level | The NRC provided a Safety Evaluation (Reference | ||
The | : 4) approving the generic technical bases and acceptability criteria for application of Code Case N-702 and BWRVIP-241, which Entergy has followed as detailed in the enclosure. | ||
: 4) approving the generic technical bases | Entergy requests approval of the proposed alternative on or before April 1, 2015 to accommodate application of this request during the next refueling outage. Entergy plans to implement this alternative during the third period of the fourth ISI interval. | ||
Entergy requests approval of the proposed alternative on or before April 1, 2015 | Although this review is neither exigent nor emergency, your prompt review is requested. | ||
Although | |||
There are no new commitments made in this submittal. | There are no new commitments made in this submittal. | ||
If you have any questions or | If you have any questions or require additional information, please contact me at (508) 830-8403.Sincerely, Joseph R. Lynch, Regulatory Assurance Manager JRL/mew | ||
Sincerely, Joseph R. Lynch, Regulatory Assurance | |||
==Enclosure:== | ==Enclosure:== | ||
Pilgrim Relief Request (PRR) -24, "Alternative Examination Requirements | Pilgrim Relief Request (PRR) -24, "Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 in Accordance with 10CFR 50.55a (a)(3)(i)" cc: Mr. William M. Dean Regional Administrator, Region 1-U. S. Nuclear Regulatory Commission 2100 Renaissance Boulevard, Suite 100 King of Prussia, PA 19406-1415 Ms. Nadiyah Morgan, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission MS O-8C2A Washington, DC 20555 NRC Senior Resident Inspector Pilgrim Nuclear Power Station Enclosure to Letter 2.14.028 Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" Sheet 1 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" 1. ASME Code Component(s) | ||
cc: Mr. William M. | Affected Code Class: 1 Component Numbers: N2 (See Attachment 1 for detailed list of components) | ||
Sheet 1 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" | Applicable Code Edition: ASME Section Xl, 1998 Edition with 2000 Addenda Examination Category: | ||
B-D (Inspection Program B)Item Numbers: B3.90 and B3.100 | |||
N2 (See Attachment 1 for detailed list of components) | == Description:== | ||
Applicable Code Edition: | |||
ASME Section Xl, 1998 Edition with 2000 | Alternative to ASME Section Xl, Table IWB-2500-1 Unit/Inspection Interval Pilgrim (PNPS) /Fourth (4th)10-year interval starting Applicability: | ||
B-D (Inspection Program B)Item Numbers: | July 1, 2005, ending June 30, 2015 2. Applicable Code Requirement ASME Section Xl, 1998 Edition with 2000 Addenda Table IWB-2500-1, Examination Category B-D, Full Penetration Welded Nozzles In Vessels -Inspection Program B requires a volumetric examination of all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles each 10-year interval. | ||
B3.90 and B3. | Additionally, for ultrasonic examinations, ASME Section Xl, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is implemented, as required and modified by 10 CFR 50.55a(b)(2)(xv). | ||
The subject components for this request for alternative are the N2 Recirculation Inlet Reactor Vessel Nozzle-to-Vessel Welds (Item B3.90) and the Reactor Vessel Nozzle Inner Radius Sections (Item B3.100).3. Reason for Request The twenty-five percent sampling level stated in Code Case N-702 (Reference 3)provides a significant cost savings and reduction in worker dose exposure. | |||
July 1, 2005, ending June 30, | PNPS has estimated that the proposed reduction of inspection requirements would result in a cost savings of approximately | ||
Additionally, | $200,000 and reduction in worker radiation exposure of approximately 3.7 Person-Rem over the remainder of the current interval while providing an acceptable level of quality and safety.Enclosure to Letter 2.14.028 Sheet 2 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" 4. Proposed Alternative Pursuant to 10 CFR 50.55a (a)(3)(i), an alternative is requested from performing the required examinations on 100% of the identified nozzle assemblies listed in Attachment | ||
is implemented, as required and modified by 10 | : 1. As an alternative, incorporation of Code Case N-702 (Reference 3)would require examination of a minimum of 25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each system and nominal pipe size as shown in Table 4-1. For three of the Reactor Recirculation nozzle assemblies shown in Table 4-1, both the inner radius region and the nozzle-to-shell weld have already been examined during the 4 th interval.Table 4-1 PNPS Summary -Affected Components of PRR-24 Minimum Number Nozzle Total to be Group Description Number Examined Comments Recirculation 3 completed in Inlet RFO17 (2009)5. Basis for Proposed Alternative Electric Power Research Institute (EPRI) Technical Report 1021005, "BWRVIP-241: | ||
The subject components for this request for alternative are the | BWR Vessel and Internals Project (BWRVIP), Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii" (Reference | ||
: 6) provides the technical basis for use of Code Case N-702. BWRVIP-241 was developed to propose a relaxation of the criteria in BWRVIP-108 allowing BWR's to obtain inspection relief for their Reactor Recirculation inlet and outlet nozzles. The evaluation found that failure probabilities due to a low temperature overpressure event at the nozzle blend radius region and nozzle-to-vessel shell weld are very low (i.e. < lX 10.6 for 40 years) with or without inservice inspection. | |||
$200,000 and reduction in worker radiation exposure of approximately 3.7 Person-Rem over the remainder of the | The report concludes that inspection of 25% of each nozzle type is technically justified. | ||
BWRVIP-108 was originally submitted to the NRC for review and approval by the BWRVIP via BWRVIP Letter 2002-323 on November 25, 2002 (Reference | |||
an alternative is requested from performing | : 2) and supplemented by Tennessee Valley Authority (TVA) letter dated November 15, 2004, and BWRVIP letters dated July 25, 2006, and September 13, 2007. Reference 4 provided the NRC Safety Evaluation approving use of ASME Code Case N-702 in accordance with BWRVIP-108 guidance.On April 19, 2013, the NRC issued a Safety Evaluation (SE) (Reference | ||
: 1. As an alternative, incorporation of Code Case N-702 (Reference 3)would require examination of a minimum of 25% of the nozzle-to-vessel welds | : 7) approving the use of BWRVIP-241. | ||
Table 4- | Within Section 5 of the SE, it states that each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the request for alternative by meeting the criteria discussed in Section 5 of the SE.Enclosure to Letter 2.14.028 Sheet 3 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with 1OCFR 50.55a (a)(3)(i)" The applicability of the BWRVIP-241 report to PNPS is demonstrated by showing the criteria within Section 5 of the SE are met for the recirculation inlet nozzles.The general terms used in the SE Section 5 applicability evaluations are: Ci.RPV = recirculation inlet nozzles (from BWRVIP-108 model) = 19332 psi CI-NOZZLE | ||
= recirculation inlet nozzles (from BWRVIP- 108 model) = 1637 psi The PNPS nozzle-specific terms to be used in the SER Section 5 applicability evaluations are as follows: Heatup / Cooldown rate = 100°F/hr p= Reactor Pressure Vessel (RPV) normal operating pressure, p = 1035 psi r= RPV inner radius, r = 113.40625" t= RPV wall thickness, t = 6.5" riN2 = inner radius for Recirculation Inlet N2 nozzles, rjN2 = 5.75" roN2 -outer radius for Recirculation Inlet N2 nozzles, roN2 = 9.125" The results of the equations in Attachment 2 demonstrate the applicability of the BWRVIP-108 1 BWRVIP-241 reports to PNPS by showing the criteria within Section 5.0 of the NRC SE is met for all nozzles listed in Attachment 1, Table 1. Therefore, the basis for using Code Case N-702 is demonstrated for the PNPS reactor recirculation inlet nozzles in Table 1.In addition, Code Case N-702 stipulates that a VT-1 examination may be used in lieu of the volumetric examination for the inner radii (Item No. B3.100). Note that PNPS is not currently using Code Case N-648-1 and has no plans of using the code case in the future.All examinations on Item B3.100, inner radius section, will be volumetric examinations. | |||
Probabilistic Fracture Mechanics Evaluation | The analyses for the N2 nozzles in BWRVIP-108 and BWRVIP-241 are based on the assumption that fluence at the nozzles is negligible. | ||
: 6) provides the technical basis for use of Code Case N-702. BWRVIP-241 was developed to propose a relaxation of the criteria in BWRVIP-108 allowing BWR's | The PNPS N2 nozzles are predicted to experience fluence in excess of 1 x 1017 by the end of the currently licensed period of extended operation. | ||
The report concludes that inspection | |||
BWRVIP-108 was originally submitted to the NRC for review and approval by the | |||
: 2) and supplemented | |||
On April 19, 2013, the NRC issued a Safety Evaluation (SE) (Reference | |||
: 7) approving | |||
Within Section 5 of the SE, it states that each licensee | |||
The applicability of the BWRVIP-241 report to PNPS is demonstrated by showing | |||
= recirculation inlet nozzles (from BWRVIP- 108 model) = 1637 | |||
Note that PNPS is | |||
The analyses for the N2 nozzles in BWRVIP-108 and BWRVIP-241 are based on | |||
The PNPS N2 nozzles are predicted to experience fluence in excess of 1 x 1017 by the end of the currently licensed period | |||
Therefore, a plant specific probabilistic fracture mechanics evaluation (Reference | Therefore, a plant specific probabilistic fracture mechanics evaluation (Reference | ||
: 8) was performed to supplement the criteria of Code Case N-702 | : 8) was performed to supplement the criteria of Code Case N-702 and BWRVIP-241 in order to demonstrate that the probability of failure remains acceptable over this period of extended operation (to 60 years). This analysis was performed using the same methods as were used in BWRVIP-241, with PNPS specific fracture mechanics analyses. | ||
The results demonstrate that for the N2 nozzles at PNPS, the probability | The results demonstrate that for the N2 nozzles at PNPS, the probability of failure was less than the limiting probability as defined in 10 CFR 50 Appendix H.Therefore, the probabilistic fracture mechanics criteria of BWRVIP-241 remain applicable to the PNPS N2 nozzles.Enclosure to Letter 2.14.028 Sheet 4 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" The analyses in BWRVIP-108 and BWRVIP-241 were based on predicted fatigue crack growth over the initial licensed operating period and assumed additional fatigue cycles in evaluating fatigue crack growth. PNPS is projected to exceed the total number of thermal cycles used in the BWRVIP analysis over its remaining life. However, the usage factor for the N2 nozzles remains below 1.0. Previous BWRVIP documents have demonstrated that fatigue crack growth is not a significant contributor to probability of failure for the N2 nozzles. Consideration of fatigue crack growth for the PNPS N2 nozzles is included in the plant specific probabilistic fracture mechanics analysis. | ||
The analyses in BWRVIP-108 and BWRVIP-241 were based on predicted fatigue | The plant specific analysis confirms that the additional fatigue cycles will have an insignificant effect on probability of failure for the PNPS N2 nozzles.In conclusion, use of ASME Code Case N-702 provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i) for Reactor Recirculation Inlet nozzle-to vessel shell full penetration welds and nozzle inner radii sections identified in Attachment 1, Table 1.6. Duration of Proposed Alternative Upon approval by the NRC staff, this request for alternative will be utilized through the remainder of Pilgrim's fourth inspection interval (July 1, 2005 -June 30, 2015) for the nozzle assemblies listed in Attachment 1, Table 1.7. Precedents None. However, since the NRC issued the SER for BWRVIP-241 in April of 2013, another plant has initiated a Relief Request to use the alternate criteria of Code Case N-702.8. References (1) EPRI Technical Report 1003557, "BWRVIP-108: | ||
Consideration of fatigue crack growth for the PNPS N2 nozzles is included in | BWR Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," dated October 2002 (2) BWRVIP letter 2002-323, Carl Terry, BWRVIP Chairman, to NRC Document Control Desk, "Project No. 704- BWRVIP-108: | ||
The plant specific | BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," November 25, 2002 (3) ASME Boiler and Pressure Vessel Code, Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section Xl, Division 1," dated February 20, 2004 (4) NRC Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," dated December 19, 2007 (5) NRC SE for Pilgrim Relief Request PRR-20, Alternative Examination Requirements For Nozzle to Shell and Inner Radii Welds Using ASME Code Case N-702 and BWRVIP-108 (TAC NO. ME3290) dated August 25, 2010 (Letter No: 1.10.035)Enclosure to Letter 2.14.028 Sheet 5 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" (6) EPRI Technical Report 1021005, "BWRVIP-241, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," dated October 2010 (7) NRC SE of the Boiling Water Reactor Vessel Internals Project (BWRVIP) -241 Report, Probabilistic Fracture Mechanics Evaluation for the Boiling-Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii (TAC NO. ME6328) dated April 19, 2013" (8) Structural Integrity Associates, "Evaluation of the Probability of Failure for Recirculation Inlet (N2) in the Nozzle-to-Shell-Welds and Nozzle Blend Radii Regions at Pilgrim Nuclear Station", 1400071.301 Revision 0, February 2014 Enclosure to Letter 2.14.028 Sheet 6 of 8 Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with 1OCFR 50.55a (a)(3)(i)" Attachment 1 Table I Table of ASME Code Components Affected at PNPS Code Component ID Description Category Code Item RPV-N2A-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2A-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2B-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2B-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2C-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2C-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2D-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2D-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2E-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2E-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2F-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2F-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2G-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2G-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2H-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2H-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2J-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2J-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2K-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2K-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 Enclosure to Letter 2.14.028 Sheet 7 of 8 Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" Attachment 2 Given the general and plant-specific terms, Pilgrim's conformance with the three (3)criteria applicable to the recirculation inlet nozzles is demonstrated as follows: (1) Max RPV Heatup/Cooldown Rate Criterion | ||
BWR Vessel and Internals | -the maximum RPV heatup / cooldown rate is limited to < 1150°F/hr In accordance with Technical Specification 3.6.A.2, Reactor Coolant System heatup and cooldown rates are procedurally limited to a maximum of 100'F when averaged over any one hour period and thus meets the requirement of criterion 1.(2) Recirculation Inlet (N2) Nozzles Equation to meet criterion: (prlt)/CRpv | ||
BWR Vessel and Internals | |||
Enclosure to Letter 2.14. | |||
(6) EPRI Technical Report 1021005, "BWRVIP-241, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and | |||
-241 Report,Probabilistic Fracture Mechanics Evaluation for the Boiling-Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii (TAC NO. ME6328) dated April 19, 2013"(8) Structural Integrity Associates, "Evaluation of the Probability of Failure | |||
1400071.301 Revision 0, February | |||
Attachment | |||
Attachment | |||
-the maximum RPV heatup / cooldown rate is limited to < 1150°F/hr In accordance with Technical Specification 3.6.A.2, Reactor Coolant System | |||
(prlt)/CRpv | |||
< 1.15[(1035)(113.40625)16.51/19332 | < 1.15[(1035)(113.40625)16.51/19332 | ||
= 0.93 < 1. | = 0.93 < 1.15 The PNPS result is 0.93 and thus meets the requirement of criterion 2 to be < 1.15.(3) Recirculation Inlet (N2) Nozzles Equation to meet criterion: (p(ro 2 + ri 2)l (ro1 -ri 2)/CN=ozzLE | ||
(p( | |||
< 1,47[1035(9.1252+5.752)/(9.1252-5.752)]11637 | < 1,47[1035(9.1252+5.752)/(9.1252-5.752)]11637 | ||
= 1.465 < 1. | = 1.465 < 1.47 The PNPS result is 1.465 and thus meets the requirement of criterion 3 to be < 1.47.Enclosure to Letter 2.14.028 Sheet 8 of 8}} | ||
Revision as of 20:23, 9 July 2018
| ML14077A175 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/12/2014 |
| From: | Lynch J R Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2.14.028 | |
| Download: ML14077A175 (10) | |
Text
{{#Wiki_filter:' En tergy Entergy Nuclear Operations, Inc.600 Rocky Hill Road Plymouth, MA 02360 Pilgrim Nuclear Power Station LETTER NUMBER 2.14.028 March 12, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Request for Alternative -Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35
REFERENCES:
- 1. ASME Boiler and Pressure Vessel Code, Code Case N-702,"Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radii and Nozzle-to-Vessel Shell Welds, Section Xl, Division 1", dated February 20, 2004 2. BWRVIP-241, "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to Shell Welds and Nozzle Blend Radii", dated October 2010 3. NRC Letter to Entergy, "Relief Request PRR-20, Alternative Examination Requirements for Nozzle-to-Shell and Inner Radii Welds Using ASME Code Case N-702 and BWRVIP-108
-Pilgrim Nuclear Power Station (TAC NO. ME3290)", dated August 25, 2010 (1.10.035)
- 4. NRC Letter to BWRVIP, "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP)-241 Report, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to Shell Welds and Nozzle Blend Radii (TAC NO. ME6328)", dated April 19, 2013
Dear Sir or Madam:
Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy Nuclear Operations, Inc. (Entergy) requests NRC authorization to implement alternative examination requirements based on the American Society of Mechanical Engineers (ASME) Code Case N-702 (Reference
- 1) and Boiling Water Reactor Vessel Inspection Program (BWRVIP) -241 (Reference
- 2) as documented in the enclosed Pilgrim Nuclear Power Station (PNPS) Relief Request (PRR) -24, "Alternative Examination Requirements for PNPS Nozzle-to Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241".
This template format is similar to the submittal the Nuclear Regulatory Commission (NRC) Staff has approved in a Safety Evaluation (Reference
- 3) for other PNPS Nozzle-to-Vessel Welds and Nozzle Inner Radii in PRR-20.
LETTER NUMBER 2.14.028 Page 2 of 2 In accordance with 10 CFR 50.55a(a)(3)(i), the proposed alternative to the referenced requirements may be approved by the NRC provided an acceptable level of quality and safety are maintained. Entergy believes the proposed alternative provides an acceptable level of quality and safety as well as a significant reduction in worker radiation exposure. The NRC provided a Safety Evaluation (Reference
- 4) approving the generic technical bases and acceptability criteria for application of Code Case N-702 and BWRVIP-241, which Entergy has followed as detailed in the enclosure.
Entergy requests approval of the proposed alternative on or before April 1, 2015 to accommodate application of this request during the next refueling outage. Entergy plans to implement this alternative during the third period of the fourth ISI interval. Although this review is neither exigent nor emergency, your prompt review is requested. There are no new commitments made in this submittal. If you have any questions or require additional information, please contact me at (508) 830-8403.Sincerely, Joseph R. Lynch, Regulatory Assurance Manager JRL/mew
Enclosure:
Pilgrim Relief Request (PRR) -24, "Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 in Accordance with 10CFR 50.55a (a)(3)(i)" cc: Mr. William M. Dean Regional Administrator, Region 1-U. S. Nuclear Regulatory Commission 2100 Renaissance Boulevard, Suite 100 King of Prussia, PA 19406-1415 Ms. Nadiyah Morgan, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission MS O-8C2A Washington, DC 20555 NRC Senior Resident Inspector Pilgrim Nuclear Power Station Enclosure to Letter 2.14.028 Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" Sheet 1 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" 1. ASME Code Component(s) Affected Code Class: 1 Component Numbers: N2 (See Attachment 1 for detailed list of components) Applicable Code Edition: ASME Section Xl, 1998 Edition with 2000 Addenda Examination Category: B-D (Inspection Program B)Item Numbers: B3.90 and B3.100
Description:
Alternative to ASME Section Xl, Table IWB-2500-1 Unit/Inspection Interval Pilgrim (PNPS) /Fourth (4th)10-year interval starting Applicability: July 1, 2005, ending June 30, 2015 2. Applicable Code Requirement ASME Section Xl, 1998 Edition with 2000 Addenda Table IWB-2500-1, Examination Category B-D, Full Penetration Welded Nozzles In Vessels -Inspection Program B requires a volumetric examination of all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles each 10-year interval. Additionally, for ultrasonic examinations, ASME Section Xl, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is implemented, as required and modified by 10 CFR 50.55a(b)(2)(xv). The subject components for this request for alternative are the N2 Recirculation Inlet Reactor Vessel Nozzle-to-Vessel Welds (Item B3.90) and the Reactor Vessel Nozzle Inner Radius Sections (Item B3.100).3. Reason for Request The twenty-five percent sampling level stated in Code Case N-702 (Reference 3)provides a significant cost savings and reduction in worker dose exposure. PNPS has estimated that the proposed reduction of inspection requirements would result in a cost savings of approximately $200,000 and reduction in worker radiation exposure of approximately 3.7 Person-Rem over the remainder of the current interval while providing an acceptable level of quality and safety.Enclosure to Letter 2.14.028 Sheet 2 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" 4. Proposed Alternative Pursuant to 10 CFR 50.55a (a)(3)(i), an alternative is requested from performing the required examinations on 100% of the identified nozzle assemblies listed in Attachment
- 1. As an alternative, incorporation of Code Case N-702 (Reference 3)would require examination of a minimum of 25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each system and nominal pipe size as shown in Table 4-1. For three of the Reactor Recirculation nozzle assemblies shown in Table 4-1, both the inner radius region and the nozzle-to-shell weld have already been examined during the 4 th interval.Table 4-1 PNPS Summary -Affected Components of PRR-24 Minimum Number Nozzle Total to be Group Description Number Examined Comments Recirculation 3 completed in Inlet RFO17 (2009)5. Basis for Proposed Alternative Electric Power Research Institute (EPRI) Technical Report 1021005, "BWRVIP-241:
BWR Vessel and Internals Project (BWRVIP), Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii" (Reference
- 6) provides the technical basis for use of Code Case N-702. BWRVIP-241 was developed to propose a relaxation of the criteria in BWRVIP-108 allowing BWR's to obtain inspection relief for their Reactor Recirculation inlet and outlet nozzles. The evaluation found that failure probabilities due to a low temperature overpressure event at the nozzle blend radius region and nozzle-to-vessel shell weld are very low (i.e. < lX 10.6 for 40 years) with or without inservice inspection.
The report concludes that inspection of 25% of each nozzle type is technically justified. BWRVIP-108 was originally submitted to the NRC for review and approval by the BWRVIP via BWRVIP Letter 2002-323 on November 25, 2002 (Reference
- 2) and supplemented by Tennessee Valley Authority (TVA) letter dated November 15, 2004, and BWRVIP letters dated July 25, 2006, and September 13, 2007. Reference 4 provided the NRC Safety Evaluation approving use of ASME Code Case N-702 in accordance with BWRVIP-108 guidance.On April 19, 2013, the NRC issued a Safety Evaluation (SE) (Reference
- 7) approving the use of BWRVIP-241.
Within Section 5 of the SE, it states that each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the request for alternative by meeting the criteria discussed in Section 5 of the SE.Enclosure to Letter 2.14.028 Sheet 3 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with 1OCFR 50.55a (a)(3)(i)" The applicability of the BWRVIP-241 report to PNPS is demonstrated by showing the criteria within Section 5 of the SE are met for the recirculation inlet nozzles.The general terms used in the SE Section 5 applicability evaluations are: Ci.RPV = recirculation inlet nozzles (from BWRVIP-108 model) = 19332 psi CI-NOZZLE = recirculation inlet nozzles (from BWRVIP- 108 model) = 1637 psi The PNPS nozzle-specific terms to be used in the SER Section 5 applicability evaluations are as follows: Heatup / Cooldown rate = 100°F/hr p= Reactor Pressure Vessel (RPV) normal operating pressure, p = 1035 psi r= RPV inner radius, r = 113.40625" t= RPV wall thickness, t = 6.5" riN2 = inner radius for Recirculation Inlet N2 nozzles, rjN2 = 5.75" roN2 -outer radius for Recirculation Inlet N2 nozzles, roN2 = 9.125" The results of the equations in Attachment 2 demonstrate the applicability of the BWRVIP-108 1 BWRVIP-241 reports to PNPS by showing the criteria within Section 5.0 of the NRC SE is met for all nozzles listed in Attachment 1, Table 1. Therefore, the basis for using Code Case N-702 is demonstrated for the PNPS reactor recirculation inlet nozzles in Table 1.In addition, Code Case N-702 stipulates that a VT-1 examination may be used in lieu of the volumetric examination for the inner radii (Item No. B3.100). Note that PNPS is not currently using Code Case N-648-1 and has no plans of using the code case in the future.All examinations on Item B3.100, inner radius section, will be volumetric examinations. The analyses for the N2 nozzles in BWRVIP-108 and BWRVIP-241 are based on the assumption that fluence at the nozzles is negligible. The PNPS N2 nozzles are predicted to experience fluence in excess of 1 x 1017 by the end of the currently licensed period of extended operation. Therefore, a plant specific probabilistic fracture mechanics evaluation (Reference
- 8) was performed to supplement the criteria of Code Case N-702 and BWRVIP-241 in order to demonstrate that the probability of failure remains acceptable over this period of extended operation (to 60 years). This analysis was performed using the same methods as were used in BWRVIP-241, with PNPS specific fracture mechanics analyses.
The results demonstrate that for the N2 nozzles at PNPS, the probability of failure was less than the limiting probability as defined in 10 CFR 50 Appendix H.Therefore, the probabilistic fracture mechanics criteria of BWRVIP-241 remain applicable to the PNPS N2 nozzles.Enclosure to Letter 2.14.028 Sheet 4 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" The analyses in BWRVIP-108 and BWRVIP-241 were based on predicted fatigue crack growth over the initial licensed operating period and assumed additional fatigue cycles in evaluating fatigue crack growth. PNPS is projected to exceed the total number of thermal cycles used in the BWRVIP analysis over its remaining life. However, the usage factor for the N2 nozzles remains below 1.0. Previous BWRVIP documents have demonstrated that fatigue crack growth is not a significant contributor to probability of failure for the N2 nozzles. Consideration of fatigue crack growth for the PNPS N2 nozzles is included in the plant specific probabilistic fracture mechanics analysis. The plant specific analysis confirms that the additional fatigue cycles will have an insignificant effect on probability of failure for the PNPS N2 nozzles.In conclusion, use of ASME Code Case N-702 provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i) for Reactor Recirculation Inlet nozzle-to vessel shell full penetration welds and nozzle inner radii sections identified in Attachment 1, Table 1.6. Duration of Proposed Alternative Upon approval by the NRC staff, this request for alternative will be utilized through the remainder of Pilgrim's fourth inspection interval (July 1, 2005 -June 30, 2015) for the nozzle assemblies listed in Attachment 1, Table 1.7. Precedents None. However, since the NRC issued the SER for BWRVIP-241 in April of 2013, another plant has initiated a Relief Request to use the alternate criteria of Code Case N-702.8. References (1) EPRI Technical Report 1003557, "BWRVIP-108: BWR Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," dated October 2002 (2) BWRVIP letter 2002-323, Carl Terry, BWRVIP Chairman, to NRC Document Control Desk, "Project No. 704- BWRVIP-108: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," November 25, 2002 (3) ASME Boiler and Pressure Vessel Code, Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section Xl, Division 1," dated February 20, 2004 (4) NRC Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," dated December 19, 2007 (5) NRC SE for Pilgrim Relief Request PRR-20, Alternative Examination Requirements For Nozzle to Shell and Inner Radii Welds Using ASME Code Case N-702 and BWRVIP-108 (TAC NO. ME3290) dated August 25, 2010 (Letter No: 1.10.035)Enclosure to Letter 2.14.028 Sheet 5 of 8 Entergy Nuclear Operations, Inc.Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" (6) EPRI Technical Report 1021005, "BWRVIP-241, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," dated October 2010 (7) NRC SE of the Boiling Water Reactor Vessel Internals Project (BWRVIP) -241 Report, Probabilistic Fracture Mechanics Evaluation for the Boiling-Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii (TAC NO. ME6328) dated April 19, 2013" (8) Structural Integrity Associates, "Evaluation of the Probability of Failure for Recirculation Inlet (N2) in the Nozzle-to-Shell-Welds and Nozzle Blend Radii Regions at Pilgrim Nuclear Station", 1400071.301 Revision 0, February 2014 Enclosure to Letter 2.14.028 Sheet 6 of 8 Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with 1OCFR 50.55a (a)(3)(i)" Attachment 1 Table I Table of ASME Code Components Affected at PNPS Code Component ID Description Category Code Item RPV-N2A-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2A-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2B-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2B-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2C-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2C-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2D-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2D-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2E-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2E-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2F-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2F-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2G-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2G-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2H-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2H-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2J-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2J-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 RPV-N2K-NV 12" Recirculation Inlet Nozzle to Vessel Weld B-D B3.90 RPV-N2K-NIR 12" Recirculation Inlet Nozzle Inner Radius B-D B3.100 Enclosure to Letter 2.14.028 Sheet 7 of 8 Pilgrim Relief Request (PRR) -24"Alternative Examination Requirements for Pilgrim Nuclear Power Station Nozzle-to-Vessel Welds and Nozzle Inner Radii Using ASME Code Case N-702 and BWRVIP-241 In Accordance with IOCFR 50.55a (a)(3)(i)" Attachment 2 Given the general and plant-specific terms, Pilgrim's conformance with the three (3)criteria applicable to the recirculation inlet nozzles is demonstrated as follows: (1) Max RPV Heatup/Cooldown Rate Criterion -the maximum RPV heatup / cooldown rate is limited to < 1150°F/hr In accordance with Technical Specification 3.6.A.2, Reactor Coolant System heatup and cooldown rates are procedurally limited to a maximum of 100'F when averaged over any one hour period and thus meets the requirement of criterion 1.(2) Recirculation Inlet (N2) Nozzles Equation to meet criterion: (prlt)/CRpv < 1.15[(1035)(113.40625)16.51/19332 = 0.93 < 1.15 The PNPS result is 0.93 and thus meets the requirement of criterion 2 to be < 1.15.(3) Recirculation Inlet (N2) Nozzles Equation to meet criterion: (p(ro 2 + ri 2)l (ro1 -ri 2)/CN=ozzLE < 1,47[1035(9.1252+5.752)/(9.1252-5.752)]11637 = 1.465 < 1.47 The PNPS result is 1.465 and thus meets the requirement of criterion 3 to be < 1.47.Enclosure to Letter 2.14.028 Sheet 8 of 8}}