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{{Adams
#REDIRECT [[L-08-111, Perry, Inservice Inspection Program Relief Request IR-054]]
| number = ML080990625
| issue date = 03/31/2008
| title = Perry, Inservice Inspection Program Relief Request IR-054
| author name = Bezilla M B
| author affiliation = FirstEnergy Nuclear Operating Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000440
| license number = NPF-058
| contact person =
| case reference number = L-08-111
| document type = Letter
| page count = 6
| project =
| stage = Other
}}
 
=Text=
{{#Wiki_filter:-I FENOC Perry Nuclear Power Station"-"% 10 Center Road FirstEnergy Nuclear Operating Company Perry, Ohio 44081 Mark B. Bezilla 440-280-5382 Vice President Fax: 440-280-8029 March 31, 2008 L-08-.111 10 CFR 50.55a ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
 
==SUBJECT:==
Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Perry Nuclear Power Plant Inservice Inspection Program Relief Request IR-054 In accordance with 10 CFR 50.55a, Nuclear Regulatory Commission (NRC) review and approval of a request for relief from certain Inservice Inspection (ISI) requirements associated with the implementation of Section Xl of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for the Perry Nuclear Power Plant (PNPP) is requested.
Enclosure 1 contains the identification of the affected components, the applicable code requirements, the description and basis of the proposed relief request, and the proposed alternative for the relief request. The relief request is proposed for use during the remainder of the current PNPP 1 0-year ISI interval.
Approval of the relief request is requested by February 20, 2009, to support PNPP.'s twelfth refueling outage.There are no regulatory commitments contained in this submittal.
If there are any questions, or if additional information is required, please contact'Mr. Thomas A. Lentz, Manager -Fleet Licensing, at (330) 761-6071.Sincerely, Mark B.Bezilla
 
==Enclosure:==
 
Relief Request IR-54 cc: NRC Region III Administrator NRC Resident Inspector NRC Project Manager ,v,7 State of Ohio Enclosure 1 L-08-111 Page 1 of 5 Perry Nuclear Power Plant Unit 1 RELIEF REQUEST No. IR-054, Rev 0 Page 1 of 5 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)
: 1. Identification of Components ASME Class 1 Reactor Pressure Vessel pressure-retaining Nozzle-to-Vessel Welds and Nozzle Inner Radii. The inservice examination identifications and descriptions are as follows: ISI Exam ID 1B13-NIA-KA 1B13-N1A-IR 1B13-N1B-KA 1B13-N1B-IR 181 3-N2A-KA 1 B13-N2A-IR 1B13-N2B-KA 181 3-N2B-IR 1B13-N2C-KA 1B13-N2C-IR 1B13-N2D-KA 1B13-N2D-IR 1B13-N2E-KA 1 B13-N2E-IR 1B13-N2F-KA 1B13-N2F-IR 1 B 13-N2G-KA 1B13-N2G-IR 1B13-N2H-KA 1 B13-N2H-IR 181 3-N2J-KA 1B13-N2J-IR 1B1 3-N2K-KA 1B13-N2K-IR 1 B1 3-N7-KA 11 B3-N7-IR 181 3-N8-KA 181 3-N8-IR 1 B13-N9A-KA 181 3-N9A-IR 1B1 3-N9B-KA 181 3-N9B-IR Description 22" Recirculation Outlet Nozzle NIA to Vessel Weld 22" Recirculation Outlet Nozzle N1A Inner Radius 22" Recirculation Outlet Nozzle N1B to Vessel Weld 22" Recirculation Outlet Nozzle N1B Inner Radius 12" Recirculation Inlet Nozzle N2A to Vessel Weld 12" Recirculation Inlet Nozzle N2A Inner Radius 12" Recirculation Inlet Nozzle N2B to Vessel Weld 12" Recirculation Inlet Nozzle N2B Inner Radius 12" Recirculation Inlet Nozzle N2C to Vessel Weld 12" Recirculation Inlet Nozzle N2C Inner Radius 12" Recirculation Inlet Nozzle N2D to Vessel Weld 12" Recirculation Inlet Nozzle N2D Inner Radius 12" Recirculation Inlet Nozzle N2E to Vessel Weld 12" Recirculation Inlet Nozzle N2E Inner Radius 12" Recirculation Inlet Nozzle N2F to Vessel Weld 12" Recirculation Inlet Nozzle N2F Inner Radius 12" Recirculation Inlet Nozzle N2G to Vessel Weld 12" Recirculation Inlet Nozzle N2G Inner Radius 12" Recirculation Inlet Nozzle N2H to Vessel Weld 12" Recirculation Inlet Nozzle N2H Inner Radius 12" Recirculation Inlet Nozzle N2J to Vessel Weld 12" Recirculation Inlet Nozzle N2J Inner Radius 12" Recirculation Inlet Nozzle N2K to Vessel Weld 12" Recirculation Inlet Nozzle N2K Inner Radius 6" Top Head Spray Spare Nozzle N7 to Vessel Weld 6" Top Head Spray Spare Nozzle N7 Inner Radius 6" Top Head Spray Nozzle N8 to Vessel Weld 6" Top Head Spray Nozzle N8 Inner Radius 4" Jet Pump Instrumentation Nozzle N9A to Vessel Weld 4" Jet Pump Instrumentation Nozzle N9A Inner Radius 4" Jet Pump Instrumentation Nozzle N9B to Vessel Weld 4" Jet Pump Instrumentation Nozzle N9B Inner Radius Enclosure 1 L-08-111 Page 2 of 5 2. Applicable Code Edition and Addenda Perry is currently in its second 10-year inspection interval and complies with the 1989 Edition of ASME XI. Additionally, for ultrasonic examinations, Section Xl, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 1995 Edition, with the 1996 Addenda, is implemented as required (and modified) by 10 CFR 50.55a.3. Applicable Code Requirement Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are given in Subsection IWB, Table IWB-2500-1, "Examination Category B-D Full Penetration Welds of Nozzles in Vessels -Inspection Program B," Item Numbers B3.90 and B3.100, respectively.
The method of examination is volumetric.
For the extent of examination, all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles must be examined each interval.4. Reason for Request Without approval of the requested relief to incorporate Code Case N-702, all of the Class 1 nozzle-to-vessel weld and nozzle inner radii examinations would be required prior to the end of Perry's current inspection interval (Refuel Outage 12, which is scheduled for February 2009). The proposed alternative provides an acceptable level of quality and safety and the reduction in scope will provide for a dose savings of at least 12,000 mRem.5. Proposed Alternative and Basis for Use.Proposed Alternative:
Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from performing the required examinations on 100% of the identified nozzle assemblies.
Alternatively, in accordance with Code Case N-702, a minimum of 25% of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size, would be performed.
For each of the identified nozzle assemblies, both the inner radius and the nozzle-to-shell weld would be examined.
The following nozzle assemblies would be selected for examination:
one of the two 22" recirculation outlet nozzle assemblies; three of the ten 12" recirculation inlet nozzle assemblies; one of the two 6" head spray nozzle assemblies; and one of the two 4" jet pump instrumentation nozzle assemblies.
Code Case N-702 proposes that VT-1 visual examination may be used in lieu of volumetric examination for the inner radii (Item B3.100). Note that Perry is already using Code Case N-648-1 in accordance with the conditions placed upon the use of that Code Case by Regulatory Guide 1.147, which allows VT-1 visual examination for nozzle inner radii. As Code Case N-648-1 is already approved for use at Perry, the Enclosure 1 L-08-1 11 Page 3 of 5 specific aspect of utilizing VT-1 visual examinations as allowed by Code Case N-702 is not a part of the request. Despite this allowance, volumetric examinations of the nozzle inner radii of the selected recirculation inlet and jet pump instrumentation nozzles will still be performed, as their nozzle inner radii are not fully accessible from inside the vessel for VT-1 examination.
Basis for Use: 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," provides the basis for Code Case N-702. 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., < 1 x 106 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 submitted to the NRC for review and approval by the Boiling Water Reactor Vessel and Internals Project (BWRVIP) via BWRVIP'letter 2002-323 on November 25, 2002.On December 19, 2007, the NRC issued a Safety Evaluation Report (SER) approving the use of BWRVIP-108 as a basis for using Code Case N-702. Within Section 5 of the SER, it states that each licensee should demonstrate the plant-specific applicability of the BWRVIP-1 08 report to their units in the relief request by meeting the criteria discussed in Section 5 of the SER.The applicability of the BWRVIP-1 08 report to the Perry plant is demonstrated by showing the criteria within Section 5 of the SER are met.The generic terms to be used in the SER Section 5 applicability evaluations are: Ci-RPV = recirculation inlet nozzles (from BWRVIP-108 model), Ci-RPV = 19332 psi C i-NOZZLE = recirculation inlet nozzles (from BWRVIP-108 model), C i-NOZZLE = 1637 psi Co-RPV = recirculation outlet nozzles (from BWRVIP-108 model), CO-aRPV= 16171 psi CO-NO77LE
= recirculation outlet nozzles (from BWRVIP-108 model), Co NOZZLE= 1977 psi The Perry-specific terms to be used in the SER Section 5 applicability evaluations are: Heatup/Cooldown rate = 1000 F/hr p = Reactor Pressure Vessel (RPV) normal operating pressure, p= 1045 psicq (maximum reactor steam dome pressure per Technical Specification 3.4.12)r = RPV inner radius, r = 119" t = RPV wall thickness, t = 7.19" Enclosure 1 L-08-111 Page 4 of 5 riN1 = inner radius for Recirculation Outlet N1 nozzles, riN1 10" roN1 = outer radius for Recirculation Outlet N1 nozzles, roN1 = 17.594" riN2 = inner radius for Recirculation Inlet N2 nozzles, riN2 = 5.813" roN2 -outer radius for Recirculation Inlet N2 nozzles, roN2 = 11.125" Given the generic and plant-specific terms, Perry's conformance with the five (5)criteria is demonstrated as follows: (1) Max RPV Heatup/Cooldown Rate Criterion
-the maximum RPV heatup/cooldown rate is limited to < 1150 F/hour In accordance with Technical Specification 3.4.11, Reactor Coolant System heatup and cooldown rates are < 1000 F in any one hour period.(2) Recirculation Inlet (N2) Nozzles Equation to meet criterion: (pr/t)/ Ci-RPV < 1.15[(1045 x 119) &#xf7;7.19] &#xf7; 19332 < 1.15 The Perry result is 0.89 < 1.15 (3) Recirculation Inlet (N2) Nozzles Equation to meet criterion:
[p(roN2 2 + riN2) -(roN2 2-riN2)] -Ci-NOZZLE
< 1.15[1045 x (11.1252 + 5.8132) &#xf7; (11.1252-5.8132)]
-1637 < 1.15 The Perry,result is 1.12 < 1.15 (4) Recirculation Outlet (N1) Nozzles Equation to meet criterion: (pr/t)/ Co-RPV < 1. 15[(1045 x 119) 7.19] &#xf7; 16171 < 1.15 The Perry result is 1.07 < 1.15 (5) Recirculation Outlet (N1) Nozzles Equation to meet criterion,:
[p(roN1 2+ riN1) &#xf7; (roN1 2-riN1)] -Co-NOZZLE
< 1.15[1045 x (17.5942 + 102) (17.5942-102)]
&#xf7;-1977 < 1.15 The Perry result is 1.03 < 1.15 Enclosure 1 L-08-111 Page 5 of 5 The results of the above equations demonstrate the applicability of the BWRVIP-108 report to the Perry plant by showing the criteria within Section 5 of the NRC SER is met. Therefore, the basis for using Code Case N-702 is demonstrated for the Perry plant.6. Duration of Proposed Alternative Upon approval by the NRC Staff, this relief request will be utilized through the remainder of Perry's second 10-Year inspection interval (November 18, 1998 through May 17, 2009, with the current 10-year inspection interval being extended by 6 months in accordance with IWA-2430(d)), or until the NRC publishes Code Case N-702 in a future revision of Regulatory Guide 1.147.7. 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," October 2002.2 ASME Boiler and Pressure Vessel Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plants," 1989 Edition with no Addenda.3 ASME Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plants," 1995 Edition with the 1996 Addenda.4. ASME Boiler and Pressure Vessel Code, Code Case N-648-1, "Alternative Requirements for Inner Radius Examinations of Class I Reactor Vessel Nozzles, Section Xl, Division 1," September 7, 2001.5. 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," February 20, 2004.6. 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, 2002.7. Matthew A. Mitchell, Office of Nuclear Reactor Regulation, to Rick Libra, BWRVIP Chairman, "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)'," December 19, 2007.}}

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