ML15278A024
| ML15278A024 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/28/2015 |
| From: | AREVA |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15278A022 | List:
|
| References | |
| 1CAN091503 ANP-3417NP, Rev. 0 | |
| Download: ML15278A024 (50) | |
Text
Attachment 2 to 1CAN091 503 ARE VA Document ANP-3417NP, Revision 0, "MRP-227-A Applicant / Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1,"
NON-PROPRIETARY
Go* *oThed Docum ent 1*
I AREVA ANP-3417NP M P-227-A ApplicntlLicensee Action Revision 0 Item #7 Analysis for Arkansas Nuclear One Unit I September 2015 AREVA Inc.
(c) 2015 AREVA Inc.
Cont~oltec ocment ANP-3417NP Revision 0 Copyright © 2015 AREVA Inc.
All Rights Reserved
Controlledc Document AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe i Nature of Changes Section(s)
Item or Page(s)
Description and Justification 1 All Initial Issue
Contl'ole Doc0umen AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe ii Contents Page
1.0 INTRODUCTION
AND PURPOSE................................................... 1-1 2.0 METHODOLOGY...................................................................... 2-1 2.1 WCAP-17096 Methodology Applicability .................................... 2-1 2.2 MRP-227-A Suggested Methodologies ...................................... 2-1 2.3 Methodology Utilized for ANO-1.............................................. 2-2 3.0 - CRGT SPACER CASTINGS ......................................................... 3-1 3.1 Background..................................................................... 3-1 3.1.1 Description of the Component Item .................................. 3-1 3.2 Evaluation Inputs .............................................................. 3-2 3.2.1 Flaw Size ............................................................... 3-3 3.2.2 Degraded Material Properties ........................................ 3-3 3.2.3 Distortion Evaluation................................................... 3-4 3.3 Evaluation ...................................................................... 3-4 3.3.1 Failure is Unlikely..................................... .................. 3-5 3.3.2 Effect of Failure on Functionality ..................................... 3-6 3.4 Conclusions..................................................................... 3-7 4.0 IMI GUIDE TUBE SPIDER CASTINGS. ............................................ 4-1 4.1 Background..................................................................... 4-1 4.1.1 Description of the Component Item ................................... 4-1 4.2 Evaluation Inputs .............................................................. 4-3 4.2.1 Likelihood of Fabrication and Service-Induced Flaws.............. 4-3 4.2.2 Driving Force for Crack Extension.................................... 4-3 4.2.3 Irradiated Fracture Toughness........................................ 4-4 4.3 Evaluation ...................................................................... 4-6 4.3.1 Likelihood of Failure ................................................... 4-7 4.3.2 Impact of Fractured Spider Casting on Functionality ............... 4-8 4.4 Conclusions .................................................................... 4-9 5.0 VENT VALVE RETAINING RINGS................................................... 5-1 5.1 Background..................................................................... 5-1 5.1.1 Description of the Component Item .................................. 5-1 5.2 Evaluation Inputs .............................................................. 5-2 5.2.1 Flaw Size ............................................................... 5-2 5.2.2 Degraded Material Properties ........................................ 5-3
Cont o C)
DoUmentsl AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A ApplicantlLicensee Action Item #7 Analysis for Arkansas Nuclear One Unit I Page iii 5.2.3 Stresses................................................................. 5-3 5.3 Evaluation....................................................................... 5-4 5.3.1 Failure is Unlikely ...................................................... 5-4 5.3.2 Effect of Failure on Functionality ..................................... 5-5 5.4 Conclusions .................................................................... 5-6 6.0 SELECT ORIGINAL VENT VALVE LOCKING DEVICE PARTS ................. 6-1 6.1 Background ..................................................................... 6-1 6.2 Evaluation....................................................................... 6-2 6.2.1 Degradation Mechanism .............................................. 6-2 6.2.2 Effect of Failure on Functionality...................................... 6-2 6.3 Conclusion...................................................................... 6-3 7.0 OVERALL CONCLUSIONS........................................................... 7-1
8.0 REFERENCES
......................................................................... 8-1
O(hontVold Docunent AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe iv Nomenclature Acronym Definition A/LAI Applicant/Licensee Action Item ANO Arkansas Nuclear One ANO-i Arkansas Nuclear One Unit 1 ASME American Society of Mechanical Engineers ASTM American Society of Testing and Materials BWR VIP Boiling Water Reactor Vessel and Internals Program CASS Cast Austenitic Stainless Steel CMTR Certified Material Test Report CRA Control Rod Assembly CRGT Control Rod Guide Tube CSS Core Support Shield EFPY Effective Full Power Year EPRI Electric Power Research Institute FIV Flow-Induced Vibration l&E Inspection and Evaluation IE Irradiation Embrittlement IMI Incore Monitoring Instrumentation LOCA Loss of Coolant Accident LR License Renewal MRP Materials Reliability Program NDE Non-Destructive Evaluation NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System PH Precipitation-Hardenable PT Dye Penetrant Testing PWR Pressurized Water Reactor RCP Reactor Coolant Pump RCS Reactor Coolant System RT Radiographic Testing RV Reactor Vessel SCC Stress Corrosion Cracking SER Safety Evaluation Report
Cont° olt *d Document AREVA Inc.
7N A Pa34e MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paoe V Acronym Definition SSE Safe Shutdown Earthquake TE Thermal Aging Embrittlement U.S. United States UT Ultrasonic Testing VT Visual Testing
GCon~o ed Document AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqie vi ABSTRACT The purpose of this document is to summarize the analyses performed for the applicable component items at ANO-1 to complete applicant/licensee action item #7 from MRP-227-A for ANO-1. The summary includes a discussion of the purpose, the methodology utilized, a summary of the background, evaluation inputs, evaluation, and conclusion for each component item, and an overall conclusion.
Conmooi Docum'~ent AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Panqe I-1
1.0 INTRODUCTION
AND PURPOSE The Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) developed inspection and evaluation (l&E) guidelines in document MRP-227-A (1) for managing long-term aging of pressurized water reactor (PWR) reactor vessel (RV) internal components. Specifically, the I&E guidelines are applicable to RV internal structural components; they do not address fuel assemblies, reactivity control assemblies, or welded attachments to the RV. The I&E guidelines concentrate on eight aging degradation mechanisms and their aging effects, such as loss of fracture toughness. The I&E guidelines define requirements for inspections that will allow owners of PWRs to demonstrate that the effects of aging degradation are adequately managed for the period of extended operation. These guidelines contain mandatory and needed requirements and an implementation schedule for nuclear units employing B&W nuclear steam supply systems (NSSS) currently operating in the United States (U.S.).
MRP-227-A includes a safety evaluation report (SER) prepared by the US Nuclear Regulatory Commission (NRC). The U. S. NRC staff determined whether the guidance contained in the report provided reasonable assurance that the I&E guidelines ensured that the RV internal components will maintain their intended functions during the period of extended operation. From the determination, seven topical report conditions and eight plant-specific applicant/licensee action items (A/LAIs) were contained in the SER to alleviate issues and concerns of the NRC staff. The plant-specific A/LAIs address topics related to the implementation of MRP-227 that could not be effectively addressed on a generic basis in MRP-227. The seventh A/LAI (A/LAI #7) addresses NRC staff concerns regarding thermal aging embrittlement (TE) and irradiation embrittlement (IE).
Con~troiled [Docu~ment AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit I Paqe 1-2 During the performance of this A/LAI, four component items were identified as requiring further aging management for ANO-1 based on material type. A fifth component item, the vent valve bodies, were also identified as being fabricated from cast austenitic stainless steel (CASS); [
]
However, this cannot be definitively confirmed unless the specific vent valve bodies currently installed at ANO-1 are known.
CQ tont oled Document AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe 1-3 Therefore, Entergy commits to record the serial numbers and heat numbers stamped on the vent valve bodies currently installed in the ANO-1 RV internals when the core barrel assembly is removed during the initial MRP-227 inspections. [
The four component items applicable to A/LAI #7 for ANO-1 are listed below:
- Control Rod Guide Tube (CRGT) Spacer Castings (Grade CF-3M)
-Screened as potentially susceptible to TE, but not IE
- Incore Monitoring Instrumentation (IMI) Guide Tube Spiders (Grade CF-8)
-Screened as potentially susceptible to IE, but not TE o Vent Valve Retaining Rings (Type 15-5 precipitation-hardenable [PH])
-Screened as potentially susceptible to TE, but not IE
- Select Original Vent Valve Locking Device Parts (Type 431)
- Screened as potentially susceptible to TE, but not IE The purpose of this document is to summarize analyses performed for these four component items for Entergy Operations, Inc.'s (hereafter referred to as Entergy)
Arkansas Nuclear One (ANO) Unit 1 (ANO-1). This document will fulfill the A/LAI for these component items; that is, to develop a plant-specific analysis for ANO-1 to demonstrate that the component items will maintain their functionality during the period of extended operation, considering the loss of fracture toughness due to TE and/or IE (whichever is applicable).
The methodology used to evaluate all four components items is similar and is illustrated in Section 2.0. Each component item has its own section (CRGT Spacer Castings -
Section 3.0, IMI Guide Tube Spider Castings - Section 4.0, Vent Valve Retaining Rings
- Section 5.0, and Select Original Vent Valve Locking Device Parts - Section 6.0).
AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe 1-4 Information considered by AREVA to be proprietary is marked with brackets: [ I
Co toled, Decument AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action item #7 Analysis for Arkansas Nuclear One Unit 1 Paqie 2-1 2.0 METHODOLOGY The purpose of this section is to provide various potential/methodologies and identify the ultimate methodology used to evaluate the component items for ANO-I.
2.1 WCAP- 17096 Methodology Applicability WCAP-1 7096 (2), including proposed edits (3), provides a methodology for developing evaluation procedures to assess the functional impacts of degradation in component items with "observed relevant conditions." [
]
2.2 ./::l"MRP-22 7-A Suggested Methodologies As dlescribed in A/LAI #7, to address the NRC staff concerns regarding TE and IE of potentially susceptible materials, applicants/licensees are required to perform a plant-specific analysis or evaluation demonstrating that certain component items will maintain their functionality during the period of extended operation. Per MRP-227-A, possible acceptable approaches may include, but are not limited to:
- Functionality analyses for the set of like components or assembly-level functionality analyses, or
- Component level flaw tolerance evaluation justifying that the MRP-227 recommended inspection technique(s) can detect a structurally significant flaw for the component in question, taking into account the reduction in fracture toughness due to IE and TE; or
<~urn~ n~
AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Pane 2-2
- For CASS, if the application of applicable screening criteria for the component's material demonstrates that the components are not susceptible to either TE or IE, or the synergistic effects of TE and IE, then no other evaluation would be necessary.
For assessment of CASS materials, the licensees or applicants for license renewal (LR) may apply the criteria in the NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components" (4) as the basis for determining whether the CASS materials are susceptible to the TE mechanism.
2.3 2.3 Methodology Utilized for A NO-I1
AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit I Paqe 3-1 3.0 CRGT SPACER CASTINGS This section summarizes the analysis performed of the ANO-1 CRGT spacer castings to fulfill AILAI #7 from MRP-227-A.
3.1 Background MRP-227-A provides I&E guidelines for the various component items including the CRGT spacer castings, which are considered a "Primary" item in MRP-227-A. The I&E guidelines specify applicability, effect and mechanism, expansion link, examination method/frequency, and examination coverage.
3.1.1 Description of the Component Item This section contains an abbreviated description, including a short description of the functionality, consequence of failure, and operating experience of the component items.
The plenum assembly (upper internals) contains 69 vertical CRGT assemblies that are welded to the plenum cover plate and bolted to the upper grid. Inside of each guide housing is a brazement subassembly consisting of ten parallel spacer castings brazed to twelve perforated vertical rod guide tubes and 4 pairs of vertical rod tube guide sectors. There are a total of 690 spacer castings in the ANO-1 RV internals. The CRGT spacer castings are made from ASTM A 351-65, Grade CF-3M castings.
During normal operation, [
] Inthe event of a reactor trip or a rod movement command from the control room, the CRAs pass through the path provided by the brazement into, or out of, the fuel assemblies. (
]
AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqie 3-2 There are openings in the lower region of the pipe to allow some of the fluid entering the CRGT assembly from the core to exit to the plenum region. [
The function of the CRGT spacer castings is to [
] The CRA consists of a control rod spider and control rods that travel vertically within the rod guide brazement.
] The CRA is guided by the brazement subassembly over the entire range of the vertical withdrawal path. In addition, the rod guide tubes limit reactor coolant cross-flow on the control rods to limit flow-induced vibration. The spacer castings do not have a core support function; however, they do have a safety function relative to control rod alignment, insertion, and reactivity issues. Degradation of the spacer castings could result in degradation in the unit shutdown capability by hindering the insertion of the control rods into the core in the normal anticipated time.
Appendix A of MRP-227-A indicates that the failure of CASS materials due to TE in the PWR RV internals has not been reported. Additionally, no known failures of CASS materials due to embrittlement have been reported in the industry.
3.2 Evaluation Inputs This section will describe the quantitative inputs for the evaluation, such as flaw size, degraded material properties, and stresses.
AREVA Inc. ANP-3417NP
~Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Page 3-3 3.2.1 Flaw Size As indicated by the MRP-227-A process, the CRGT spacer castings are not screened as potentially susceptible to service induced flaws (i.e., irradiation-assisted stress corrosion cracking [IASCC], SCC, or fatigue). Therefore, the following section will focus on the potential for flaws in the 'as-built' condition from the manufacturing process. The non-destructive evaluation (NDE) methods [
3.2.2 Degraded M~aterial Properties
AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paole 3-4 3.2.3 Distortion Evaluation 3.3 Evaluation The results of the methodology utilized are organized into several conclusions as discussed in the following sections.
AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqie 3-5 3.3.1 Failure is Unlikely
] In 2012 and 2013, two B&W units performed visual testing (VT)-3 examinations of the CRGT spacer castings, per the guidance in MRP-227-A. These visual examinations with 100% coverage of accessible surfaces at each of the four CRGT spacer casting screw locations revealed no recordable indications. [
I
Con~trolled D* *e ,mri AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkan.£v. Ntil-.kar flne I lnit I Pinp ~-R 3.3.2 Effect of Failure on Functionality
AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Pacie 3-7 3.4 Conclusions Cast austenitic stainless steel materials are known to be potentially susceptible due to TE after exposure at PWR RV internals temperatures for long periods of time, especially those containing higher levels of ferrite and molybdenum.
] The CRGT spacer casting material that does not exceed the screening criteria is not considered potentially susceptible to TE.
AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paae 3-8 Based on the discussion above, it is concluded that the CRGT spacer castings will maintain functionality during the period of extended operation.
,, onoIed D ocu ment AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit I Pa~qe 4-I 4.0 IMII GUIDE TUBE SPIDER CASTINGS This section summarizes the analysis performed of the ANO-1 IMI guide tube spider castings to fulfill A/LAI #7 from MRP-227-A.
4.1 Background MRP-227-A provides I&E guidelines for the various component items including the IMI guide tube spider castings, which are considered a "Primary" item in MRP-227-A. The l&E guidelines specify applicability, effect and mechanism, expansion link, examination method/frequency, and examination coverage.
4.1.1 Description of the Component Item This section contains an abbreviated description, including a short description of the functionality, consequence of failure, and operating experience of the component items.
The IMI guide tube spider castings are part of the lower internals assembly. Fifty-two IMI guide tube assemblies provide support and protection for the IMI along the path from the RV bottom head IMI nozzles, through the lower internals, and into the instrument tubes in the fuel assemblies.
The IMI guide tube spider castings are ASTM A351-65 Grade CF-8 material and resemble a four eared butterfly nut. The IMI guide tube spider casting has a center hub with four integral "L" shaped legs extending outward. The inner diameter of each IMI guide tube spider casting center hub is chrome plated. Each of the 52 IMI guide tube spider castings is custom machined to fit within the lower grid rib section.
AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe 4-2 Each of the four IMI guide tube spider casting legs is fillet welded to the walls of the lower grid rib section. The welds are a stainless steel filler metal (ER 308/308L). The tip at the upper end of the IMI guide tube slides inside the chrome-plated center hub of the IMI guide tube spider casting. [
]
The lower end of the IMI guide tube is solidly welded to the flow distributor head (in some locations with the use of gussets) with additional support provided at its midsection via a threaded guide tube nut to the IMI guide support plate.
The function of the IMI guide tube spider casting is to provide lateral restraint for the IMI guide tube and the function of the spider fillet welds is to hold the IMI guide tube spider casting in place. The IMI guide tube provides the continuous protected guide path for the in-core monitoring instrumentation from their entry into the RV through the RV instrumentation nozzles to the entrance into the fuel assembly instrument guide tube.
] Loss of function of the in-core monitoring instrument guide path would be a sufficient misalignment at the fuel assembly instrument tube entrance to prohibit entry of the in-core instrument. In addition, failure of the guide path could result in wear of the IMI sheath due to flow-induced vibration (FIV) and therefore would be considered a loss of function.
Appendix A of MRP-227-A indicates no cracking has been reported in the PWR RV internals as being attributed to embrittlement for CASS materials. Cast stainless steels are also used extensively in pressure-boundary components such as piping components, valve bodies, and pump casings. However, no cases of embrittlement requiring corrective action have been reported in the industry as of 2015.
Contr oled D cu mern, AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe 4-3 4.2 Evaluation In puts This section will describe the quantitative inputs for the evaluation, such as flaw size, degraded material properties, and stresses.
4.2.1 Likelihood of Fabrication and Service-Induced Flaws
] acceptable by the American Society of Mechanical Engineers (ASME) Code for castings in pressure boundary applications.
Additionally, service-induced flaws such as those resulting from SCC, IASCC, and fatigue, were evaluated and not expected for the ANO-1 IMI guide tube spider castings.
4.2.2 Driving Force for Crack Extension A structural analysis evaluated the design configuration for the ANO-1 IMI guide tube spider castings and considered two loading configurations:
AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqie4-4 4.2.3 Irradiated Fracture Toughness
[
] The parameter J characterizes the crack driving force based on an integration of loading work per unit volume (e.g., strain energy density for elastic bodies) around a crack front. J1c characterizes the crack driving force just prior to the onset of significant stable tearing crack extension. J2.5mm characterizes the crack driving force required to achieve a crack extension of 2.5mm.
AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqie 4-5 4.2.3.1 Fracture Toughness Characterized by J1 c There is a paucity of fracture toughness data available for CASS material, particularly in the ( ] relevant to the ANO-1 IMI guide tube spider castings. No measured fracture toughness properties were identified in this task for CF-8 materials irradiated in light water reactors. Reference (5) reports fracture toughness properties (measured at room temperature) for a CF-8 material irradiated at 32500 in a fast-breeder reactor between roughly 0 and 12 dpa. The Reference (5) data are summarized in Figure 54 of Reference (6) as "CF-8 (Burke et al.)", including an additional measurement at 19 dpa. A lower bound to this data determined per engineering judgment suggests that the [
J based on fracture toughness categorizations described in Reference (6).
Figure 61 of Reference (6) shows that for irradiated materials [
The key point from this discussion is that the ANO-1 IMI guide tube spider casting [
I
Conta olled ©ocument AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe 4-6 4.2.3.2 Fracture Toughness Characterized by J2.5mm The NRC has adopted J2 .5mm = 255 kJim2 as a conservative criterion for piping materials to differentiate between non-significant and potentially significant reduction in fracture toughness for CASS subject to thermal aging embrittlement (4). A joint Boiling Water Reactor Vessel and Internals Program (BWRVIP)/MRP Working Group on CASS has compiled information of J2.5mm for irradiated CF-8 materials as a function of neutron exposure. Most of this J2.5mrn data is from the same CF-8 testing from which J1scwas discussed in Section 4.2.3.1 and shows that J 2 .5mm = 255 kJ/m 2 is not reached until about 3.3 dpa. In addition, the BWRVIP/MRP Working Group reported example calculations of J2.5mm for RV internals components with large flaws. These calculations show that a crack driving force of Japplied = 255 k Jim2 is unlikely to be achieved in RV internals components, adding further conservatism to the use of J2.5rmm = 255 kJ/m 2.
] There is also margin between Japplied = 255 kJim2 and RV internals loading conditions.
The key point from this discussion [
] Thus, the reduction in fracture toughness due to IE is not considered significant.
4.3 Evaluation The results of the methodology utilized are organized into several conclusions as discussed in the following sections.
Controfled Documeint AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe 4-7 4.3.1 Likelihood of Failure Three parameters must be considered to evaluate the likelihood of failure due to reduced fracture toughness: 1) likelihood for flaw to be present, 2) driving force for crack extension and 3) material fracture toughness. Considering each of these parameters in turn, it is unlikely that an ANO-1 IMI guide tube spider casting will fail due to irradiation embrittlement.
~ontroWed Do : ne AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paie 4-8 4.3.2 impact of Fractured Spider Casting on Functionality
Controlled Document AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit I Paqe 4-9 4.4 Conclusions Therefore, the IMI guide tube spider castings are expected to perform their function for the period of extended operation.
AREVA Inc. ANP-3417NP Revision 0 MRP-227-A ApplicantlLicensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Page 5-1 5.0 VENT VALVE RETAINING RINGS This section summarizes the analysis performed of the ANO-1 vent valve retaining rings to fulfill A/LAI #7 from MRP-227-A.
5.1 Background MFRP-227-A provides I&E guidelines for the various component items including the vent valve retaining rings, which are considered a "Primary" item in MRP-227-A. The l&E guidelines specify applicability, effect and mechanism, expansion link, examination method/frequency, and examination coverage.
5.1.1 Description of the Component Item This section contains an abbreviated description, including a short description of the functionality, consequence of failure, and operating experience of the component items.
ANO-1 has eight vent valves installed in the core support shield (OSS) cylinder. Each vent valve is mounted in a vent valve mounting ring (also called vent valve nozzle) which is welded into the CSS cylinder. For all normal operating conditions, the vent valve is closed but in the event of a pipe rupture in the RV inlet pipe, the valve will open to permit steam generated in the core to flow directly to the break, and will permit the core to be flooded and adequately cooled after emergency core coolant has been supplied to the RV. Each valve assembly includes two retaining rings with varying thicknesses that have integral threaded bosses at both ends to accept the jackscrews.
They are fabricated from AMS 5658 Type 15-5 PH stainless steel in the H1100 condition.
Oontrolleu Docum ,rnc AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One. t nit I Paqe 5-2 As of the evaluation date, there is no known cracking or failures of the vent valve retaining rings; there are several known instances of more susceptible types of PH stainless steel materials (e.g., Type 17-4 PH) in other components and systems failing.
5.2 Evaluation Inputs This section will describe the quantitative inputs for the evaluation, including inputs such as flaw size, degraded material properties, and stresses.
5.2.1 Flaw Size
- °* * ' *i * * * "I' AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe 5-3 5.2.2 Degraded Material Properties 5.2.3 Stresses
Controlled Document ARE VA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Page 5-4 5.3 Evaluationf The results of the methodology utilized are organized into several conclusions as discussed in the following sections.
5.3.1 Failure is Unlikely
[
] Additionally, as of the publication of MRP-227-A, there is no known cracking of the vent valve retaining rings. [
I
- o~*,*~le Dou~a AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 -Paqe 5-5 5.3.2 Effect of Failure on Functionality While failure of the vent valve retaining rings is not expected, this section describes the outcome, should a vent valve retaining ring fail, on the two functions of the vent valve retaining ring.
One of the functions of the vent valves is to relieve pressure in the interior of the core support assembly during a cold leg large break LOCA. The retaining rings, if damaged due to TE (cracked, fractured material, surface irregularities, etc.), (
] Therefore, it is likely that degradation of the vent valve retaining ring material due to TE will not affect the function of the vent valve during a cold leg large break LOCA.
Contro led Docuen AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit I Paqe 5-6 5.4 Conclusions Therefore, the vent valve retaining rings are expected to perform their function for the period of extended operation and in the unlikely event of failure, the primary vent valve functions is not expected to be impaired and the secondary vent valve function that could possibly be impaired would be detectable.
Controlld ocument AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit I Paqie 6-1 6.0 SELECT ORIGINAL VENT VALVE LOCKING DEVICE PARTS The vent valve locking devices were not included in the MRP-227-A I&E guidelines; however they were screened and evaluated in the response to Applicant/Licensee Action Item #2 for ANO-1.
As of February 2015, the four vent valves near the outlet nozzles have modified locking devices (7); therefore both the modified and original locking devices were evaluated in Applicant/Licensee Action Item #2 for ANO-I. Engineering evaluation and assessment was performed after consideration of screening parameters, failure, modes, effects, and criticality analysis (FMEGA), and severity categorization, for select original and modified vent valve locking device parts in Applicant/Licensee Action Item #2 for ANO-1, including two component items fabricated from martensitic stainless steel. The [
] (both fabricated from martensitic stainless steel) within the original vent valve locking device are within the scope of LR for ANO-1 and require aging management.
6.1 Background ANO-1 has eight vent valves installed in the CSS cylinder. Each vent valve is mounted in a vent valve mounting ring (also called vent valve nozzle) that is welded into the CSS cylinder. Each vent valve consists of a hinged disc, a valve body with sealing surfaces, a split-retaining ring and fasteners (that retain and seal the perimeter of the valve assembly), and an alignment device (to maintain the correct orientation). For all normal operating conditions, the Vent valve is closed but in the event of a pipe rupture in the RV inlet pipe, the valve will open to permit steam generated in the core to flow directly to the break, and will permit the core to be flooded and adequately cooled after emergency core coolant has been supplied to the RV.
Contro ld Document AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paae 6-2 6.2 Evaluation The following section contains a discussion of the applicable degradation mechanism and effect of failure on functionality for select original vent valve locking device parts
[ ] fabricated from martensitic stainless steel (Type 431).
Li 6.2.1 Degradation Mechanism 6.2.2 Effect of Failure on Functionality I
Cont:rolled Document AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas~ Ni clear One Unit I c~
6.3 Conclusion
ControlledA©ocument AREVA Inc. ANP-341 7NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Panqe 7-1 7.0 OVERALL CONCLUSIONS A/LAI#7 is applicable to CRGT spacer castings, IMI guide tube spider castings, vent valve retaining rings, and select original vent valve locking device parts for ANO-I.
Based on the extensive evaluations summarized above, failure during the period of extended operation was found to be improbable for the CRGT spacer castings, vent valve retaining rings, and IMI guide tube spider castings. In the unlikely event of a failure occurring for these component items, the intended function of the component items is expected to be maintained or the failure will be detectable.
For the applicable original vent valve locking device parts, the evaluation concluded that failure of one or both of the original locking devices on a vent valve assembly would not impact the primary function of the vent valve assembly [
I
Cont otled Documen AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Paqe 8-1
8.0 REFERENCES
- 1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA:
2011. 1022863.
- 2. WCAP-1 7096-NP, Revision 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," December 2009, NRC Accession No. ML101460157.
- 3. Letter MRP 2013-019 from Tim Wells (EPRI IC Chairman) to NRC Document Control Desk, "Proposed Edits to WCAP-1 7096-A Draft,"
August 2013, NRC Accession No. ML13219A183.
- 4. Letter from Christopher I. Grimes (NRC) to Douglas J. Walters (NE I),
"License Renewal Issue No. 98-0030, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," May 19, 2000, NRC Accession No. ML003717179.
- 5. Kim, C., et. al., "Embrittlement of Cast Austenitic Stainless Steel Reactor Internals Components," Proc. 6th Intl. Symp. on Contribution of Materials Investigations to Improve the Safety and Performance of LWRs, Vol. 1, Fontevraud 6, French Nuclear Energy Society, SFEN, Fontevraud Royal Abbey, France, Sept. 18-22, 2006.
- 6. NUREG/CR-7027, "Degradation of LWR Core Internal Materials Due to Neutron Irradiation," NRC ADAMS Accession No. ML102790482.
- 7. Attachment 3 to Entergy Operations, Inc. Letter 1CAN021 503, "Responses to Request for Additional Information, Reactor Vessel Internals Aging Management Program Plan, Arkansas Nuclear One, Unit 1, Docket No. 50-313, License No. DPR-51 ," February 10, 2015, NRC Accession No. ML15043A102, contained in NRC Accession No. ML15043A103.
Controx ed Documen :
AREVA Inc. ANP-3417NP Revision 0 MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1 Panqe 8-2
- 8. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-1 75). EPRI, Palo Alto, CA: 2005. 1012081.
to 1 CAN091 503 Affidavit
Cont~roled Document AFFIDAVIT COMMONWEALTH OF VIRGINIA )
) SS, CITY OF LYNCHBURG)
- 1. My name is Morris Byram. I am Manager, Product Licensing, for ARE VA Inc.
(AREVA) and as such I am authorized to execute this Affidavit.
- 2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary. I am familiar with the policies established by AREVA to ensure the proper application of these criteria.
- 3. I am familiar with the AREVA information contained in the topical report ANP-3417P, Revision 0, "MRP-227-A Applicant/Licensee Action Item #7 Analysis for Arkansas Nuclear One Unit 1," dated September, 2015, and referred to herein as "Document." Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.
- 4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.
- 5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is
o'tf~ ~eQ~ Do u~e~A~
requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information.
- 6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary:
(a) The information reveals details of AREVA's research and development plans and programs or their results.
(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.
(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for ARE VA.
(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for ARE VA in product optimization or marketability.
(e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.
The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(d) above.
- 7. In accordance with ARE VA's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA only as required and Under suitable agreement providing for nondisclosure and limited use of the information.
- 8. AREVA policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
Cont ~o1 ,d Document
- 9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.
~7#2o,~
SUBSCRIBED before me this Lt day of t' L ,2015.
,)
Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg. # 7079129
Attachment 4 to 1CAN091503 List of Commitments to 1 CAN 0915O3 Page 1 of 1 List of Commitments This table identifies actions discussed in this letter for which Entergy commits to perform. Any other actions discussed in this submittal are described for the NRC's information and are not commitments.
TYPE (Check one) SCHEDULED COMMITMENT COMPLETION ONE-TIME CONTINUING DATE ACTION COMPLIANCE Therefore, Entergy commits to record the serial numbers and heat numbers stamped on the vent valve bodies currently installed v"Completion of in the ANO-1 RV internals when the core 1R26 (Fall 2016) barrel assembly is removed during the initial MRP-227 inspections.______