ML16365A027
| ML16365A027 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/31/2016 |
| From: | AREVA |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16365A016 | List: |
| References | |
| 1CAN121601 ANP-3486NP, Rev 0 | |
| Download: ML16365A027 (43) | |
Text
Attachment 2 to 1CAN121601 AREVA Document ANP-3486NP, Revision 0 MRP-227-A Applicant I Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 NON-PROPRIETARY
A AREVA MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report October 2016 AREVA Inc.
(c) 2016 AREVA Inc.
ANP-3486NP Revision 0
Copyright © 2016 AREVA Inc.
All Rights Reserved ANP-3486NP Revision 0
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page i Item 1
Section(s) or Page(s)
All Nature of Changes Description and Justification Initial Issue
AREVA Inc. 1
' ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page ii Contents Page
1.0 INTRODUCTION
AND PURPOSE.................................................................... 1-1
2.0 BACKGROUND
................................................................................................. 2-1 3.0 OVERALL METHODOLOGY.............. :.............................................................. 3-1 4.0 CORE BARREL CYLINDER EVALUATION...................................................... 4-1 4.1 Background............................................................................................. 4-1 4.2 Detailed Methodology............................................................................. 4-2 4.3 Evaluation............................................................................................... 4-2 4.3.1 Facto.rs Affecting Susceptibility fo Failure by Irradiation Embrittlement.............................................................. 4-3 4.3.2 [
]......................... 4-4 4.3.3 [
4.3.4 4.3.5 4.3.6 4.3.7
]................................................................... 4-4
]................................. 4-5.
' ]............................. 4-7
]................................................. :.................................... 4-7
]........................................................................... 4-8 4.4 Conclusion.............................................................................................. 4-8 5.0 FORMER PLATE EVALUATION....................................................................... 5-1 5.1 Background.............................................. ;...,"......................................... 5-1 5.2 Detailed Methodology.............................................................................. 5-1.
5.3 Evaluation............................................................................................... 5-2 5.3.1 Factors Affecting Susceptibility to Failure by Irradiation Embrittlement.............................................................. 5-2 5.3.2 [
]......... -................ 5-3 5.3.3 [
5.3.4 5.3.5 5.3.6 5.3.7
]............................................................................. 5-3
]................................. 5-4
]............................................. 5-4
]........................ 5-4
]......................................................................................... 5-5 5*,4 Conclusion<............................................................................................. 5-5
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page iii 6.0 APPLICABLE CORE BARREL ASSEMBLY BOLTING EVALUATION***************************************************'************************************************ 6-1 6.1 Background............................................................................................. 6-2 6.2 Detailed Methodology............................................................................. 6-2 6.3 Evaluation............................................................................................... 6-3 6.3.1 Failure of the Locking Devices and Locking Welds of the External Baffle-to-Baffle Bolts and Core Barrel-to-Former Bolts................................................................................ 6-4 6.3.2 Irradiation-Assisted Stress Corrosion Cracking of the Baffle-to-Baffle Bolts and Core Barrel-to-Former Bolts................ 6-4 6.3.3 Irradiation Embrittlement.............................................................. 6-6 6.3.4 Irradiation-Enhanced Stress Relaxation/Irradiation Creep Effects on Susceptibility to Wear and Fatigue................... 6-7 6.3.5 'Void Swelling of Baffle Plates Effects on Bolting.......................... 6-7 6.3.6 Summary and Results.................................................................. 6-7 6.4 Conclusion............................................................................................ 6-10 7.0 INTERDEPENDENCE OF INACCESSIBLE AND NON-INSPECTABLE COMPONENT ITEMS.............................................................. 7-1 8.0 OVERALL CONCLUSIONS............................................................................... 8-1
9.0 REFERENCES
.................................................................................................. 9-1
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page iv List of Tables Table 2-1 MRP-227-A A/LAI 6 Applicable Reactor Vessel Internals Component Items and Welds Primary-to-Expansion Linkage....................................... 2-4 Table 6-1 Applicable Age-Related Degradation Mechanisms of AN0-1 Core Barrel Assembly Bolting and Locking Devices and Locking Welds........... 6-2 Table 6-2 [
]................................................ 6-9 Table 6-3 [
]................ 6-9
(_
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page v List of Figures Figure 2-1 Overview of Typical Babcock & Wilcox Reactor Vessel Internals (some items rotated for clarity).................................................................. 2-1 Figure 2-2 Core Barrel Assembly (Thermal Shield not Shown).................................... 2-2 Figure 2-3 Core Barrel Assembly (Cross-Section)....................................................... 2-3
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee,A.ction Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page vi Acronym A/LAI AN0-1 B&W B-B B-F CB CBA CB-F CMTR css cw dpa EFPY EPRI l&E IASCC IE ISR/IC LWR MeV MRP n/cm2 NOE NRC OE PT PWR RT sec
- SER U.S.
UT vs VT-3 Nomenclature Definition Applicant/Licensing Action Item Arkansas Nuclear One Unit 1 Babcock & Wilcox Baffle-to-Baffle Baffle-to-Former Core Barrel Core Barrel Assembly Core Barrel-to-Former Certified Material Test Report Core Support Shield Cold-Worked Displacements per Atom Effective Full Power Year Electric Power Research Institute Inspection and Evaluation Irradiation-Assisted Stress Corrosion Cracking Irradiation Embrittlemerit Irradiation-Enhanced Stress Relaxation/Irradiation Creep Light Water Reactor Million (Mega) Electron Volt Materials Reliability Program Neutrons per Square Centimeter Non-Destructive Examination Nuclear Regulatory C9mmission Operating Experience Liquid (Dye)-Penetrant Test(ing)
Pressurized Water Reactor Radiographic Test(ing)
Stress Corrosion Cracking Safety Evaluation Report United States Ultrasonic Test(ing)
Void Swelling Visual Test(ing) Method {defined in American Society of Mechanical Engineers Boil~r & Pressure Vessel Code,Section XI)
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page vii ABSTRACT This document addresses MRP-227-A Applicant/Licensee Action Item 6 for Arkansas Nuclear One Unit 1. Evaluations are included herein that justify the acceptability of the following inaccessible or non-inspectable reactor vessel internals component items for continued operation through the period of extended operation:
Core barrel cylinder (including the vertical and circumferential seam welds),
which is susceptible to a reduction in toughness due to irradiation embrittlement (Section 4.0)
Former plates, which are susceptible to a reduction in toughness due to irradiation e.mbrittlement (Section 5.0)
Applicable core barrel, assembly bolting (i.e., internal baffle-to-baffle bolts, external baffle-to-baffle bolts and the associated locking devices and locking welds, and the core barrel-to-former bolts and the associated locking devices and locking welds), which are susceptible to multiple age-related degradation mechanisms (Section 6.0)
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 1-1
1.0 INTRODUCTION
AND PURPOSE The Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) developed im;;pection and evaluation (l&E) guidelines in MRP-227-A (Reference 1) for managing the long-term aging of reactor vessel internal component items of pressurized water reactors (PWRs). The l&E guidelines define requirements for inspections that will allow owners of PWRs to demonstrate that the effects of age-related degradation are adequately managed for the period of extended operation.
MRP-227-A includes a safety evaluation report (SER) prepared by the United States (U.S.) Nuclear Regulatory Commission (NRC). The NRC Staff determined whether the guidance ensured that the reactor vessel internals component items will maintain their intended functions during the period of extended operation. From the NRC Staff determination, seven (7) topical report conditions and eight (8) plant-specific applicant/licensee action items (A/LAls) were contained within the SER to alleviate issues and concerns of the NRC Staff. The plant-specific A/LAls address topics related to the implementation of MRP-227-A that could not be effectively addressed on a generic basis. A/LAI 6 addresses the NRC Staff concerns regarding inaccessible (or non-inspectable) reactor vessel internals component items. A/LAI 6 reads as follows:
As addressed in Section 3.3.6 in this SE, MRP-227 does not propose to inspect the following inaccessible components: the B&W core barrel cylinders (including
/'
vertical and circumferential seam welds), B&W former plates, B&W external baffle-to-baffle bolts and their locking devices, B&W core barrel-to-former bolts and their locking devices, and B&W core barrel assembly internal baffle-to-baffle bolts. The MRP also identified that although the B&W core barrel assembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectable using currently available examination techniques.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 1-2 Applicants/licensees shall justify the acceptability of these components for continued operation through the period of extended operation by performing an evaluation, or by proposing a scheduled replacement of the components. As part of their application to implement the approved version of MRP-227, applicants/licensees shall provide their justification for the continued operability of each of the inaccessible components and, if necessary, provide their plan for the replacement of the components for NRG review and approval.
This document will address the applicable reactor vessel internals component items to fulfill A/LAI 6 of MRP-227-A for Entergy Operations Inc., Arkansas Nuclear One Unit 1 I
(AN0-1 ). The applicable reactor vessel internals component items are as follows:
Core barrel (CB) cylinder, which includes the vertical and circumferential seam welds Former plates Applicable core barrel assembly (CSA) bolting, which includes the internal baffle-to-baffle (8-8) bolts, external 8-8 bolts and their associated locking devices and locking welds, and the core barrel-to-former (CB-F) bolts and their associated locking devices and locking welds Information considered proprietary to AREVA Inc. is enclosed in brackets: []
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 2-1
2.0 BACKGROUND
The CBA is a flanged cylinder, with its top flange bolted to the bottom flange of the core support shield (CSS) assembly and its bottom flange bolted to the top flange of the lower internals assembly (Reference 1 ). The CBA consists of the CB cylinder, baffle plates, former plates, thermal shield, and the associated bolting, as shown in Figure 2-1,
Figure 2-2, and Figure 2-3. Its functions are to direct the flow of coolant and to support the lower internals assembly. In addition, the thermal shield reduces the amount of radiation that reaches the reactor vessel.
A Plenum Cover Plenum
~ Assembly Assembly 1
yVentValve
~
r.o Core SuppOI Shield
&E
- c. 31
~~
,r I Cl l2 a.,
uz Ill Plenum Cylinder
_L I
Assembly 25
~
~
Upper Gnd ~mb E
~
31 1
Q
~
I I g
~
Ill Thermal Shield
~
co 8
~
Core 6a1Tel 8
I I
\\,,~a lower Grid ~embly
"'c: E
! a; 31
.....5 ~
I
!MI Gulde Tube Figure 2-1 Overview of Typical Babcock & Wilcox Reactor Vessel Internals (some items rotated for clarity)
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 2-2 Baffle to Baffle Bolts I
Core Barrel_;
Figure 2-2 Core Barrel to Former Bolts Core Barrel Assembly (Thermal Shield not Shown)
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 2-3 Top Flang.e Thermal Shield fii ii Upper Restraint Assembly I.. * * * * * * * * * * * * * * *** * * * -
Thermal Shield
~
- *** 0 Core Barrel Cylinder
~ * *..
0* **..... * * * * *** * * * -
Formers Baffle Pia es Bottom Flange Figure 2-3 Core Barrel Assembly (Cross-Section)
The MRP-227-A A/LAI 6 applicable reactor vessel internals component items (as listed in Section 1.0) are shown in Table 2-1 along with the Primary component item-to-Expansion component item linkage (Reference 1 ). Unless otherwise noted, the Expansion component items have the same age-related degradation mechanisms as the linked Primary component items.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 2-4 Table 2-1 MRP-227-A A/LAI 6 Applicable Reactor Vessel Internals Component Items and Welds Primary-to-Expansion Linkage Primary Component ltem(s)
Age-Related Degradation Expansion Component ltem(s)
Mechanism(s)
Baffle-to-Former (B-F) Bolts Irradiation-Assisted Stress
- External Baffle-to-Baffle (B-B)
Corrosion Cracking (IASCC),
Bolts Irradiation Embrittlement (IE),
- Core Barrel-to-Former (CB-F) and Irradiation-Enhanced Stress Bolts Relaxation/Irradiation Creep (ISR/IC)
IE, Overload, and ISR/IC
- Internal Baffle-to-Baffle (B-B)
Bolts Baffle Plates IE
- Core Barrel (CB) Cylinder (including vertical and circumferential seam welds)
- Locking Devices (including locking welds) of Baffle-to-locking welds) of Core Barrel-Former (B-F) Bolts and Internal to-Former (CB-F) Bolts and Baffle-to-Baffle (B-B) Bolts External Baffle-to-Baffle (B-B)
Bolts Note 1: The MRP-227-A Expansion component items and welds are not linked to overload failure, but only to IASCC and IE.
J AREVA Inc.
ANP-3486NP
- Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 3-1 3.0 OVERALL METHODOLOGY The following is the overall methodology for addressing MRP-227-A A/LAI 6:
First, justify that the CB cylinder (which includes vertical and circumferential seam welds) is unlikely to fail during the period of extended operation, and, thus are expected to maintain functionality through the AN0-1 period of extended operation (detailed methodology in Section 4.2)
Second, justify that the former plates are unlikely to fail during the period of extended operation, and thus, are expected to maintain functionality through the AN0-1 period of extended operation (detailed methodology in Section 5.2)
,~
Third, justify that sufficient failures bf applicable CBA bolting (i.e., internal B-B bolts, external B-B bolts and their associated locking devices and locking welds, and CB-F bolts and their associated locking devices and locking welds), which would affect CBA functionality, are unlikely to occur during the period of extended operation (detailed methodology in Section 6.2)
Lastly, use the above information to address potential interdependence between the inaccessible (or non-inspectable) reactor vessel internals component items I
during the period of extended operation
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis'for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 4-1 4.0 CORE BARREL CYLINDER EVALUATION This section will justify that the AN0-1 CB cylinder, which includes vertical and circumferential seam welds, is unlikely to faii* during the AN0-1 period of extended operation and, thus, is expected to maintain functionality through the AN0-1 period of extended operation. The functions of the CB cylinder are to:
- 1) Direct flow down through the annulus between the CB cylinder and the reactor
- vessel,
- 2) Support the former plates and baffle plates,
- 3) Support the lower internals, and
- 4) Support the thermal shield.
These functions support the function of the CBA (i.e., [
to support the lower internals assembly).
4.1 Background
] and In accordance with the results of the MRP-227-A process (Reference 1 ), the CB cylinder is susceptible to irradiation embrittlement (IE), which is the phenomenon of reduction in ductility and fracture toughness from exposure to high energy neutrons (E>1.0 Million (Mega) electron Volts (MeV)). This decrease in ductility is accompanied by a corresponding increase in yield strength. The resultant effects on mechanical properties are due to (in part) point defects created in the atomic lattice from the neutron bombardment, which impede or hinder plastic deformation.
Failure of the CB cylinder herein refers to [
]
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 4-2 4.2 Detailed Methodology The following methodology was used to address MRP-227-A A/LAI 6 for the AN0-1 CB cylinder:
[
First, discuss the factors affecting susceptibility to failure as the result of reduced fracture toughness from IE (Section 4.3.1)
. [
. [
. [. [
. [
]
]
]
]
]
Lastly, use the results of the above steps to justify that failure of the CB cylinder is unlikely during the AN0-1 period of extended operation and, thus, the CB cylinder is expected to maintain functionality through the AN0-1 period of extended operation (Section 4.4)
]
4.3 Evaluation This section will implement the methodology stated above.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 4-3 4.3.1 Factors Affecting Susceptibility to Failure by Irradiation Embrittlement Failure related to IE is defined by rapid unstable crack growth (i.e., fast fracture) due to the effects of IE (loss of fracture toughness) and high tensile stresses. This can be caused by two scenarios:
- 1) Overload from an applied tensile stress greater than the yield strength; or
- 2) The existence of a flaw with an applied stress, potentially significantly lower than the yield strength, which exceeds the reduced fracture toughness of the material.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 4-4 4.3.2
[
1 4.3.3
[
1
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 4-5 4.3.4
[
l
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 4-6
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 4-7 4.3.5
[
1 4.3.6
[
1
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 4-8 4.3.7
[
]
4.4 Conclusion In accordance with the MRP-227-A process, the AN0-1 CB cylinder, which includes vertical and circumferential seam welds, is susceptible to a reduction in toughness due to IE. Failure related to IE is characterized by rapid unstable crack growth (i.e., fast fracture), [
] Therefore, the AN0-1 CB cylinder is unlikely to fail due to the effects of IE during the AN0-1 period of extended operation and, thus, the CB cylinder is expected to maintain its functionality through the AN0-1 period of extended operation.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 5-1 5.0 FORMER PLATE EVALUATION This section will justify that the AN0-1 former plates are unlikely to failt during the AN0-1 period of extended operation and, thus, are expected to maintain functionality through the AN0-1 period of extended operation. The functions of the former plates are to:
- 1) Support the baffle plates (in conjunction with the B-F and CB-F bolts) and
- 2) Allow flow through the annulus between the baffle plates and CB cylinder (through flow holes in the former plates).
Both of these functions contribute to the [
function of the CBA.
] portion of the
5.1 Background
In accordance with the results of the MRP-227-A process (Reference 1 ), the former plates are susceptible to IE, which is the phenomenon of reduction in ductility and fracture toughness from exposure to high energy neutrons (E>1.0 MeV). This decrease in ductility is accompanied by an increase in yield strength.
5.2 Detailed Methodology The following methodology was used to address MRP-227-A A/LAI 6 for the AN0-1 former plates:
First, discuss the factors affecting susceptibility to failure as the result of reduced fracture toughness from IE (Section 5.3.1)
. [
]
t Failure of the former plates herein refers to [
]
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 5-2
[
. [
. [. [
. [
]
]
]
]
Lastly, use the,results of the above steps to justify that failure of the former plates is unlikely during the AN0-1 period of extended operation and, thus, the former plates are expected to maintain functionality through the AN0-1 period of extended operation (Section 5.4)
]
5.3 Evaluation This section will implement the methodology stated above.
5.3.1 Factors Affecting Susceptibility to Failure by Irradiation Embrittlement The discussion provided in Section 4.3.1 is also applicable here to the former plates.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 5-3 5.3.2
[
]
5.3.3
[
1 r*.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 5-4 5.3.4
[
]
[
]
5.3.5
[
]
5.3.6
[
]
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 5-5 5.3.7
[
]
5.4 Conclusion In accordance with the MRP-227-A process, the former plates are susceptible to a reduction in toughness due to IE. Failure related to IE is characterized by rapid unstable crack growth (i.e., fast fracture), [
]
Therefore, the AN0-1 former plates are unlikely to fail due to the effects of IE during the AN0-1 period of extended operation and, thus, the former plates are expected to maintain functionality through the AN0-1 period of extended operation.
AREVA Inc.
_ ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-1 6.0 APPLICABLE CORE BARREL ASSEMBLY BOLTING EVALUATION This section will justify that sufficient failures:t: of applicable CBA bolting (i.e., internal B-B bolts, external B-B bolts and their associated locking devices and locking welds, and CB-F bolts and their associated locking devices and locking welds), which would affect CBA functionality, are unlikely to occur during the AN0-1 period of extended operation.
Note that the applicable CBA bolting does not have any safety function beyond supporting the [
] function of the CBA. The specific function of each applicable CBA bolting item is discussed below for reference.
- I: Failure of an inaccessible or non-inspectable bolt is defined as the bolt no longer performing its function due to one or more of the applicable age-related degradation mechanisms (but most likely due to IASCC).
The same is true of the associated locking devices and locking welds.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A ApplicanULicensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-2
6.1 Background
In accordance with the MRP-227-A process (Reference 1 ), the applicable CBA bolting is susceptible to the following degradation mechanisms: IASCC, IE, overload (due to void swelling (VS) of the baffle plates), and ISR/IC, as shown in Table 6-1. The CBA bolting Primary component items and welds include the B-F bolts and their locking devices and locking welds and the internal B-B bolt locking devices and locking welds and are accessible for inspection. The remaining CBA bolting component items and welds are Expansion component items and inaccessible with the exception of the internal B-B bolts.
Table 6-1 Applicable Age-Related Degradation Mechanisms of AN0-1 Core Barrel Assembly Bolting and Locking Devices and Locking Welds Applicable Age-Related Degradation Mechanism Component ISR/IC Leading to VS of Baffle Plate IASCC Wear and Fatigue IE Leading to Bolt Overload Bolts CB-F Bolts x
x x
Internal B-B Bolts x
x x
External B-B Bolts x
x x
Locking Devices and Locking Welds x
x CB-F Bolt Locking Devices and Locking Welds External B-B Bolt Locking Devices and x
x Locking Welds 6.2 Detailed Methodology The following methodology was used to address MRP-227-A A/LAI 6 for the applicable AN0-1 CBA bolting:
J
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-3
[
o Lastly, use the above to justify that sufficient failures of the applicable CBA bolting to affect the functionality of the CBA are unlikely to occur during the AN0-1 period of extended operation 1
6.3 Evaluation This section will implement the methodology stated above.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-4 6.3.1 Failure of the Locking Devices and Locking Welds of the External Baffle-to-Baffle Bolts and Core Barrel-to-Former Bolts The locking devices and the locking welds for the internal B-B bolts and the B-F bolts (MRP-227-A Primary component items) are leading indicators for age-related degradation for the locking devices and the locking welds for the external B-B bolts and CB-F bolts (linked MRP-227-A Expansion component items). These Primary component items are leading indicators for the Expansion component items because they are susceptible to overload (due to slip between the baffle plates and bolts) in addition to IASCC and IE, as noted in Table 2-1. As of the release date of this document, three (3)
B&W units have completed their MRP-227-A required visual testing (VT-3) examinations of their locking devices and locking welds for the internal B-B bolts and the B-F bolts, which resulted in no indication of active age-related degradation (References 6 and 7). [
]
6.3.2 Irradiation-Assisted Stress Corrosion Cracking of the Baffle-to-Baffle Bolts and Core Barrel-to-Former Bolts
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-5
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-6 6.3.3 Irradiation Embrittlement Failure of the CB-F and B-B bolts due to the reduction in toughness from IE requires the presence of a flaw. This is conservatively addressed by the IASCC evaluation herein,
[
]
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-7
[
]
6.3.4 Irradiation-Enhanced Stress Relaxation/Irradiation Creep Effects on Susceptibility to Wear and Fatigue The effects of ISR/IC in the CB-F and B-B bolts can cause a reduction in preload, which could increase susceptibility to wear and/or fatigue. [
]
6.3.5 Void Swelling of Baffle Plates Effects on Bolting The consequence of VS of the baffle plate on inaccessible or non-inspectable bolts [
]
6.3.6 Summary and Results Of the age-related degradation mechanisms applicable to the inaccessible or non-inspectable reactor vessel internals bolting components items and welds, [
]
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-8
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-9 Table 6-2
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Table 6-3
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AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 6-10 6.4 Conclusion Of the age-related degradation mechanisms applicable to the inaccessible or non-inspectable reactor vessel internals bolting components items, [
] Given the 8-F bolt OE at B&W units to date (one observed UT indication in three separate unit MRP-227-A inspections) and the results of the evaluation discussed above, sufficient failures of the AN0-1 CB-F bolts to affect CBA functionality is unlikely to occur during the AN0-1 period of extended operation.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 7-1 7.0 INTERDEPENDENCE OF INACCESSIBLE AND NON-INSPECTABLE COMPONENT ITEMS The AN0-1 CB cylinder, former plates, and applicable CBA bolting are not accessible for inspection. [
] Therefore, interdependence is not a concern at this time.
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 8-1 8.0 OVERALL CONCLUSIONS This report summarized the analyses performed for the applicable reactor vessel internals component items at AN0-1 to complete A/LAI 6 from MRP-227-A. The following conclusions were reached:
The CB cylinder is unlikely to fail during the AN0-1 period of extended operation and, therefore, is expected to maintain functionality through the AN0-1 period of extended operation The former plates are unlikely to fail during the AN0-1 period of extended operation and, therefore, is expected to maintain functionality through the AN0-1 period of extended operation Sufficient failures of applicable CBA bolting (i.e., internal B-B bolts, external B-B bolts and their associated locking devices and locking welds, and the CB-F bolts and their associated locking devices and locking welds), which would affect CBA functionality, are unlikely to occur during the AN0-1 period of extended operation and, therefore, the CBA is expected to maintain functionality
AREVA Inc.
ANP-3486NP Revision 0 MRP-227-A Applicant/Licensee Action Item 6 Analysis for Arkansas Nuclear One Unit 1 (AN0-1)
Licensing Report Page 9-1
9.0 REFERENCES
- 1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.
- 2. Attachment to AREVA letter from T. Natour to J. Molkenthin, AREVA-13-02949, "AREVA Revised Text for Draft WCAP-17096-NP Revision 2 for Transmittal to EPRI to Address NRC Reviewer Comments," October 23, 2013, NRC Accession Number ML16043A095.
- 3. "Reactor Internals Acceptance Criteria Methodology and Data Requirements,"
WCAP-17096-NP, Revision 2, December 2009, NRC Accession Number ML101460157.
- 4. Materials Reliability Program: PWR Internals Age-Related Material Properties, Degradation Mechanisms, Models, and Basis Data-State of Knowledge (MRP-211). EPRI, Palo Alto, CA: 2007. 1015013.
- 5. Materials Reliability Program: Screening, Categorization, and Ranking of B&W-Designed PWR Internals Component Items (MRP-189-Rev. 1). EPRI, Palo Alto, CA: 2009. 1018292.
Subject:
Biennial Report of Recent MRP-227-A Reactor Internals Inspection Results (Project 694),
dated May 12, 2014, NRC Accession Numbers ML14135A383, ML14135A384, and ML14135A385.
Subject:
Biennial Report of Recent MRP-227-A Reactor Internals Inspection Results, dated May 18, 2016, NRC Accession Number ML16144A789.