2016/10/04 Vogtle COL Docs - Preservice Inspection Draft … · Sent: Tuesday, October 04, 2016...

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Vogtle PEmails

From: Hoellman, JordanSent: Tuesday, October 04, 2016 12:18 PMTo: Vogtle PEmailsCc: Patel, Chandu; Gleaves, BillSubject: Preservice Inspection Draft Alternative to Support October 20, 2016 Pre-submittal

MeetingAttachments: ND-16-1706 ISI-ALT-6 PSI for Presubmittal Meeting.pdf

Attached is a draft preservice inspection alternative to support an October 20, 2016 Pre-submittal Meeting. This draft alternative does not contain any proprietary or SUNSI information; therefore, it may be released to the public. This will be followed by slides next week that will be used to facilitate the presubmittal discussion. Thanks, Corey Thomas Southern Nuclear Licensing

Hearing Identifier: Vogtle_COL_Docs_Public Email Number: 55 Mail Envelope Properties (1beaf406593f429ebf953242015b0ab6) Subject: Preservice Inspection Draft Alternative to Support October 20, 2016 Pre-submittal Meeting Sent Date: 10/4/2016 12:17:33 PM Received Date: 10/4/2016 12:17:38 PM From: Hoellman, Jordan Created By: [email protected] Recipients: "Patel, Chandu" <[email protected]> Tracking Status: None "Gleaves, Bill" <[email protected]> Tracking Status: None "Vogtle PEmails" <[email protected]> Tracking Status: None Post Office: HQPWMSMRS01.nrc.gov Files Size Date & Time MESSAGE 409 10/4/2016 12:17:38 PM ND-16-1706 ISI-ALT-6 PSI for Presubmittal Meeting.pdf 1719735 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date: Recipients Received:

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Southern Nuclear Operating Company

ND-16-1706

Enclosure

Vogtle Electric Generating Plant (VEGP) Units 3 and 4

Proposed Alternative VEGP 3&4-PSI-ALT-06 in Accordance with 10 CFR 50.55a(z)(1)

Regarding Preservice Inspection Requirements for Specific Valve to Pipe Welds

(Enclosure consists of 45 pages, including this cover page.)

ND-16-1706EnclosureProposed Alternative VEGP 3&4-PSI-ALT-06 in Accordance with 10 CFR 50.55a(z)(1)

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Plant Site-Unit: Vogtle Electric Generating Station (VEGP) – Units 3 and 4

Interval-Interval Dates: Applies to Preservice Inspection

Requested Date for Approval:

Approval is requested by April 28, 2017 to support performance of theSection XI PSI Class 1 & 2 piping examinations.

ASME Code Components

Affected:

Components affected consist of ASME Section III, Subsection NB Class1 and a limited population of ASME Section III, Subsection NC Class 2Valve to Pipe welds (ASME Section XI, Table IWB-2500-1 ExaminationCategory B-J and Table IWC-2500-1 Examination Category C-F-1,respectively) as identified in Table 1 and Table 2 of this AlternativeRequest.

Applicable Code Edition and

Addenda:ASME Section XI Code, 2007 Edition through the 2008 Addenda

Applicable CodeRequirements:

In accordance with ASME Section XI, IWA-2200 (c), all nondestructive examinations of the required examination surface or volume shall be conducted to the maximum extent practical. When performing VT-1,surface, radiographic, or ultrasonic examination on a component with defined surface or volume, essentially 100% of the required surface or volume shall be examined. Essentially 100% coverage is achieved when the applicable examination coverage is greater than 90%; however, in no case shall the examination be terminated when greater than 90% coverage is achieved, if additional coverage of the required examination surface or volume is practical.

ASME Section XI, Figure IWB-2500-8 and Figure IWC-2500-8requires that for Category B-J and Category C-F-1, respectively, piping welds that the inner 1/3 T of the weld be examined for a distance of 1/4 inch into the base metal (on the ID) on each side of the weld. In this Alternative Request, the examination volume consists of the valve and pipe base metal and the weld.


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Reason for Request:

Class 1 and 2 austenitic valve-to-pipe welds do not allow for essentially 100% of the required examination volume coverage as defined in ASME Section XI, IWA-2200 (c) based on current ASME Section XI, Appendix VIII, Supplement 2 qualified ultrasonic examination procedures, equipment, and personnel.

Current ASME Section XI, Appendix VIII, Supplement 2 qualified ultrasonic examination procedures, equipment and personnel as defined by the Performance Demonstration Initiative (PDI) program require dual side access of wrought austenitic pipe welds to obtain qualified examination coverage of the examination volume. Single-sided examinations to detect discontinuities on the far side of the weld have been demonstrated to be representative of ‘best effort’ technology but have not been able to satisfy Supplement 2 single-sided demonstration requirements as mandated by 10 CFR 50.55a (b)(2)(xv)(A)(2). In addition, these qualified ultrasonic examination procedures, equipment and personnel are not qualified for examination of cast austenitic materials.

There are two categories of austenitic valve-to-pipe welds addressed in this request.

Category 1 austenitic valve-to-pipe welds are those with a cast austenitic valve body welded to a wrought austenitic pipe. The cast austenitic material combined with the outer diameter surface configuration of the valve prevent the application of the required ultrasonic techniques needed to detect and size inner diameter surface initiated planar flaws. Whereas, Appendix VIII does notaddress performance demonstration requirements for cast austenitic materials, it does mandate, in section 3110 (c), that ultrasonic examination procedure requirements shall be in accordance with Section XI, Appendix III as supplemented by Table I-2000-1 of Section XI, Appendix I. However appropriate ultrasonic test techniques applied from the valve austenitic body and consistent with Section XI, Appendix III are not practical given the restricted outer diameter surface configuration. Qualified ultrasonic examination coverage will be approximately 50% of the examination volume. The examination volume not covered is that associated with the austenitic valve material. There are fifty-two (52) valves in Category 1 with a total of eighty-five (85) welds on each unit. The specific welds are identified in Table 1. The nominal pipe sizes for these valve-to-pipe welds range from NPS 3” to NPS 14”. Figure 1 shows an example of a Category 1, 3” gate valve to pipe weld design configuration, ASME Section XI examination volume, and the non-credited ASME Section XI examination volume on the valve side of the weld.

Category 2 austenitic valve-to-pipe welds are those with a wrought valve body welded to a wrought austenitic pipe. The outer diameter surface configuration of the valve prevents the application of ultrasonic techniques from the valve side of the weld needed to detect and size inner diameter surface initiated planar flaws consistent with the


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Reason for Request

(Continued):

requirements of Section XI, Appendix VIII, Supplement 2. Qualified ultrasonic examination coverage will be approximately 50% of the examination volume. The examination volume not covered is that associated with the austenitic valve material. There are six (6) valves in Category 2 with a total of eight (8) welds on each unit. The specific welds are identified in Table 2. The nominal pipe size for these valve-to-pipe welds are NPS 4” and NPS 8”. Figure 2 shows the Pressurizer Spray valve to pipe weld design configuration, ASME Section XI volume, and the non-credited ASME Section XI examination volume on the valve side of the weld.

Both of these categories are limited due to the outer diameter surface configuration of the valve and the dual sided ultrasonic examination flaw detection and length sizing capabilities as defined in qualified ultrasonic examination procedures (for example, PDI-UT-2: PDI Generic Procedure for the Ultrasonic Examination of Austenitic Pipe Welds, Revision G) and as mandated by PDI performance demonstration documentation (for example, PDQS No. 859: PDI-UT-2– PDI Generic Procedure for the Ultrasonic Examination of Austenitic Pipe Welds, Revision G). The outer diameter surface configuration which includes a 30° maximum taper for a distance of 1.5 times the minimum pipe wall thickness from the valve-to-pipe weld is shown in ASME Section III, Subsection NB, Figure NB-4250-1 and Subsection NC, Figure NC-4250-1. The taper transition angle is dictated by ASME B16.34 and ASME Section III valve design requirements. Whereas a ‘no taper’ or 0° end transition with an adequate surface extent would allow the potential for dual sided access to the examination volume, but the ASME boundary and pipe nozzle loading cannot be met.

The dual sided ultrasonic examination capabilities are not possible due to current examination technology related to austenitic weld material and the welding process itself rather than the requirements of ASME Section XI. Weld material attenuation of the ultrasonic energy, weld crowns, weld shrinkage and weld root reflectors prevent a successful outcome for a single-sided Section XI, Appendix VIII performance demonstration through the PDI program. These examination constraints are not addressed in ASME Section III Subsections NB or NC (NB/NC-4000). ASME Section III Subsection NB only addresses surface finish on the pipe side of the weld (NB-4424.2). Whereas NB-4424.2 Pre-Service Examination does provide preservice examination requirements, NC does not.


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Proposed Alternative and Basis for Use:

The proposed alternative is to implement a similar approach for the examination volume on the valve side of the weld that has been adopted in the ASME Code Section XI for Examination Category B-M-2: Valve Bodies. In ASME Code Section XI editions prior to the 2008 Addenda, welds in valve bodies were required to undergo volumetric examinations. However in the 2008 Addenda of ASME Code Section XI, the volumetric examination requirement was removed and substituted with a VT-3 visual examination of the internal surfaces when a valve is disassembled for maintenance or repair. This change was justified based on the high flaw tolerance of cast and wrought austenitic materials.

While the proposed alternative could be to eliminate the requirement for a volumetric examination of the austenitic valve material in a valve-to-pipe weld, ‘best effort’ outer diameter surface ultrasonic examinations of the valve side examination volume using ultrasonic test techniques from the pipe side of the weld, on the conditioned weld surface and where practical from the valve side of the weld will be performed.

Scanning will be directed in at least one axial direction from the pipe toward the valve and two circumferential directions. It is noted that by design the weld crown is to be flush or smooth across the entire weld surface and the transition to the pipe is to blended smooth to provide less than or equal to 1/32-inch/inch (0.8 mm/mm) flatness. This allows for scanning on the weld surface with lower angles (30° to 50° inside surface impingement beam angles) in order to minimize the distance traveled in the austenitic weld and base materials in an effort to reduce beam attenuation.

For Category 2 welds (Table 2), the ‘best effort’ ultrasonic test techniques will be as defined in qualified ASME Code Section XI, Appendix VIII procedures. Such techniques will include 1.0 – 1.5MHz, dual element, transmit-receive, refracted longitudinal wave probes focused to within 75% to 125% of the nominal wall thickness of

60% to 110% of the nominal thickness of the examination volume for beam angles > 52°. The probes will be contoured to the outer diameter surface in accordance with the procedure requirements.

For Category 1 welds (Table 1) which are not addressed in current qualified ASME Code Section XI, Appendix VIII procedures, the ‘best effort’ ultrasonic test techniques will include dual element, transmit-receive refracted longitudinal wave probes with frequencies up to 1.5 MHz, with inside surface impingement beam angles between 30° to 50°, and focusing within 80% to 110% of the examination volume wall thickness. At least one beam angle will be greater than or equal to 55°. The probes will be contoured to the outer diameter surface such that there is no more than 1/32-inch gap between the probe and the valve/pipe surface along the scan length.

For both the Category 1 and 2 welds, the mandatory preservice ASME


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Proposed Alternative and

Basis for Use (Continued):

Code Section XI, Appendix VIII ultrasonic examinations of the pipe side of the valve-to-pipe examination volume will also be performed with examination coverage of essentially 100%.

These outer diameter surface applied ultrasonic test techniques will be applied for the preservice examination in order to obtain a baseline volumetric examination of the ASME Code Section XI defined examination volume. It is noted that volumetric examination of the weld using the radiographic examination method will have already been performed in accordance with ASME Code Section III. Additionally, per the examination requirements of ASME Section III NB-2541 and NB- 2571, a liquid penetrant examination of all external and accessible interior portions of the valve bodies and machined surfaces (including the weld prep) has been completed prior to N-stamping the valves. For cast austenitic valves, visual examinations shall be performed in accordance with design specifications.

Basis for Use:The technical justification for this proposed alternative is included in Attachment 1 to this alternative. Attachment 2 provides historical information relative to compliance with Code Case N-481 and NRCacceptance of the code case. In summary the technical basis for the reduced volumetric examination of the cast and wrought austenitic valve body material for a valve-to-pipe weld is based on research, prior approval and implementation of ASME Section XI Code Cases N-481 and N-770-1 as detailed in the attachment.

Duration of Proposed

Alternative:The duration of the proposed Alternative is the Section XI Preservice Examinations.

References:

ASME Boiler & Pressure Vessel Code, Section III, Division 1, Subsection NB, Class 1 Components and Subsection NC, Class 2 Components, “Rules for Construction of Nuclear Power Plant Components”, 1998 Edition through 2000 Addenda.

ASME Boiler & Pressure Vessel Code, Code, Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components,” 2007 Edition through the 2008 Addenda

ASME Code Case N-824: Ultrasonic Examination of Cast Austenitic Piping Welds from the Outside Surface.

ASME Code Case N-481: Alternative Examination Requirements for Section XI, Division 1, Examination of Pump Casings

ASME Code Case N-770-1: Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section XI, Division 1

Status: Awaiting NRC Authorization


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Figure 1: AP1000™ 3” Motor Operated Gate Valve to Pipe Weld Representing Category 1 Cast Austenitic Valve to Pipe Welds – Design Configuration

(a) Valve-to-Pipe Weld with ASME Section XI Examination Volume Shown in Blue

(b) Representation of coverage limitations using a Phased Array Axial Beam UT Probe in Compliance with ASME Code Case N-824 Applied from the Valve Body OD Surface; the Non-Credited ASME Section XI Examination Volume is Shown in Red


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Figure 2: AP1000™ Pressurizer Spray Valve to Pipe Weld Representing Category 2 Wrought Austenitic Valve to Pipe Welds – Design Configuration

(a) Valve-to-Pipe Weld with ASME Section XI Examination Volume Shown in Blue

(b) Representation of a Conventional Axial Beam UT Probe in Compliance with PDI-UT-2Applied from the Valve Body OD Surface; the Non-Credited ASME Section XI Examination Volume is Shown in Red


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Table 1: Category 1 Cast Austenitic Valve-to-Wrought Austenitic Pipe Welds

ISO Number Valve ID Line Size (inch) Connecting Welds Comments

PXS-PLW-01K V015A 8 FW14PXS-PLW-01V 8 FW8

PXS-PLW-01K V014A 8 FW12PXS-PLW-01V 8 FW10

PXS-PLW-02X V122B 8 SW6 Class 3 weld on other side of valve; no volumetric examination required

PSX-PLW-02Y V013B 8 SW6, SW7

PXS-PLW-010 V013A 8 SW4, SW5

PXS-PLW-014 V028A 8 FW9, FW10Class 3 weld on other side of valve; no volumetric examination required

PXS-PLW-014 V029A 8 FW5, FW6

PXS-PLW-016 V122A 8 SW4 Class 3 weld on other side of valve; no volumetric examination required

PXS-PLW-017 V124A 8 SW4 Class 3 weld on other side of valve; no volumetric examination required

PXS-PLW-02U V124B 8 SW7 Class 3 weld on other side of valve; no volumetric examination required

PXS-PLW-020 V014B 8 SW7, SW8




PXS-PLW-023 V027B 8 FW1Class 3 weld on other side of valve; no volumetric examination required


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PXS-PLW-035 V101 14 FW1RCS-PLW-034 14 FW1

PXS-PLW-041 V109 14 SW9RCS-PLW-04A 14 FW1

PXS-PLW-041 V108A 14 FW13, FW14

PXS-PLW-041 V108B 14 FW3, FW4

PXS-PLW-050 V002A 8 FW10, FW11

PXS-PLW-060 V002B 8 FW5RCS-PLW-061 8 FW1

RNS-PLW-014 V001A 10 SW20, SW21

RNS-PLW-014 V001B 10 SW10, SW11

RNS-PLW-014 V002A 10 SW23

RNS-PLW-014 V002B 10 SW7, SW8

RNS-PLW-016 V023 10 SW3 Class 3 weld on other side of valve; no volumetric examination required

RNS-PLW-017D V011 8 FW7Class 3 weld on other side of valve; no volumetric examination required

RNS-PLW-093 V022 10 FW3Class 3 weld on other sideof valve; no volumetric examination required

RNS-PLW-184 V015A 6 FW1Class 3 weld on other side of valve; no volumetric examination required

RNS-PLW-184 V017A 6 FW8PXS-PLW-185 6 FW11

RNS-PLW-192 V015B 6 SW13 Class 3 weld on other side of valve; no volumetric examination required


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RNS-PLW-192 V017B 6 SW1PXS-PLW-02K 6 FW7

RCS-PLW-01E V002B 8 FW9Class 3 weld on other side of valve; no volumetric examination required

RCS-PLW-01E V012B 8 FW4, FW5

RCS-PLW-01F V013B 8 FW7, SW8

RCS-PLW-01F V003B 8 SW12 Class 3 weld on other side of valve; no volumetric examination required

RCS-PLW-03A V014A 14 FW12, FW13

RCS-PLW-03B V014C 14 FW8, FW9

RCS-PLW-03D V014D 14 FW6, FW7

RCS-PLW-015 V002A 8 FW9Class 3 weld on other side of valve; no volumetric examination required

RCS-PLW-015 V012A 8 FW4, FW5


RCS-PLW-016 V013A 8 SW7, SW8


RCS-PLW-024 V111B 4 FW10, FW11


RCS-PLW-070 V001B 4 FW12Class 3 weld on other side of valve; no volumetric examination required



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CVS-PLW-090 V091 3 SW4 No volumetric examination required on other side of valve

CVS-PLW-532 V090 3 FW8No volumetric examination required on other side of valve

Table 2: Category 2 Wrought Austenitic Valve to Wrought Austenitic Pipe Welds

ISO Number Valve ID Line Size(inch) Connecting Welds Comments



PXS-PLW-01K V016A 8 SW4 Upstream weld

PXS-PLW-020 V016B 8 SW3 Upstream weld

PXS-PLW-01K V017A 8 FW1 Upstream weld

PXS-PLW-02P V017B 8 FW6 Upstream weld

ND-16-1706Enclosure Attachment 1Proposed Alternative VEGP 3&4-PSI-ALT-06 in Accordance with 10 CFR 50.55a(z)(1)

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ATTACHMENT 1: Technical Basis for Proposed Alternative: Pre-service Examination Limitations for Austenitic Class 1 and Class

2 Valve to Pipe Welds in AP1000 Plants

IntroductionFor Class 1 and Class 2 valve to pipe welds, ASME Code Section XI Figure IWB-2500-8requires that the inner third of the wall thickness of the weld be volumetrically examined over an extent that includes the weld and 0.25 inch of base metal on each side of the weld. As such, a portion of the valve base material is part of the required examination volume. ASME Code Section XI, IWA-2200 requires that essentially 100% of the total examination volume be examined. Essentially 100% coverage is achieved when the defined examination volume coverage is greater than 90%.

For austenitic valve to pipe welds the required 100% examination volume coverage cannot be obtained due to a combination of the geometry on the valve end near the weld and the austenitic base and weld materials associated with the weld joint. Current ASME Code Section XI, Appendix VIII qualifications for austenitic piping welds require access to both sides of the weld, which cannot be met.

These inspection limitations are generally consistent with those which exist in the current operating fleet, and are primarily associated with Class 1 and Class 2 austenitic valve (and pump) to pipe welds.

The purpose of this work is to provide the technical basis for a proposed alternative that includes an alternative of the examination volume defined in ASME Code Section XI, Figure IWB-2500-8 for austenitic valve to pipe welds.

The Proposed Alternative for the Pre-Service Examination is summarized as follows:

Eliminate the austenitic valve base material from the required examination volume. Perform a ‘best effort’ outer diameter surface ultrasonic examination of the valve side volume to establish a volumetric examination baseline for future examinations.

There are several complementary bases for this proposed alternative, as will be described in the text to follow. It should be noted that the actual situations described here are not identical to the valve to pipe welds of interest here, but the situation is sufficiently similar that the same arguments are applicable. For example, the materials are the same, the operating history has been equally good, and the conclusion that any flaw indication acceptable by inspection would remain acceptably small throughout service life remains valid. The details of the technical basis are outlined below:


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The first example of a situation where volumetric examinations of cast stainless steel were addressed is in Code Case N-481. This Code case replaced the requirement of a volumetric examination of pump casings with a visual examination. The technical basis for this Code Case will be reviewed in detail, but the key points were that cast stainless steel is highly resistant to corrosion, has low stress due to the cast design, and has not had any issues with degradation during service. A requirement for a flaw tolerance evaluation was added to the Code Case, and many submittals were made, to cover the range of different CASS pump casing designs.

Code Case N-481 was incorporated into the ASME Code at some time in the 1980s, at which time valve bodies were added to the exemption. The NRC insisted that the flaw tolerance requirement be dropped, because the staff had seen enough such evaluations to be convinced of the flaw tolerance of these components. The existing inservice examination rules for pump and valve welds in ASME Section XI are now allowed to bedone visually.

Recently, Code Case N-770 was developed to provide enhanced examination requirements for Alloy 600 and its associated welds, because of observed degradation during service. It is important to note that N-770-1, which has now been made mandatory through 10 CFR50.55a, requires only a ‘best effort’ volumetric examination of the cast stainless steel portions of these susceptible welds. The key arguments of the technical basis are summarized below.

The NRC staff has approved both of these Code documents and has not identified any concerns in this area, even with the recently issued Generic Aging Lessons Learned document for second license renewal (GALL-SLR).

Technical Basis for Code Case N-481The technical basis for Code Case N-481 was based on application of the key aspects of Leak-Before-Break to the reactor coolant pump casings of Westinghouse plants. Since the results were similar for all the different pump models, it was assumed that applications to similar cast structures, such as other pump designs would yield the same results. A summary of thetechnical basis paper [1] is provided in the paragraphs to follow.

In the years preceding the acceptance of Code Case N-481, Section XI of the ASME Boiler and Pressure Vessel Code required examination of reactor coolant pump casing welds as well as pump internal surfaces once during each ten years of plant operation. Category B12.10 (B-L-1) specified a volumetric examination of the pump casing weld(s). Category B12.20 (B-L-2) specified a visual examination of the internal pressure boundary surfaces of the pump casing. One pump in each group of pumps performing similar functions in the system was required to be examined during each ten year interval.


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Both the volumetric and visual examinations required by ASME XI at that time would require disassembly of the pump. The only effective volumetric examination method available for welds in the heavy sections cast stainless steel pump casings was radiography using the miniature linear accelerator (MINAC). In order to perform the radiographic examination the pump must be disassembled completely, including removal of the diffuser adapter and casing adapter. This amount of disassembly is significantly greater than that performed during normal pump maintenance. Because the pump bowl must be dry for installation of the MINAC, a complete core unload is typically required since the reactor coolant system water level must be drained below the level necessary to maintain shutdown cooling for the core. Disassembly, preparation, examination, and reassembly of the pump also resulted in high levels of radiation exposure to maintenance and examination personnel. Utilities performing such examinations reported overall radiation exposures ranging from 35 to 100 Man-Rem.

The remainder of this summary reviewed the safety and serviceability for the class of reactor pump casings. Three dimensional finite element stress analysis results of the dominant pump casing designs have been summarized. Fatigue and fatigue crack growth results have been reviewed. Extensive elastic-plastic fracture mechanics results were performed for large postulated through-wall flaws sized by leak rate calculations to criteria an order of magnitude greater than the detection capacity of the plants. Significant margins in flaw size and load were demonstrated. In short, the leak-before-break approach applied to reactor coolant loop piping systems was applied to pump casings.

The methodology used in the Westinghouse Owners Group study for evaluating the leak-before-break for the pump casings is consistent with that recommended in NUREG 1061, Volume 3 [2]. Note that this methodology was later implemented into the US NRC Standard Review Plan [3]. The methodologies involved:

1. Establishing material properties including fracture toughness values2. Performing stress analyses of the structure3. Review of operating history of the structure4. Selection of locations for postulating flaws5. Determining a flaw size giving a detectable leak rate6. Establishing stability of the selected flaw7. Establishing adequate margins in terms of leak rate detection, flaw size and load8. Showing that a flaw indication acceptable by inspection remains small throughout

service life


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The pump casings are all fabricated from cast stainless steel. Such materials are highly resistant to corrosion issues, and exhibit high fracture toughness values (i.e., JIc and Tmat) in the pre-service product form; however, such materials are subject to thermal aging embrittlement at nominal operating temperatures of nuclear plants. The materials for each pump casing were evaluated for thermal aging embrittlement consistent with the methodology approved by the NRC, and were all found to meet the acceptance criteria.

The pump casings were designed to the ASME code and other requirements in force at the time of fabrication. Design, normal, operating, and faulted applied loads on the nozzles are very conservative when compared to actual service conditions. The stresses are such that code conditions are easily met.

Elastic-plastic fracture mechanics J-integral analyses were made at critical locations. The selection of critical locations was based on fracture toughness, normal stress level and faulted stress level. For example, the six critical locations for the model 93A pump casing are shown in Figure A1-1. For all the pump casings examined, a total of fourteen critical locations were evaluated.

Margin on flaw size was demonstrated by showing stability (i.e., the fracture criteria are met) exists under faulted loading for a flaw twice the size giving the selected leak rate (10 gpm). Margin on load was established by showing stability for faulted loads increased by up to a factor of 1.4 using the flaw size giving the selected leak rate.

To determine the sensitivity of the reactor coolant pump casing to cyclic fatigue loading, fatigue analyses were performed. Two types of fatigue analyses were included, one a conventional ASME Section III evaluation to determine a maximum cumulative usage factor, and the other an ASME Section XI fatigue crack growth analysis assuming various initial flaw sizes. The three dimensional finite element stress analysis results at the worst stressed location were used. Deadweight, pressure and all normal and upset transient loadings were considered.

A cumulative usage factor of approximately 0.15 was found, which is much less than the allowable usage factor of 1.0. Fatigue crack growth was small. A flaw having the maximum acceptable depth allowed by the ASME Code was seen to grow only 0.16 in. for 40 years of service.

In summary, service conditions are such that flaws are not expected to occur throughout the life of the plant. If small flaws are present, they will grow only a small amount throughout the service life. If a flaw should penetrate the wall it will leak at a very detectable rate and will not be unstable thus allowing a shutdown of the plant. In conclusion, successful demonstrations of leak-before-break were provided which are applicable to the primary pump casings of all Westinghouse design PWRs. Since the results were similar for all the different pump models, it was assumed that applications to similar cast structures, such as other pump designs would yield the same results. As the Code Case was applied, it became clear that similar conclusions could be reached for cast valve bodies, and the applicability of the Code Case was expanded to cover valve bodies in the year 2000.


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Adoption of Code Case N-481 into the ASME CodeThe applicability to Westinghouse units was demonstrated formally in [4], which is summarized in the Appendix, paragraph A-1. The NRC issued a formal approval of this report in 1993, and the text of the approval appears in Section A-2 of the Appendix. Several examples of plant-specific applications are shown in Section A-3 of the Appendix.

The general applicability of the Code Case N-481 became clear as it was applied at more and more operating plants. In the year 2000, the Code Case was incorporated into the body of the ASME Code (as reproduced in Table A1-1). However, The flaw tolerance calculation was removed, because multiple flaw tolerance calculations had been completed, and reviewed by NRC, which demonstrated that these components were both flaw tolerant and resistant to degradation by corrosion mechanisms (The NRC representative on the appropriate Code Committee insisted on this action, since he considered the continued review of such calculations an unwarranted expenditure of funds.)

In 2000, valve bodies were treated in the same manner because they were of similar design, and had likewise been trouble-free for their entire operating life. Similar examination requirements for valve bodies were inserted in Table IWB-2500, which is reproduced as Table A1-2. In 2008, this table of examination requirements was also simplified, to have one category for pumps, and a second category for valves. Again, this change was driven by the excellent operating experience in both components.

There was no formal written basis for the 2008 revision, other than the explanation which accompanied the action, which is reproduced below, courtesy of Reedy Engineering:


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In addition to the explanation above, a search was made of INPO and other industry databases, and failed to identify any examples of service degradation of pump casings or valve bodies resulting from corrosion or any other cracking mechanism.

The NRC formally accepted the 2000 Addenda of the ASME Code, through endorsement in 10 CFR 50.55a, in 2002 [5]. Similarly, The NRC endorsed the 2008 Addenda to the 2007 Edition of the Code, publishing the endorsement in 10 CFR 50.55a, in 2011[6]. In both cases, there were no exceptions taken in these portions of Section XI. Since the Code Case had been incorporated in an edition of the Code endorsed by NRC, the Code Case was annulled in 2004.

Code Case N-770: The Latest Treatment of This Issue in Section XITo demonstrate the consistency with which the NRC has treated the issue of examination requirements for Cast Stainless Steel, a review will be provided of the background and basis of this Code Case. The basis was published as a technical paper at the ASME Pressure Vessels and Piping Conference in 2010 [7].

The goal of Code Case N-770, and its first revision, N-770-1, was to increase the examination requirements for Alloy 600, Alloy 82, and Alloy 182 materials, to account for their susceptibility to PWSCC, and to identify appropriate examination requirements after mitigations were performed. In section 3.2 of the technical basis for N-770-1 “Ultrasonic Testing”, the followingstatement is made:

For cast stainless steel items for which no supplement is available in Appendix VIII, the examination volume shall be examined by Appendix VIII procedures to the maximum extent practicable… If 100% coverage of the required volume for axial and circumferential flaws cannot be met, 100% coverage for circumferential flaws (of the susceptible material) is to be met. This is the practical solution for the examination of the weld and buttering of the susceptible material when the base metal is cast stainless steel, or otherwise not completely inspectable. The examination coverage requirements will then be considered to be met.

Code Case N-770-1 was endorsed, and made mandatory by the NRC with the endorsement of the 2008 Addenda to the 2007 Code [6], further confirming the treatment of cast stainless steel materials which started with the original acceptance of Code Case N-481.

Summary and ConclusionsThe issue of examination capabilities for stainless steel, and their implementation into ASME Code requirements, has been recognized for many years. Stainless steel in general, and cast stainless steel in particular is highly resistant to corrosion, and has very high fracture toughness, which counteracts the concern about inspectability. Also, service experience in PWR plants over the past 40 years has been excellent.


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The inspectability issue was first addressed in Code Case N-481, which eliminated the need for volumetric examination of pump casings, and experience was good with this case, so it has been implemented into the ASME code for more than 14 years. The NRC approved the Code Case initially, and also was part of the driving force behind its implementation into the Code, which was endorsed with no exceptions in the Code of Federal Regulations.

Recently, the approach was continued by the NRC’s endorsement of Code Case N-770 as mandatory, again with no exceptions in the area of cast stainless steel examinations. Therefore the basis which has been employed for many years is offered to support the proposed alternative.

Table A1-1: Initial Implementation of Code Case N-481 into Section XI (2000 -2007)


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Table A1-2: Revision of Examination Requirements for Pumps and Valves in the 2008 Edition


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Figure A1-1: Location of the Six Flaws Postulated for Analyses in the Model 93A Pump Casing


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References

1. Kammerdeiner, Greg, “Code Case on Alternative Examination Requirements for Section XI, Division 1, Examination of Pump Casings”, ASME Committee Correspondence, Aug. 18, 1988.

2. “Report of the US Nuclear Regulatory Commission Piping Review Committee: Evaluation for Pipe Breaks” NUREG-1061, Vol.3, November 1984.

3. Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol.52, No.167/Friday, August 28, 1987/Notices, pp. 32626-32633.

4. Witt, F.J. and Petsche, J.F., “Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Nuclear Steam Supply Systems,” Westinghouse Electric Report WCAP-13045, Sept. 1991.

5. NRC endorsement of ASME Code, Section XI, 2000 Addenda: 67FR60520, Sept. 26, 2002

6. NRC Endorsement of ASME Code, Section XI, 2008 Addenda to 2007 Edition: 76FR36232, July 21, 2011.

7. Donavin, P., Elder, G.G., and Bamford, W.H., “Technical Basis Document for Alloy 82/182 Weld Inspection Code Case N-770 and N-770-1,” in Proceedings, ASME Pressure Vessels and Piping Conference, July, 2010.


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ATTACHMENT 2: Compliance with Code Case N-481, and NRC Acceptance Thereof

Code Case N-481 was accepted by the NRC, which led to the submittal of a topical report, WCAP-13045. The highlights of this report are provided in Section A2-1, which follows. The NRC accepted this report in a letter dated April 13, 1993, which is reproduced in Section A2-2below. This letter stated that individual submittals by individual plants were not necessary, butthe documentation of compliance with the Code Case should be maintained at each plant site, available for NRC audit.

Summaries of several such reports are provided in Section A2-3 of this appendix. The use of these reports has also been extended for applicability to the license renewal period from 40-60years, and an example of NRC acceptance in this situation is found in Section A2-4 of this appendix.


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Attachment A2-1: Summary of WCAP-13045, Demonstrating Compliance with N-481 for Westinghouse Pump ModelsA2-1: Summary of WCAP-13045, “Compliance to ASME Code CASE N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems”, September 1991.

WCAP-13045 is a generic integrity evaluation applicable to all Westinghouse design pump casings which demonstrates compliance to ASME Code Case N-481. WCAP-13045 does not provide plant specific pump casing evaluations because it was not found feasible to qualify each plant specifically. Instead, WCAP-13045 provides an enveloping or bounding criteria, whereby, a specific utility need only show that the pump casings of interest fall under the umbrella established in WCAP-13045.

There are eight different models of pumps in the Westinghouse type pressurizer water reactors, Models 63, 70, 93, 93A, 93A-1, 93D, 100A and 100D. Pump Models 63, 70, 93 and 93D have a tangent outlet nozzle while pump Models 93A and 93A-1 have outlet nozzles that are radially oriented. Pump Models 100A and 100D have a general shape somewhat like Model 93 but a radially oriented outlet nozzle like Model 93A. Models 93A and 93A-1 are single casting pump casings with no welds. About 90% of the domestic plants use the Models 93, 93A and 93A-1.The materials for the pump casings are SA351 CF8 and SA351 CF8M for all but three plants which are made of SA351 CF8.

Instead of modeling each different model of pump, a model representative of each of the outlet nozzle configurations were chosen for the 3D finite element stress analyses and fracture evaluation. Model 93A was chosen for the type of pumps with a radial outlet nozzle and Model 93 was chosen for the type of pumps with a tangential outlet nozzle. Models 93 and 93A were chosen because they are the most typically used pump in the domestic plants. Axisymmetric models for the Model 63 and Model 100A pumps were developed which did not contain the outlet nozzles.

The loads used in the analysis were selected to be conservative for a majority of the plants. The primary loads of interest are internal pressure and the force and moment loadings on the nozzles. The highest stresses in the 3D models were found to be in the outlet nozzle crotch region. The surface stress exceeded 50 ksi for the Model 93A pump casing and 40 ksi for Model 93 pump casing. The selections of the locations for the postulated quarter thickness cracks were selected by looking at the stress results for the 3D and axisymmetric models. The selected flaw locations were then used in the plant specific WCAP evaluations, depending on the plants model pump casing. In total, eleven locations on the 3D models and 4 locations on the axisymmetric models were chosen for postulated flaws, due the stress and locations of the welds.

A stability evaluation was performed in order to determine if enough margin is available to meet the stability criteria. The stability analysis completed in WCAP-13045 required re-evaluation for each plant that’s plant specific loads are not bounded by WCAP-13045. The stability analysis was conducted for Models 93, 93A, 63 and 100A at the previously selected critical locations for postulating quarter thickness flaws. Flaws had a 6 to 1 aspect ratio with exceptions in certain pump casing models, where the depth of the crack is taken as one-fourth the nominal casing wall thickness, not one-fourth the distance from the nozzle knuckle to the nozzle crotch. The critical locations for the four different pump casings are shown in the Figures below. The normal and faulted loadings for the pump models are determined to be stable because the stability


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results are lower than the bounding criteria. It can be concluded that flaws postulated in the pump casings per the ASME Code Case N-481, when subject to the normal and faulted loadings outlined in WCAP-13045 are determined to be stable.

In the stability analyses, cracks are postulated at various locations in the pump casings. These postulated cracks are subject to various cyclic conditions the pump casings experience. Therefore the sensitivity of the postulated cracks to cyclic loadings was evaluated as a generic fatigue crack growth analysis. The highest stress location for each pump casing model was chosen for fatigue crack growth. Through the fatigue crack growth analysis, the end-of-period flaw values are well less than the stable flaw sizes established in the stability analyses.

The Westinghouse reactor coolant system primary loops have an operating history that demonstrates the inherent operating stability characteristics of the design. This includes a low susceptibility to cracking failure from the effects of corrosion. Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Contaminant concentrations are kept below the thresholds known to be conducive to stress corrosion cracking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. Thus during plant operation, the likelihood of stress corrosion cracking is minimized.

Overall, there is a low potential for water hammer in the RCS since it is designed and operated to preclude the voiding condition in normally filled lines. Additionally, the operating transients of the RCS primary piping are such that no significant water hammer can occur.

An evaluation of the low and high cycle fatigue loadings was carried out as part of this study. An assessment of the low cycle fatigue was performed in the fatigue crack growth assessment. High cycle fatigue loads in the system would result primarily from pump vibrations. During operation, an alarm signals the exceedance of the vibration limits. Field measurements have been made on a number of plants during hot functional testing. Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold forfatigue crack growth.

It is concluded that the primary loop pump casings of all models are in compliance with Item (d) of ASME Code Case N-481 as long as the plant specific loads are bounded by the loads used in WCAP-13045. If the loads are not bounded the information provided in WCAP-13045 requires to be updated based on the plant specific loads.


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Critical Locations for Postulating Quarter Thickness Flaws


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Attachment A2-2: NRC Acceptance of WCAP-13045


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Attachment A2-3: Example Summaries of plant Specific Reports Demonstrating Compliance with Code Case N-481

A2-3.1: Summary of WCAP-16957-P, “A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casing of Plant A for the License Renewal Program,” March 2009.

The primary loop pump casings for the Plant A Units are the Westinghouse Model 93A design. These pump casings are fabricated from SA351 CF8 (this material has Type 304 stainless steel chemistry) cast stainless steel. In order to ensure the generic WCAP-13045 report was bounding for the Plant A pump casings, the loadings used in WCAP-13045 needed to be compared to the unit specific loads for Plant A. The faulted nozzle loads (i.e., the normal plus safe shutdown earthquake nozzle loads) for the Plant A Units are compared with the Level A screening (i.e., enveloping) faulted loads from WCAP-13045. The Plant A normal loads are compared with the Level C screening normal loads for evaluating the loss-of-load condition in WCAP-13045. The Plant A loads compare to the generic WCAP-13045 loads as follows:

The Plant A normal forces at the inlet and outlet nozzle are not bounded by WCAP-13045. The Plant A normal forces at the inlet and outlet nozzle are larger than the loads in WCAP-13045.The Plant A normal moments at the inlet and outlet nozzle are bounded by WCAP-13045. The Plant A normal moments at the inlet and outlet nozzle are smaller than the loads in WCAP-13045.The Plant A faulted forces, moments and total stresses at the inlet and outlet nozzle are bounded by WCAP-13045. The Plant A faulted force and moment at the inlet and outlet nozzle are smaller than the loads in WCAP-13045.

Since not all the normal loading conditions are bounded by WCAP-13045, a stability evaluation had to be performed in order to determine if enough margin is available to allow the Plant AUnits to meet the stability criteria. The stability analysis completed in WCAP-13045 was re-evaluated using the Plant A loads. The stability analysis for Plant A was conducted at fourcritical locations for postulating quarter thickness flaws in 93A pump casings as described in WCAP-13045. Flaws have a 6 to 1 aspect ratio with exception of the two flaws in the outer outlet nozzle knuckle region 3-93A and 4-93A, where the depth of the crack is taken as one-fourth the nominal casing wall thickness, not one-fourth the distance from the nozzle knuckle to the nozzle crotch. One postulated flaw was a circumferential crack in the inlet nozzle, the location selected being in the high stress region near the nozzle to pipe weld. The second flaw is in a nominal weld location in the casing proper. The last two flaws were selected in the high stress regions of the outlet nozzle associated with the nozzle crotch. The critical locations for the Plant A pump casings are shown in the below figure. The normal and faulted loadings are determined to be stable because the stability results are lower than the bounding criteria. It can be concluded that flaws postulated in the Plant A pump casings per the ASME Code Case N-481, when subject to the normal and faulted loadings are determined to be stable.

In the stability analyses, cracks are postulated at various locations in the pump casings. These postulated cracks are subject to various cyclic conditions the pump casings experience.


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Therefore the sensitivity of the postulated cracks to cyclic loadings was evaluated as a generic fatigue crack growth analysis. The highest stress location, Flaw 4-93A, was chosen for fatigue crack growth. The postulated flaw at the outlet nozzle knuckle is in the axial direction of the pump casing. Through the fatigue crack growth analysis, the end-of-period flaw values are well less than the stable flaw sizes established in the stability analyses. Any reasonably sized flaws in the pump casings will exhibit only minimal crack extension during 40 years, such flaws remaining well below the flaw sizes shown to be stable. Since the transients and cycles for the 60 year plant are the same as the 40 year plant, the fatigue crack growth assessment is applicable to 60 year.



An evaluation of the low and high cycle fatigue loadings was carried out as part of this study. The assessment of the low cycle fatigue was performed in the fatigue crack growth assessment. High cycle fatigue loads in the system would result primarily from pump vibrations. During operation, an alarm signals the exceedance of the vibration limits. Field measurements have been made on a number of plants during hot functional testing. Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold for fatigue crack growth (See the Figure below).

It is concluded that the primary loop pump casings of Plant A are in compliance with Item (d) of ASME Code Case N-481 and applicable for the license renewal period.


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Westinghouse Model 93A Pump Casing Postulated Flaw Locations Used in Plant A Analysis


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A2-3.2: Summary of WCAP-15355, “A Demonstration of Applicability of ASME Code Case N-481 to the Primary Pump Casings of Plant B,” January 2000.The primary loop pump casings for the Plant B Units are the Westinghouse Model 93 design. These pump casings are fabricated from SA351 CF8 (this material has Type 304 stainless steel chemistry) cast stainless steel. In order to ensure the generic WCAP-13045 report was bounding for the Plant B pump casings, the loadings used in WCAP-13045 needed to be compared to the unit specific loads for Plant B. The faulted nozzle loads (i.e., the normal plus safe shutdown earthquake nozzle loads) for the Plant B Units are compared with the Level A screening (i.e., enveloping) faulted loads from WCAP-13045. The Plant B normal loads are compared with the Level C screening normal loads for evaluating the loss-of-load condition in WCAP-13045. The Plant B loads compare to the generic WCAP-13045 loads as follows:

The Plant B normal moments at the inlet and outlet nozzle is bounded by WCAP-13045. The Plant B normal moments at the inlet and outlet nozzle is smaller than the load in WCAP-13045.The Plant B normal forces at the inlet and outlet nozzle is not bounded by WCAP-13045. The Plant B normal forces at the inlet and outlet nozzle is larger than the load in WCAP-13045.The Plant B faulted force and moment at the inlet and outlet nozzle is bounded by WCAP-13045. The Plant B faulted force and moment at the inlet and outlet nozzle is smaller than the load in WCAP-13045.

Since not all the normal loading conditions are bounded by WCAP-13045, a stability evaluation had to be performed in order to determine if enough margin is available to allow the Plant BUnits to meet the stability criteria. The stability analysis completed in WCAP-13045 was re-evaluated using the Plant B loads. The stability analysis for Plant B was conducted at seven critical locations for postulating quarter thickness flaws in 93 pump casings as described in WCAP-13045. Flaws have a 6 to 1 aspect ratio with one exception. The exception is the flaw selected at the outlet nozzle knuckle. An aspect ratio is not defined but the crack front curvature is representative of the crack front curvature for a crack having a 6 to 1 aspect ratio. For the outlet nozzle knuckle, the depth of the crack is taken as one-fourth the nominal casing wall thickness, not one-fourth the distance from the nozzle knuckle to the nozzle crotch. The critical locations for the Plant B pump casings are shown in the below figure. The normal and faulted loadings are determined to be stable because the stability results are lower than the bounding criteria. It can be concluded that flaws postulated in the Plant B pump casing per the ASME Code Case N-481, when subject to the normal and faulted loadings are determined to be stable.

In the stability analyses, cracks are postulated at various locations in the pump casings. These postulated cracks are subject to various cyclic conditions the pump casings experience. Therefore the sensitivity of the postulated cracks to cyclic loadings was evaluated as a generic fatigue crack growth analysis. The highest stress location, Flaw 5-93, was chosen for fatigue crack growth. The postulated flaw at the outlet nozzle knuckle is in the plane of the weld. Through the fatigue crack growth analysis, the end-of-period flaw values are well less than the stable flaw sizes established in the stability analyses. Any reasonably sized flaws in the pump casings will exhibit only minimal crack extension during 40 years, such flaws remaining well below the flaw sizes shown to be stable. Since the transients and cycles for the 60 year plant are the same as the 40 year plant, the fatigue crack growth assessment is applicable to 60 year.


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An evaluation of the low and high cycle fatigue loadings was carried out as part of this study. The assessment of the Low cycle fatigue was performed in the fatigue crack growth assessment. High cycle fatigue loads in the system would result primarily from pump vibrations. During operation, an alarm signals the exceedance of the vibration limits. Field measurements have been made on a number of plants during hot functional testing. Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below thethreshold for fatigue crack growth (See the Figure below).

It is concluded that the primary loop pump casings of Plant B are in compliance with Item (d) of ASME Code Case N-481.


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Westinghouse Model 93 Pump Casing Posulated Flaw Locations Used in Plant B Analysis


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A2-3.3: Summary of WCAP-16377, “A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casing of Plant C for the License Renewal program ,” January 2005.The primary loop pump casings for the Plant C Unit are the Westinghouse Model 93A design. These pump casings are fabricated from SA351 CF8 (this material has Type 304 stainless steel chemistry) cast stainless steel. In order to ensure the generic WCAP-13045 report was bounding for the Plant C pump casings, the loadings used in WCAP-13045 needed to be compared to the unit specific loads for Plant C. The faulted nozzle loads (i.e., the normal plus safe shutdown earthquake nozzle loads) for the Plant C Unit are compared with the Level A screening (i.e., enveloping) faulted loads from WCAP-13045. The Plant C normal loads are compared with the Level C screening normal loads for evaluating the loss-of-load condition in WCAP-13045. The Plant C loads compare to the generic WCAP-13045 loads as follows:

The Plant C normal force and moment at the inlet nozzle are bounded by WCAP-13045. The Plant C normal force and moment at the inlet nozzle is smaller than the load in WCAP-13045.The Plant C normal force at the outlet nozzle is bounded by WCAP-13045. The PlantC normal force at the outlet nozzle is smaller than the load in WCAP-13045.The Plant C normal moment at the outlet nozzle is not bounded by WCAP-13045.The Plant C normal moment at the outlet nozzle is larger than the load in WCAP-13045.The Plant C faulted force and moment at the inlet nozzle is bounded by WCAP-13045. The Plant C faulted force and moment at the inlet nozzle is smaller than the load in WCAP-13045.The Plant C faulted force at the outlet nozzle is bounded by WCAP-13045. The Plant C faulted force at the outlet nozzle is smaller than the load in WCAP-13045The Plant C faulted moment at the outlet nozzle is not bounded by WCAP-13045.The Plant C faulted moment at the outlet nozzle is larger than the load in WCAP-13045.

Since not all the normal and faulted loading conditions are bounded by WCAP-13045, a stability evaluation had to be performed in order to determine if enough margin is available to allow the Plant C Unit to meet the stability criteria. The stability analysis completed in WCAP-13045 was re-evaluated using the Plant C loads. The stability analysis for Plant C was conducted at two critical locations for postulating quarter thickness flaws in 93A pump casings as described in WCAP-13045. Flaws have a 6 to 1 aspect ratio and quarter thickness depths where the depth of the crack is taken as one-fourth the nominal casing wall thickness, not one-fourth the distance from the nozzle knuckle to the nozzle crotch. The two flaws were selected in the high stress regions of the outlet nozzle associated with the nozzle crotch, axially oriented with respect to the casing. The critical location for the Plant C pump casings are shown in the below Figure. The normal and faulted loadings are determined to be stable because the stability results are lower than the bounding criteria. It can be concluded that flaws postulated in the Plant C pump casing per the ASME Code Case N-481, when subject to the normal and faulted loadings are determined to be stable.


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In the stability analyses, cracks are postulated at various locations in the pump casings. These postulated cracks are subject to various cyclic conditions the pump casings experience. Therefore the sensitivity of the postulated cracks to cyclic loadings was evaluated as a generic fatigue crack growth analysis. The highest stress location, Flaw 4-93A, was chosen for fatigue crack growth. The postulated flaw is at the outlet nozzle knuckle in the axial direction of the pump casing. Through the fatigue crack growth analysis, the end-of-period flaw values are well less than the stable flaw sizes established in the stability analyses. Any reasonably sized flaws in the pump casings will exhibit only minimal crack extension during 40 years, such flaws remaining well below the flaw sizes shown to be stable. Since the transients and cycles for the 60 year plant same as the 40 year plant, the fatigue crack growth assessment is applicable to 60 year

It is concluded that the primary loop pump casings of Plant C are in compliance with Item (d) of ASME Code Case N-481 and applicable for the license renewal period.


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Westinghouse Model 93A Pump Casing Postulated Flaw Locations Used in Plant C Analysis


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A2-3.4: Summary of WCAP- 15169, “A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casing of Plant D,” March 1999.The primary loop pump casings for the Plant D are the Westinghouse Model 100A design. These pump casings are fabricated from SA351 CF8 (this material has Type 304 stainless steel chemistry) cast stainless steel. In order to ensure the generic WCAP-13045 report was bounding for the Plant D pump casings, the loadings used in WCAP-13045 needed to be compared to the unit specific loads for Plant D. The faulted nozzle loads (i.e., the normal plus safe shutdown earthquake nozzle loads) for the Plant D Units are compared with the Level A screening (i.e., enveloping) faulted loads from WCAP-13045. The Plant D normal loads are compared with the Level C screening normal loads for evaluating the loss-of-load condition in WCAP-13045. The Plant D loads compare to the generic WCAP-13045 loads as follows:

The Plant D normal force at the inlet nozzle is not bounded by WCAP-13045. ThePlant D normal force at the inlet nozzle is larger than the load in WCAP-13045.The Plant D normal moment at the inlet nozzle is bounded by WCAP-13045. The Plant D normal moment at the inlet nozzle is smaller than the load in WCAP-13045.The Plant D normal force and moment at the outlet nozzle are not bounded by WCAP-13045. The Plant D normal force and moment at the outlet nozzle is larger than the load in WCAP-13045.The Plant D Faulted force and moment at the inlet nozzle are bounded by WCAP-13045. The Plant D faulted force and moment at the inlet nozzle is smaller than the load in WCAP-13045.The Plant D faulted force and moment at the outlet nozzle are not bounded by WCAP-13045. The Plant D faulted force and moment at the outlet nozzle is larger than the load in WCAP-13045.

Since not all the normal and faulted loading conditions are bounded by WCAP-13045, a stability evaluation had to be performed in order to determine if enough margin is available to allow the Plant D Units to meet the stability criteria. The stability analysis completed in WCAP-13045 was re-evaluated using the Plant D loads. The stability analysis for Plant D was conducted at one critical location as described in WCAP-13045. The critical location for the Plant D pump casings was 1-100A, as shown in the below Figure. The crack was chosen at the stress concentration which is representative of the weld location. The normal and faulted loadings are determined to be stable because the stability results are lower than the bounding criteria. It can be concluded that flaws postulated in the Plant D pump casing per the ASME Code Case N-481, when subject to the normal and faulted loadings are determined to be stable.

It is concluded that the primary loop pump casings of Plant D are in compliance with Item (d) of ASME Code Case N-481.


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Westinghouse Model 100A Pump Casing Postulated Flaw Location Used in Plant D Analysis


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Attachment A2-4: NRC Acceptance of Compliance with N-481 for License Renewal


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Key Excerpts From the Report:


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