ML20081E071
| ML20081E071 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 03/10/1995 |
| From: | Coe R, Fleck J, Helmey J BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19325F503 | List: |
| References | |
| BAW-10204, BAW-10204-R01, BAW-10204-R1, NUDOCS 9503210134 | |
| Download: ML20081E071 (177) | |
Text
BWNT 30311 A-8 (4/90)
BGWNUCLEAR BWTECHNOLOGIES LICENSING DOCUMENT APPROVAL File Point:
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Title:
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Response to NRC Questions Document Preparer 4
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For Topical Reports Only is a list of source references required?
O Yes liis No if yes, complete source references on reverse side of this form kk 7 Prepare'rs Preparers initials Managers initials 9503210134 950315 PDR ADOCK 05000498 P
BAW-10204 1
REVISION 01 MARCH 1995 SOUTH TEXAS PROJECT TUBE REPAIR CRITERIA FOR ODSCC AT TUBE SUPPORT PLATES BWNT Non-proprietary B&WNUCLEAR TECHNOLOGIES P.O. BOX 10935 LYNCHBURG, VA 24506-0935
o 2
1 f
i This-document is the non-proprietary version of the proprietary document BAW-10204P-01.
In order for this document'to meet the non-proprietary criteria, certain blocks of information were withheld.
The basis for determining what information to' withhold was based on the criteria listed below.
Depending.upon the application criteria, the criteria codes below represent the withheld information.
I
-(h)
The-use of the information by a competitor would decrease his expenditures,.in time or resources, in designing, producing or marketing a similar product.-
(d)
The information consists of test data or other similar data:
concerning a process, method or component, the application-of which results'in a~ competitive-advantage to BWNT..
(e)
The infirmation reveals special aspects of a process, method, component or the like, the exclusive use of which results in a competitive advantage'to BWNT.
[]" The information contains data that is directly proprietary to BWNT and the corresponding code from above is applied to the information as needed between the brackets.
[]" The information contains data that is directly proprietary to EPRI and HL&P, as a member of EPRI has declared that this information is proprietary to EPRI and therefore was removed.
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TABLE OF CONTENTS I
Page
1.0 INTRODUCTION
1-1 i
2.0 EXECUTIVE
SUMMARY
2-1 3.0 GENERIC. LETTER' APPLICABILITY TO STP 3-1 4.0 GENERIC LETTER EXCEPTIONS FOR STP 4-1 5.0 STP STEAM GENERATOR DESIGN INFORMATION 5-1 6.0 REPAIR LIMITS 6-1 7.0 NDE INSPECTION CRITERIA 7-1 8.0 TUBE REMOVAL AND EXAMINATION / TESTING 8-1 9.0 VOLTAGE DISTRIBUTIONS AND PROJECTIONS 9-1 10.0 PROBABILITY OF BURST 10-1 11.0 EVALUATION OF LEAKAGE 11-1 12.0 OPERATIONAL LEAKAGE LIMITC 12-1 13.0 REPORTING REQUIREMENTS 13-1 I
14.0 EXCLUSION OF INTERSECTIONS 14-1
15.0 CONCLUSION
S 15-1
16.0 REFERENCES
16-1 APPENDIX A - NDE DATA ACQUISITION AND ANALYSIS REQUIREMENTS A-1 i
y f
s r
s o
B&W NUCLEAR TECHNOLOGIES i
1
LIST OF TABLES Pane TABLE 4-1 DIFFERENCES BETWEEN DRAFT GENERIC LETTER 4-2 AND STP ARC SUBMITTAL TABLE 5-1 W-E DESIGN AND OPERATING CHARACTERISTICS 5-3 TABLE 6-1 STP-1 ARC REPAIR LIMITS TO SATISFY STRUCTURAL 6-4 REQUIREMENTS TABLE 8-1 RESULTS OF TUBE PULL EXAMINATIONS ON STP-1 8-5 (A-D) 1993 H/L TSP INTERSECTIONS TABLE 9-1 STP-1 FIELD BOBBIN CALLS AND RE-SIZED CALLS 9-8 FOR DSIs AT H/L TSPs 1993 INSPECTION TABLE 9-2 STP-2 FIELD BOBBIN CALLS FOR DSIs AT H/L 9-11 TSPs 1993 INSPECTION TABLE 9-3 BOUNDING VOLTAGE GROWTH DISTRIBUTION FOR 9-12 DOMESTIC PLANTS TABLE 10-1 EXAMPLE FOR PROBABILITY OF BURST CALCULATION 10-4 TABLE 10-2 STP-1 EOC-5 PROBABILITY OF BURST RESULTS 10-6 TABLE 10-3 VOLTAGE-BURST-LEAK RATE DATA TABLE 10-13 TABLE 10-4 BASIS FOR EXCLUDING DATA FROM THE 3/4 INCH 10-17 BURST CORRELATION TABLE 10-5 COMPARISON OF SAMPLE CALCULATIONS USING 10-19 MONTE CARLO AND LOOK-UP TABLE METHODS TABLE 11-1 PROBABILITY OF LEAKAGE PARAMETERS 11-6 TABLE 11-2 BASIS FOR EXCLUDING INDICATIONS FROM THE 11-15 3/4 INCH PROBABILITY OF LEAK CORRELATION TABLE 11-3 BASIS FOR EXCLUDING INDICATIONS FROM 11-16 LEAK RATE CORRELATIONS TABLE 11-4 INDEX OF DETERMINATION FOR VARIOUS 11-9 SELECTION OF COORDINATE SCALES TABLE 14-1
SUMMARY
OF TUBES TO BE EXCLUDED FROM ARC 14-17 B&W NUCLEAR TECHNOLOGIES il
LIST OF FIGURES Pace FIGURE 1-1
SUMMARY
OF THE MAJOR ASPECTS OF THE STP 1-3 ALTERNATE REPAIR CRITERIA FIGURE 5-1 W-E RSG GENERAL ARRANGEMENT 5-4 FIGURE 9-1 STP-1 C S/G BOC 5 VOLTAGE DISTRIBUTION 9-13 FIGURE 9-2 EPRI VOLTAGE GROWTH DISTRIBUTION 9-14 (FROM TABLE 9-3)
FIGURE 9-3 STP-1 C S/G EOC 5 PREDICTED VOLTAGE 9-15 DISTRIBUTION FIGURE 10-1 BURST PRESSURE vs. BOBBIN AMPLITUDE FOR 3/4 10-20 INCH ALLOY 600 S/G TUBE MODEL BOILER AND FIELD DATA FIGURE 10-2 BURST PRESSURE vs. BOBBIN AMPLITUDE 10-21 RESIDUALS vs. PREDICTED VALUES FIGURE 10-3 BURST PRESSURE vs. BOBBIN AMPLITUDE ACTUAL 10-22 vs. EXPECTED CUMULATIVE PROBABILITY FIGURE 11-1 2650 PSI MSLB LEAK RATE vs. BOBBIN 11-17 AMPLITUDE 3/4" TUBES, MODEL BOILER AND FIELD DATA FIGURE 11-2 RESIDUALS vs. PREDICTED LOG OF LEAK RATES 11-18 3/4" TUBES, MODEL BOILER AND FIELD DATA FIGURE 11-3 CUMULATIVE PROBABILITY OF RESIDUAL LEAK 11-19 RATES FIGURE 14-1 TSP MODEL 14-7 FIGURE 14-2 MODEL ELEMENTS 32 DEGREE WEDGE LOCATION 14-8 FIGURE 14-3 MODEL ELEMENTS 16 DEGREE WEDGE LOCATION 14-9 FIGURE 14-4 TUBE BUNDLE SHOWING TSP ELEVATIONS FOR 14-10 IDENTIFYING WEDGE GROUPS FIGURE 14-5 WEDGE GROUP LOCATIONS: TSPs 1-11 14-11 FIGURE 14-6 WEDGE GROUP LOCATIONS: TSP 12 14-12 FIGURE 14-7 EXCLUDED TUBE REGION: TSPs 1-11(NOZZLE SIDE) 14-13 FIGURE 14-8 EXCLUDED TUBE REGION: TSPs 1-11(MANWAY SIDE) 14-14 l
B&W NUCLEAR TECHNOLOGIES iii
LIST OF FIGURES (CONT.)
Ea92 FIGURE 14-9 EXCLUDED TUBE REGION: TSP 12(NOZZLE SIDE) 14-15 FIGURE 14-10 EXCLUDED TUBE REGION: TSP 12(MANWAY SIDE) 14-16 FIGURE A-1 PROBE WEAR STANDARD SCHEMATIC A-27 FIGURE A-2 BOBBIN COIL AMPLITUDE ANALYSIS OF ODSCC A-28 AT TSP FIGURE A-3 BOBBIN COIL AMPLITUDE ANALYSIS OF ODSCC A-29 INDICATION AT TSP-IMPROPER IDENTIFICATION OF FULL FLAW SEGMENT RESULTING IN REDUCED VOLTAGE MEASUREMENT WHEN COMPARED WITH FIGURE A-2 FIGURE A-4 BOBBIN COIL AMPLITUDE ANALYSIS OF ODSCC A-30 INDICATION AT TSP-IMPROPER IDENTIFICATION OF FULL FLAW SEGMENT RESULTING IN REDUCED VOLTAGE MEASUREMENT WHEN COMPARED TO FIGURE A-2 FIGURE A-5 CORRECT PLACEMENT OF VOLTAGE SET POINTS ON A-31 MIX 1 LISSAJOUS TRACES FOR R18C103 FIGURE A-6 CORRECT PLACEMENT OF VECTOR DOTS ON MIX 1 A-32 LISSAJOUS TRACES FOR R22C40 FIGURE A-7 INCORRECT PLACEMENT OF VECTOR DOTS ON MIX 1 A-33 LISSAJOUS TRACES FOR R18C103 FIGURE A-8 INCORRECT PLACEMENT OF VECTOR DOTS ON MIX 1 A-34 LISSAJOUS TRACES FOR 322C40 FIGURE A-9 INCORRECT MAXIMUM VOLTAGE DERIVED FROM A-35 PLACEMENT OF VECTOR DOTS ON TRANSITION REGION OF 550 kHz RAW FREQUENCY DATA LISSAJOUS TRACE FOR R42C44 FIGURE A-10 CORRECT PLACEMENT OF VECTOR DOTS ON MIX 1 36 LISSAJOUS FIGURE FOR R42C44 FIGURE A-11 PLACEMENT OF VECTOR DOTS BASED SOLELY ON a-37 MIX 1 LISSAJOUS FIGURE (NO SIGNIFICANT SHARP TRANSITIONS IN ANY OF THE RAW FREQUENCIES)
R10C44 FIGURE A-12 PLACEMENT OF DOTS MARKING MIX 1 LISSAJOUS A-38 FIGURE FOR R16C26 B&W NUCLEAR TECHNOLOGIES iv
f LIST OF FIGURES (CONT.)
Page FIGURE A-13 INCORRECT PLACEMENT OF VECTOR DOTS MARKING A-39 MIX 1 LISSAJOUS FIGURE FOR R30C74 FIGURE A-14 CORRECT PLACEMENT OF DOTS TO EFFECT MAXIMUM A-40 VOLTAGE - R30C74 FIGURE A-15 EXAMPLE OF BOBBIN COIL FIELD DATA - MIX A-41 RESIDUAL DUE TO ALLOY CHANGE FIGURE A-16 EXAMPLE OF MRPC DATA FOR SINGLE AXIAL A-42 INDICATION (SAI) ATTRIBUTED TO ODSCC PLANT S FIGURE A-17 MRPC DATA FOR SINGLE AXIAL ODSCC A-43 INDICATION (SAI) - PLANT S FIGURE A-18 MRPC DATA FOR MULTIPLE AXIAL ODSCC A-44 INDICATIONS (MAI) - PLANT S FIGURE A-19 MRPC DATA FOR CIRCUMFERENTIAL ODSCC A-45 INDICATIONS AT DENTED UPPER AND LOWER TSP EDGES FIGURE A-20 EXAMPLE OF BOBBIN COIL FIELD DATA-A-46 FLAW SIGNALS FOR ODSCC.AT DENTED TSP INTERSECTION FROM PLANT A FIGURE A-21 EXAMPLE OF BOBBIN COIL FIELD DATA -
A-47 FLAW SIGNALS FOR ODSCC AT DENTED TSP INTERSECTION FROM PLANT A FIGURE A-22 EXAMPLE OF BOBBIN COIL FIELD DATA -
A-48 i
FLAW SIGNALS FOR ODSCC AT DENTED TSP INTERSECTION FROM PLANT A FIGURE A-23 EXAMPLE OF BOBBIN COIL FIELD DATA -
A-49 FLAW SIGNALS FOR ODSCC AT DENTED TSP j
INTERSECTION FROM PLANT A l
FIGURE A-24 EXAMPLE OF BOBBIN COIL FIELD DATA -
A-50 FLAW SIGNALS FOR ODSCC AT DENTED TSP INTERSECTION FROM PLANT A FIGURE A-25 EXAMPLE OF BOBBIN COIL FIELD DATA -
A-51 FLAW SIGNALS FOR ODSCC AT DENTED TSP INTERSECTION FROM PLANT A FIGURE A-26 LOCATION OF ONE END OF AN INDICATION A-52 USING AN RPC PROBE i
B&W NUCLEAR TECHNOLOGIES v
Il 1
LIST OF ABBREVIATIONS
. ARC
-Alternate Repair Criteria.
ASME American Society of Mechanical Engineers BC..
Bobbin ~ Coil BOC:
Beginning of Cycle BWNT-B&W Nuclear Technologies C/L Cold Leg L
EC
. Eddy Current ECT Eddy Current Testing EDM Electric Discharge Machining EFPM Effective Full Power Month EFPY Effective Full Power Year EOC End of Cycle EPRI Electric Power Research Institute F
-Fahrenheit FSAR Final Safety Analysis Report GDC General Design Criteria' GL
' Generic Letter GPD Gallons per Day GPM Gallons per Minute H/L Hot Leg HL&P Houston Lighting and Power Company ID Inside Diameter IGA Intergranular Attack IN/SEC Inch per Second kHz Kilo-Hertz LBB Leak Before Break LB/HR Pounds per Hour L/HR Liter per Hour-
-LOCA Loss-of-Coolant-Accident LTL Lower Tolerance Limit MB Model Boiler MRPC Motorized Rotating Pancake Coil MSLB Main Steam Line Break NDD No Detectable Degradation NDE Non-Destructive Examination NRC Nuclear Regulatory Commission OD Outside Diameter ODSCC Outside Diameter Stress Corrosion Cracking PCT Peak Clad Temperature POD Probability of Detection POL Probability of Leakage PPM-Parts per Million PSI Pounds per Square Inch PSIA Pounds per Square Inch Atmospheric PSID Pounds per Square 1 Inch Differential PWR Pressurized Water Reactor RCS Reactor Coolant System RG Regulatory Guide RHR Residual Heat Removal RL Repair Limit B&W NUCLEAR TECHNOLOGIES vi
LIST OF ABBREVIATIONS (CONT.)
RSG Recirculating Steam Generator S/G Steam Generator SG Steam Generator SRSS Square Root of the Sum of the Squares SSE Safe Shutdown Earthquake STP South Texas Project STP-1 South Texas Project Unit 1 STP-2 South Texas Project Unit 2 SU Ultimate Tensile Stress SY Yield Stress TS Technical Specifications TSP Tube Support Plate TYP Typical UT Ultrasonic Testing Vu.
Voltage Structural Limit Vu.
Voltage Repair Limit Van NDE Voltage Measurement Error Veo Voltage Growth Anticipated Between Cycles IRE 04 Refueling Outage 4 1RE05 Refueling Outage 5 l
B&W NUCLEAR TECilNOLOGIES vii
1.0 INTRODUCTION
1.1 Purnose The purpose of this document is to provide a technical justification to implement an e.1 ternate steam generator tube repair limit for outer diameter stress corrosion cracking (ODSCC) at the tube-to-tube support plate intersections in the South Texas Project Units 1 and 2 steam generators.
This justification addresses the criteria and guidance contained in NRC Draft Generic Letter 94-xx: Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking (August 1994)(1).
This justification relies on the industry's recommended approach and methodologies, as developed by the Committee for Alternate Repair Limits for ODSCC at TSPs through the Electric Power Research Institute (EPRI). This recommended approach is defined in EPRI Technical Report TR-100407, Revision 2A, "PWR Steam Generator Tube Repair Dimits Technical Support Document for Outside Diameter Stress Corrosion Cracking at Tube Support Plates (6).
1.2 Backcround Stress corrosion cracking initiating on the outer diameter of Alloy 600 steam generator tubes has been diagnosed in the tube support plate (TSP) region of the South Texas Project Units 1 and 2 steam generators, as well as at many other pressurized water reactor (PWR) steam generators throughout the world. If existing tube plugging limits based on crack depth were applied, many tubes would require repair that is unnecessary from either a safety or reliability standpoint.
B&W NUCLEAR TECHNOLOGIES 1-1
To address this issue, the PWR industry, working through EPRI, has developed an approach to define an alternate repair criterion (ARC) that does not set limits on depth of cracks to ensure tube integrity. Instead, this criterion relies on correlating the eddy current voltage amplitude from a bobbin coil probe with the more specific measurement of burst pressure and leak rate.
In turn, these items are related to assuring the structural integrity of the tubes and the safe
-operation of the plant.
Allowing tubes with axial ODSCC to remain in service can be justified based on a combination of enhanced in-service inspection, a repair limit based on eddy current testing voltage, a limit on the number of ARC tubes remaining in service (determined by leakage limits for faulted loads), and a reduced primary-to-secondary allowable leak rate at normal operating conditions.
1.3 Orcanization of Report Each section of this report addresses a different NRC Generic Letter requirement as specified in Reference 1.
The requirements are listed and then STP's compliance with that particular requirement is presented in a manner as to justify the STP results and position on the requirement.
Section 4 contains a Table that summarizes STP's differences with the requirements specified in Reference 1.
Figure 1-1 depicts the major steps involved in developing the ARC for the South Texas Project.
The major requirements for the implementation of the voltage-based repair criteria are shown in this figure and the related section of this report is referenced for each.
B&W NUCLEAR TECHNOLOGIES 1-2
,P' I
e FIGURE 1-1
SUMMARY
OF.THE MAJOR ASPECTS OF THE'STP ALTERNATE _ REPAIR CRITERIA
'[
(c) l$I i
B&W NUCLEAR TECHNOLOGIES 1-3 i
2.0 EXECUTIVE
SUMMARY
This report documents the technical justification for an Alternate Repair Criteria (ARC) for Outer Diameter Stress Corrosion Cracking (ODSCC) indications at Tube Support Plates (TSP) for South Texas Project Units 1&2 steam generators. This assessment demonstrates that a correlation relating tube burst pressure to bobbin voltage and main steam line break (MSLB) leakage to bobbin voltage can be used to conservatively satisfy the Reg Guide 1.121 guidelines for tube integrity at South Texas Project Units 1
& 2.
2.1 Overall Conclusions The South Texas Project Unit 1 pulled tubes (1993 outage) show that the degradation morphology for indications at TSPs can be described as axial ODSCC within the TSP length. The burst test behavior of these indications is consistent with the data base supporting the repair limits of this report.
Application of the generic ODSCC ARC methodology developed through EPRI is appropriate for South Texas Units 1&2 ODSCC flaws, satisfies the NRC Generic Letter on ODSCC ARC, and is consistent with other approved ODSCC ARC methodologies.
2.2 Reauirements for Imolementation of ARC The following requirements for South Texas Project ARC conservatively satisfy Reg Guide 1.121 guidelines for tube integrity:
o Bobbin coil indications having flaw voltages greater than 1.0 volt and confirmed as flaws by MRPC inspection shall be repaired.
o Tubes with bobbin coil indications greater than 1.0 volt may be repaired as an alternative to MRPC i
inspection.
B&W NUCLEAR TECHNOLOGIES 2-1
_ _. _. _. ~
- +
cc i
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U 4
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Bobbin coil indications.having. flaws-greater than (d)"P-o
. volts shall'be repaire'd'independentfof MRPC-confirmation of a flaw.
-Projected leakage for a' postulated steam line--break-o-
^"
event at end of. cycle (EOC) conditions shallLbe'less~
than the bounding leakage necessary to remain within
- applicable dose limits'(10 CFR.100,'NUREG-0800, and GDC 19).
o Projected tube burst probability for a postulated-steam line break at EOC conditions shall be calculated and compared to a. threshold.value of 1 percent (1.0 x 10")
for the most limiting steam generator, o
Tubes identified as subject to significant deformation j
at a TSP under a postulated LOCA + SSE event shall be excluded from application of the ARC at that TSP location.
Inspection Requirements o
The inspection shall include 100% bobbin coil inspection of all hot leg intersections and cold leg-
{
intersections down to the lowest TSP for which'the ARC is to be applied.
1 o
Bobbin coil flaw indications above 1.0 volt and below
)
[d)"' volts shall be inspected by MRPC to evaluate ' for detectable MRPC indications and to support ODSCC as the degradation mechanism.
o Eddy current analysis guidelines shall be equivalent to the requirements given in Appendix A.
Operating Leak Rate Limit o
The normal operating leak rate requiring plant shutdown shall be limited to 150 gpd per steam generator.
j B&W NUCLEAR TECHNOLOGIES 2-2
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'. 3. 0 GENERIC. LETTER APPLICABILITY TO STP 7f 3.1 Introduction
- In the ' Generic ' Letter, the NRC ' describes ' the -information necessary to justify. the use of an. ARC for
- ODSCC at' TSP intersections.- Voltage-based repair criteria are-considered
' applicable'only to indications at support plate intersections share the degradation mechanism is dominantly axial ODSCC with ne significant cracks extending outside the thickness of the support plato.
s For the purposes of the Generic Letter, ODSCC refers to degradation whose dominant morphology consists of cial stress corrosion cracks which occur either singularly or in networks of multiple cracks, sometimes with limited patches of general-intergranular attack (IGA).
Circumferential~ cracks may sometimes occur in the IGA affected regions resulting'in a grid-like pattern of axial and circumferential cracks, termed cellular corrosion.
Cellular corrosion is assumed to be relatively shallow (based on available data.'from tube specimens removed from the field) f transitioning to dominantly i
I
'n depth.
The.
axial cracks as the cracking progresses circumferential cracks are assumed (based on available data).
not to be of sufficient size to produce a discrete, crack-like circumferential indications during g field nondestructive examinations (NDE) inspections.
Thus,.the failure mode of ODSCC is axial and the burst pressure is controlled by the geometry of the most limiting axial crack or array of axial cracks.
It is also assumed for purposes of the Generic Letter that ODSCC is confined to within the thickness of the tube support plate, based on available data from tube specimens removed from the field.
Very shallow microcracks are sometimes observed on these specimens that initiate at locations slightly outside the thickness of the tube support plate; however, these micrueracks are small compared to the cracks within the thickness of the support plate and are too small to produce an eddy current response.
The effects of these B&W NUCLEAR TECHNOLOGIES l
3-1 j
l microcracks is considered as negligible and are considered in ARC analyses.
Confirmation that the degradation mechanism is dominantly axial ODSCC should be accomplished by periodically removing tube specimens from the steam generators and by examining and testing these specimens as specified in Section 4 of the Generic Letter.
The acceptance criteria should ccnsist of demonstrating that the dominant degradation mechanism affecting the burst and leakage properties. of the tube is axially oriented ODSCC.
In addition, results of inservice inspections with motorized rotating pancake coil (MRPC) probes would be evaluated in accordance with section 3.b of the Generic Letter to confirm the absence of detectable crack-like circumferer.tial indications and detectable ODSCC indications extending outside the tube support plate thickness.
3.2 Generic Letter Applicability The criteria in the Generic Letter are only applicable to ODSCC located at the tube-to-tube support plate intersections in Westinghouse designed steam generators. These criteria are not applicable to other forms of steam generator tube degradation, nor are they applicable to ODSCC that occurs at other locations within a steam generator.
The voltage-based repair criteria can be applied only under the following constraints:
(1)
The repair criteria of the Generic Letter apply only to Westingbouse designed steam generators with 1.9 cm (3/4-inch) and 2.2 cm (7/8-inch) diameter tubes and drilled hole tube support plates, (2)
The repair criteria of the Generic Letter apply only to predominately axially oriented ODSCC confined within the tube-to-tube support plate intersection and, (3)
Certain intersections are excluded from the application of the voltage-based repair criteria as discussed in Section 1.b of Enclosure 1 of the Generic Letter.
B&W NUCLEAR TECHNOLOGIES 3-2
7 1
Compliance with tho' Generic Letter Requirements 3.2.1 STP Generic'Le.tter Applicability.
In compliance with Section 2 (1). of. the. Generic
. Letter, STP has 4 Westinghouse Model E. steam
-l generators.
'The tubes are ' O.749"
- 1. 0.005" OD x 0.043" i 0.004" wall mill-annealed nickel-chromium-iron' alloy UNS N06600 -tube: per ASME. material-specification SB-163 (10)..
In compliance.with Section 2(2) of the l Generic i
Letter, STP has confirmed, by pulling tubes from the STP-1 steam generators, that' axially oriented ODSCC exists and is the dominate degradation i
mechanism within the tube-to-tube support plate
{
intersections.
At this point in their operating history,.the STP units have not had a large number j
During their last inspection ontages, the units had [ (d)
)" flaws, i
respectively.
In compliance with Section 2 (3) of the Generic Letter, certain intersections are excluded from the application of the voltage-based repair criteria as discussed in Section 1.b of the Generic Letter.
Section 14.0 of this report discusses the intersections excluded from application of ARC for STP.
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~ B&W NUCLEAR TECHNOLOGIES 3-3
]
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4.0 GENERIC LETTER EXCEPTIONS FOR STP 4.1 Introduction Since the' Draft Generic Letter was issued on August 12,
'1994, there have been several meetings between the1 NRC and-the industry in order to evaluate and. modify the guidelines-provided in NRC Generic Letter 94-xx: " Voltage-Based Repair Criteria for the Repair of Westinghouse Steam' Generator i
Tubes Affected by outside Diameter Stress Corrosion Cracking".
Table 4-1 documents the results of these meetings and summarizes the differences between the requirements of the Draft Generic Letter issued August 12,-
1994 and the STP ARC Submittal.
t
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1 B&W NUCLEAR TECHNOLOGIES 4-1
BWNT PROPRIETARY TABLE 4-1 DIFFERENCES BETWEEN DRAFT GENERIC LETTER AND STP ARC SUBNITTAL No.
Generic Letter Sumsmary of CL Requiressents Summary of SfF Submittal Disposiaan Section(s) i 3(2)
The uPrer and lower voltage repair limits for 3/4" Per 1/18M5 Meeting (17), the lower repair limit will STP win osa en upper sepair lisait of[dl* for tubing are fixed at 1.0 and 2.7.
stay at 1.0 volt for every plant, however, the upper IRE 05.
repair limit will be plant specific. Based on NDE uncertainties discussed in Section 6 of this Report, the upper repair limit for STP was determined to be
[d]" based on a st:uctural repair limit ofldi voka at a pressure of 3790 psi.
2 2.a. I Burm data exclusions should not be excluded Database is consistent with the EPRI [3] database for STP will meet this requirenwns for IRE 05 by specifically under critena 2a & 2b ending NRC 3/4* tubing. Data excluded under criteria I,2a & 2b performing the borst calculation using the EPki P
Staff review, are excluded from the STP database per 1/18M5 database which is consistent with the exclusion Meeting (17].
criteria discussed at the I/I8/95 Meeting [17].
3 2.h.2(2)
Growth rates are to be taken from the most Per Ref.17, if fewer than 200 ODSCC indications Ifinspections during IRE 05 indicate higher limiting of the last two inspections if both were present in prior inspections, the bounding growth rates than EPRi l6], the plant specific ec ployed ARC guidelines or from the current growth rate distribution from EPRI[6] should be growth rate will be used to determine the inspection ifit is the first to employ ARC used. Sirce the growth rates from the be two IRE 06 EOC vohage distribution.
guidelines. Negative growth rates should be inspections were not available and fewer than 200 included as zero growth rates in the assumed indications were found during IRE 04, the EPRI growth distribution.
growth rate discussed in Ref. 6 was used in this report for the IRE 05-EOC coltage growth as discussed in Section 9 of this report.
4 2.b.3(l)
He probability ofleakage (POL) model should Per Ref.17, data ex:Inded under criteria 2a and 2b STP will meet this requirement for IRE 05 by include all applicable da a consistent with industry are acceptable and the Generic Letter will be revised perfwming the probabilig ofleakage recommendations with certain exceptions. Data accordingly. He database for the POL model is calculation using the EPRI database and excluded under criteria 2a and 2b should not be consistent with EPRI[6] and data excluded from the excluding the data specified in the 1/18M5 excluded pending staff review and approval of STP database is discussed in Section 11 of this Meeting (17].
these criteria.
report.
5 2.b.3(2) leak data exclusions should not be excluded under Database is consistent with the EPRI[3] database for STP will meet this requirement for IRE 05 by criteria 2s,2b,3a,3b. & 3c. In addition, MSLB 3/4* tubing. Data excluded under criteria 1 & 2 performing the leak rate calculation using the leak rate of 2496 t/hr for VC Summer Tube from database is acceptable per the I/ ISMS Meeting EPRI database and including the data specified R2RC4I should be included in database.
[l71. The data point excluded under criteria 3 (VC in the IllRM5 Meeting l17l.
Summer Tube R28C41) in EPRI l3l is included in STP database per 1/18/95 Meeting 117].
6 3 b.2, 3.h.3 Tubes with copper deposits at intersections should Copper is not fband in the secondary at STP.
Does not apply to STP.
be MRPC inspected and any indications detected should resuh in tube repair.
B&W NUCLEAR TECHNOLOGIES 4-2
BWNT PROPRIETARY TABLE 4-1 DIFFERENCES BETWEEN DRAFT GENERIC LETTER AND STP ARC SUBNITTAL Mo.
Generic letter Summary of GL Requirements Seminary of STF Submittal DW Swtion(s) 7 3.c.2 Bobbin coil calibration is to be setup on 4-100%
Per 1/18/95 Meeting [17]. new probe variability STP will meet titis requerenseist by fonownig through-wsII holes.
requirernent will be modified to reflect calibration on the ECT Guidelines conessent wish the the 4-20% holes. Appendix A of this report Appendix A requ'nniemen in this Report. -
addresses ECT Analysis Requirements. Inspection Guidelines will address the bobbin coil calibration.
8 3.c.4 All tubes inspected since the last successful pmbe Per 1/18/95 Meeting [171. GL will be enodified to STP will meet this requirement by following wear check shall be reinspected with a new pmbe.
permit afternatives to the method specified in the the ECT Guidelines consistent with the Draft GL, Appendix A of this report addresses Appendix A requirements in this Report.
ECT Analysis Requirements. Inspection Guidelines will address probe wear check.
9 3.c.8 Smaller diana er probes are acceptable if STP plans to use standard 0.610 diameter probes for ne use of the.610 probes has been demonstrated to be statistically equivalent to larger IRE 05.
demonstrated to be acceptable.
probes for detection and response capabilities.
10 4.s Six intersections of tubes should be pulled every Section 8 of this Report addresses tube removal and STP will meet this requirement by pulling 3 other outage or participate in an industry tube pull examination / testing requirements. Per 1/1R/95 tubes 9 intersection from IRE 05 in addition to database.
Meeting (17], Generic lener to be revised to three 12 intersections pulled in 1993.
(3) intersections should be pulled at each second refueling outage (or 36 EFPM whichever is greater) fc'Iowing initial ARC implementation or the initial tube pull, whichever occurred sooner.
II 5.c Tubes with known leaks must be repaired prior to Not addressed in this report STP will comply with this requirement by returning SGs to service.
plugging tubes with known leaks prior to returning the SGs to service.
12 6.b(a) he resuhs of metallurgical examinations of tubes Tube pull resuhs to be submitted f> wing restart.
STP wi!I use best efforts to eneet the 90 day removed are to be reported within 90 days requirement. Depending on extent of results following restart.
expected to be submitted,90 days may not allow sufficient time to obtain results and generate the report.
B&W NUCLEAR TECHNOLOGIES 4-3
5.0 STP STEAM GENERATOR DESIGN INFORMATION 5.1 Introduction The ARC evaluation shall consider as a minimum, the design and operational loading conditions for South Texas Project as summarized in this section, including Figure 5-1 and Table 5-1.
Each unit at STP has 4, Westinghouse Model E steam generators.
The steam generator tubes in both units are 0.749" i 0.005" OD x 0.043" i 0.004" wall mill-annealed nickel-chromium-iron alloy UNS N06600 tube per ASME material specification SB-163 [10].
The tubing in the "as-built" condit10n was not stress relieved in the regions of interest at the TSPs.
5.2 Differences Between Unit 1 and 2 The major differences, with respect to application of ARC, between the steam generators at Unit i versus those at Unit 2 are indicated on Figure 5-1.
The tube support plates at Unit 2 are made of SA-240 type 405 stainless steel while those at Unit 1 are SA-285 Gr. C carbon steel.
Both materials have similar strength characteristics.
The top tube support plate at Unit 2 however, is approximately 50%
thicker than that for Unit 1.
This difference in thickness is important when evaluating LOCA-induced tube deformation near wedge locations, and can be conservatively bounded by using the Unit 1 plate thickness.
At Unit 1, the tubes are roller expanded into the tuba sheet and the tube support plates are carbon steel with drilled holes.
Unit 2 has its tubes hydraulically expanded into the i
tube sheet and has drilled stainless steel support plates.
5.3 Tube Exclusions Based on Analysis Considerations The analysis as required per the NRC Generic Letter, Section 1.b.,
shall consider the effect of SSE and LOCA with respect B&W NUCLEAR TECHNOLOGIES 5-1
461 to the maximum loads that may'be generated in a TSP'and reacted at the wedge locations.
As specified in GDC-4 (52 FR 41288), with NRC acceptance,' leak-before-break (LBB) has
~
been evaluated at STP and thus may be used to determine support plate loads (18,19).
With LBB,.the large. primary
- pipe breaks are eliminated and the next largest' branch pipe break in the primary system, not included in LBB, shall'be considered.
The' bounding branch line LOCA break is a 12" Schedule 140. attachment line.
The use'of LBB at STP:is also consistent with an NRC letter [12] which says that LBB of primary pipingfis acceptable in evaluating internals for steam generator replacements, provided that an assumed break occurs in the branch piping.
With this in mind,Jthe structural analysis to identify tube exclusion areas was performed with the following considerations:
[
n i
(c) b r
1 j ne EiW NUCLEAR TECHNOLOGIES 5-2 j
-I i
l l
TABLE 5-1 WESTINGHOUSE SERIES-E STEAM GENERATOR DESIGN AND OPERATING CHARACTERISTICS [10) i 1
9 9
t h
(c) r B&W NUCLEAR TECHNOLOGIES 5-3
l i
FIGURE 5-1 W-E RSG GENERAL ARRANGEMENT I
(c)
UY B&W NUCLEAR TECHNOLOGIES 5-4
L 6.0 REPAIR LIMITS 6.1 - Jntroduction.
TheLvoltage-based repair limits of the Generic Letter were.
- determined considering the entire range of design basis events.that could challenge tube. integrity.
The. voltage i
repair limits ensure structural integrity and' leakage limits for all postulated design basis events.
The structural criteria are intended to ensure that tubes subjected to the-voltage repair limits will be able to withstand a pressure of 1.4 times ~the maximum possible main' steam line break.
(MSLB) differential pressure postulated to occur at the end of the operating cycle,: consistent with the criteria of Regulatory Guide 1.121.
The induced leakage under worst-case MSLB conditions calculated using licensing basis assumptions will.not result in offsite. dose releases that.
exceed the applicable' limits'of 10 CFR 100.
6.2 Voltace ReDair Limits Der the Generic Letter Per the Generic Letter, the voltage repair limits for 1.9 cm
[3/4 inch) diameter tubes are:
Indications below 1.0 volt as measured by bobbin coil may remain in service; Indications between 1.0 and 2.7 volts as measured by bobbin' coil can remain in service if MRPC inspections do not confirm the indications; and Indications between 1.0 and 2.7 volts as' measured.
by bobbin coil that are confirmed by MRPC and indications exceeding 2.7 volts as measured by' I
bobbin coil must be repaired.
1 i
B&W NUCLEAR TECHNOLOGIES 6-1 i
_-2
.~
d
/
BTP Compilance with Requirements
^
b 6.2.1 Voltage Repair Limit-Methodology' 3
The' Generic-Letter' states that the~ upper andilower repair limit are fixed and not plant specific..
l However, Reference 17 states'that the upper' repair limit should be plant specific and not fixed.
Per-Reference 17,-the lower repair limit of 1~.0' volt ~
for 3/4-inch diameter' tubing will remain fixed.-
l
.However, the upper repair limit will be plant l
specific and the Generic Letter will be modified to include the new requirements.
In compliance with the new modification to the Generic Letter, the tube repair limit methodology described in-this report is conservatively developed to preclude free span tube burst.
The correlation between burst pressure and bobbin voltage amplitude, discussed in Section 10 of this report, i
is derived from the combined model boiler and pulled tube specimens discussed in Reference 3.
The. burst pressure versus-bobbin voltage correlation was adjusted to account for operating.
j temperature and minimum material properties.
To l
establish the voltage structural limit (Vn) that l
satisfies the RG 1.121 guidelines for' margin.
against tube burst,:the burst correlation must be evaluated at the higher of 1.43 times the faulted pressure.or three times the normal operating pressure differential.
For ETP, this value is 1.43 times the faulted pressure, or 3790 psid.
The voltage structural limit must be reduced to allow for NDE measurement error and ODSCC growth between steam generator tube inspections.
l I
B&W NUCLEAR TECHNOLOGIES 6-2
.y
- f
^
5 The voltage repair limit provides margins against tube rupture, consistent with RG.1.121 guidelines, including allowances for NDE measurement error'and.
defect growth, and can be. expressed as.follows:
(
]"
Eq. 6-1 (6) or
[
]"
voltage limit for tube repair, where:
Vn.
=
NDE voltage measurement error, Vmm
=
voltage growth anticipated Vaa
=
between inspections,.and voltage structural. limit from V
=
SL the burst pressure and bobbin voltage correlation (volts).
The value for %Vmm has been determined from available data and is provided in Reference 6 for.
the industry standard.' As discussed in Reference 6, the %Vmm = ()" with the use of a transfer standard.
The value for %Vco has been determined in Reference 6 for the industry standard and is
% Voo = ()"/EFPY.
Therefore, for a cycle length of
[
()" EFPY, %Vaa = ()".
These specific values for NDE uncertainty and voltage growth are utilized in the development of the repair limit voltage only, because they are the most conservative of their i
respective distributions, as presented in i
Reference 6.
The distributions for NDE uncertainty and voltage growth are utilized during a Monte Carlo simulation, in order to project an EOC voltage distribution that is used to determine the probability of burst and leak rate during MSLB conditions, i
B&W NUCLEAR TECHNOLOGIES 6-3
_s g
'HL&P plans to implement the 1 0 volt criterion at' STP as the lower repair limit and a (d]nP-volt criterion-as the upper. repair limit; Table 6-1 i
summarizes the' development of.the ARCLrepair limit.
based on reducing the structural 1 voltage limit of
((d)] volts by allowances for growth and NDE uncertainties at STP.- At the lower []"
.i prediction ~ interval, a bobbin ~ voltage of [(d)]sP volts establishes.the structural requirement for 1.43 faulted pressure (3790 psi) tube burst ~
capability as shown on Figure 10-1.
Equation.10-5 is utilized to determine the exact value of the structural voltage limit, given the value for the pressure.
After adjusting.the structural limit voltage by the allowances for. growth and NDE uncertainties, the resulting equivalent' ARC repair
-i limit is ((d)]BP volts.
The ARC repair limit-is used to define an upper bobbin voltage' limit for leaving unconfirmed bobbin indications in service.
TABLE 6-1 STP-1 ARC REPAIR LIMITS TO SATISFY STRUCTURAL REQUIREMENTS i
[
(d)
} BP B&W NUCLEAR TECHNOLOGIES 6-4
7.0 NDE INSPECTION CRITERIA 7.1 Introduction In order to apply the ARC to the applicable tube support plate (TSP) intersections, the Generic Letter (GL) requires that the bobbin coil inspection guidelines listed below be implemented.
These guidelines ensure that the techniques used to inspect the steam generators are consistent with the techniques used in the development of the voltage-based repair limit methodology.
7 7.2 Bobbin Coil Inspection Scope and Samolina The bobbin coil inspection should include 100% of the hot-leg TSP intersections and cold-leg intersections down to the lowest cold-leg TSP with known ODSCC.
The determination of TSPs having ODSCC should be based on the performance of at least a 20% random sampling of tubes inspected full length.
STP Complianco with Requiroment For Unit 1, Cycle 6, implementation of the TSP ARC requires 100% bobbin coil inspection for all of the H/L TSP intersections and all cold leg intersections down to the lowest C/L support plate with ODSCC indications.
The determination of the TSP intersections having ODSCC indications shall be based on the performance of at least a 20% random sampling of tubes inspected over their full length.
7.3 Motorized Rotatina Pancake Coil (MRPC) Inspection MRPC inspections should be conducted as specified in the GL for the purposes of obtaining additional characterization of the ODSCC flaws found with the bobbin coil insoection and to j
further inspect intersections with significant bobbin interference signals which may influence the bobbin coil
]
measurement or impair the detectablility of an ODSCC flaw.
One of the main reasons for performing MRPC inspections is B&W NUCLEAR TECHNOLOGIES 7-1
K N I
",q-to ensure the absence of detectable crack-like circumferential indications and detectable indications extending beyond the boundary of the TSP. - The voltage-based repair criteria are not applicable to TSP' intersections containing these types of-indications, and special. reporting requirements pertaining to the discovery of such indications are described in Section 6 of the' Generic Letter.
a 7.3.1 MRPC inspection should-be performed for all 1
indications exceedingL1.0 volts as measured by bobbin coil, for 3/4 inch tubing.
7.3.2 The voltage-based criteria of the guidance contained within the GL does not apply to intersections with copper deposits, dent signals greater than 5 volts, and large mixed residuals.
7.3.3 All intersections with bobbin coil signals indicative of-copper deposits should be inspected with MRPC.
Any indication found at this type of intersection should cause the tube to be repaired.
7.3.4 All intersections with dents signals greater than S volts should be inspected with MRPC and any j
indication found should cause the' tube to be i
repaired.
7.3.5 All intersections with large mixed residuals should also be inspected with MRPU.
Large mixed residuals are those that could cause a 1.0 volt bobbin signal to be misread or missed.
Any indication found by MRPC at such an intersection should cause the tube to be repaired.
7.3.6 A minimum of 100 intersections should be inspected with MRPC to further satisfy these requirements.
B&W NUCLEAR TECHNOLOGIES 7-2
t e
STP Compliance with Requirement HL&P will address each.of the requirements listed above, and'
. specified in the Generic Letter, during-steam generator-inspection outages whenLARC will be implemented.
The specificLrequirements for the'MRPC inspection scope.are contained !.n~ Appendix A of this report.
7.4 Data Acauisition and Analysis.
These inspection guidelines are intended to maintain a level of consistency for all plants utilizing an alternate repair criteria.
They ensure that the inspection techniques used
- by the plants are consistent with those used in the development of the tube-integrity methodologies for the-i voltage-based repair limits.
1 7.4.1 The bobbin coil calibration standard'should be i
calibrated-against the reference standard used in-the laboratory in the development of the voltage-based approach by direct testing or through the.
use of a transfer standard.
i 7.4.2 Bobbin coil probes should be calibrated based on i
four 20% through-wall holes, per' Reference 17.
j 7.4.3 Once the probe has been calibrated on the 20%
through-wall holes, the voltage response of new bobbin coil probes for the 20-80% ASME through-wall holes should not differ from the nominal voltage by more than 10%.
l 7.4.4 Probe wear should be monitored by an in-line measuring device or through the use of periodic wear measurement.
When utilizing the wear measurement approach, if a probe is found to be
]
out of specification, all tubes that were~found to be within the repair limit voltage by the extent that the calibration' exceeded the 15% allowance since the last successful calibration, should be B&W NUCLEAR TECHNOLOGIES 7-3 e -,
-m-m------
m,-
r, + - -
e.-,
---i-
+ - - * *
=+~n+
1 e
reinspected with the new calibrated probe, or
~
obtain preliminary voltages with no probe wear.
standard, then-reinspect preliminary voltage; indications ~that are above 75%.of the~ repair limit voltage with a new probe.. Wear measurements should be taken before and,after the reinspection.
7.4.5-Data analysts should be trained and~ qualified in-the use of the guidelines and. procedures.
Analyst performance should be consistent with the assumptions made for analyst measurement variability in Sections 2.b.2(1) of the Generic Letter and-used in'the tube integrity analysis.
(Section 9).
7.4.6 Quantitative noise criteria, that which-results from electrical noise, tube noise, or calibration standard noise, should be included in the data analysis guidelines.
Data that fails to meet the criteria should be rejected and the tube reinspected.
-7.4.7 Data analysts should review the mixed residuals on the standard itself and take action as necessary
.to minimize these residuals.
7.4.8 Smaller diameter probes can be used to inspect tubes where it is impractical to utilize a full-sized probe, provided that the probes and procedures have been demonstrated on a statistical basis to give equivalent voltage response and detection capability when compared to a full size probe.
i STP Compliance with Requirement HL&P will address each of the requirements listed above, and specified in the Generic Letter, during steam generator inspection outages that ARC will be implemented.
The B&W NUCLEAR TECHNOLOGIES 7-4 i
- i.
A) specific requirements pertaining to the items. listed above
+
are contained in Appendix A of this report.
't I
t i
8 t
i 4
I t
. I B&W NUCLEAR TECHNOLOGIES 7-5
F- -
, +
3 p_
S ~
e 8.0 TUBE REMOVAL AND EXAMINATION / TESTING 8.1 Introduction Implementation of voltage-based pluggingferiteria should-include a program of tube removals for testing and examination as described below.
The purpose of this-program is to confirm axial ODSCC as the dominant degradation mechanism atithe TSP intersections and to-provide' additional data to enhance the burst pressure, probability of leakage,.
l and conditional leak rate correlations, as described.in Sections 10 and 11.
8.2 Number and Frecuency of Tube Pulls As stated in the Generic Letter, pulled tube specimens for at least six TSP intersections should be obtained for each plant either during-the plant steam generator inspection.
outage that implements the voltage-based repair limith or, during the inspection outage preceding the initial application of the repair criteria.
The laicst agreement between the industry and the NRC'in Reference 17, however, states that three intersections should be pulled at each second refueling outage or 36 effective full power months (EFPM), whichever is more, following. initial ARC implementation or the initial tube. pull, whichever occurred sooner.
Alternatively, the request to acquire pulled tube specimens may be met by participating'in an industry sponsored tube pull program endorsed by the NRC that meets the objectives of the Generic Letter.
STP Compilance with Requirement During the September, 1993, outage at STP-1, HL&P elected to pull four' tubes from the Unit 1. steam generators (S/G).
Three tubes were pulled from the
'D' S/G and one from the
'C' S/G.
Each of these tubes contained three hot leg TSP intersections, thus producing 12 TSP intersections for D&W NUCLEAR TECHNOLOGIES 8-1 4
laboratory-testing.
HL&P has therefore met this requirement-for STP-1, as specified.
HL&P will continue to pull-tubes from STP-2 in accordance with the requirements outlined in-this report, during future inspection outages.
In addition,-
HL&P plans to pull three additional tubes from STP-1,during' the March, 1995, inspection outage to provide additional ~
information about the flaw ~ morphology at the TSPs.
8.3 Candidate Selection Criteria
+
The selection of tubes as candidates for pulling should.
consider the following criteria:
1.
Tubes with large voltage indications.
2.
Tubes should cover a range of voltages, including
[
intersections with no detectable degradation (NDD).
3.
Selected tube intersections comprising the total data.
set should include at least a representative number of intersections with MRPC signatures indicative of a i
single dominant crack as compared to intersections with MRPC signatures indicative of two or more dominant cracks'about the circumference.
STP Compliance with Requirement The STP-1 1RE04 pulled tubes met theTrequirements specified in the Generic Letter.
During future outages at STP,.HL&P will follow'the guidelines listed above, and specifically in-l Reference 1, for developing a list of tube pull candidates to support the voltage-based repair criteria for axial ODSCC at TSPs.
8.4 Examination and Testina Removed tube intersections should be subjected to leak and burst test under simulated MSLB conditions to confirm that the failure mode and the leakage rates are consistent with that assumed in the development of the voltage-based repair criteria.
Additionally, these data may be used to enhance the supporting data sets for the burst pressure and leakage l
B&W NUCLEAR TECHNOLOGIES 8-2 i..
correlations subject to NRC review and approval.
Subsequent to burst testing, the intersections should be destructively examined to confirm that the degradation morphology is consistent with that assumed for ODSCC.
STP Compliance with Requirement As previously stated, four tubes were pulled from STP-1 during the 1RE04 outage, 1 from the
'C' steam generator, and 3 from the
'D' steam generator.
The tubes that were pulled had various types of defects confirmed through the laboratory tests.
However, only the tubes with axial-oriented ODSCC at TSP intersections are pertinent for the purposes of the ARC implementation.
[
(d)
} BP B&W NUCLEAR TECHNOLOGIES 8-3
During future outages at STP, HL&P will follow the guidelines listed above, and specifically in Reference 1, for developing a list of. tube pull candidates to support the voltage-based repair criteria for axial ODSCC at TSPs.
8.5 General Criteria for Burst and Leakace Models and Suncortina
. Test Data only the use of NRC approved burst and leakage models and
' correlations are to be utilized; this includes the approval of the data that supports the models and correlations.
STP Compliance with Reguirement.
The use of the EPRI burst and leak databases as contained in.
References 2, 3,
and 6 were' utilized in the development of the data correlations for the STP ARC Any modification performed on any of the data came only from NRC requirements on exclusion or inclusion of data points within the correlations [1, 17].
These exclusions and inclusions of data are discussed in Sections 110 and 11 of this report.
l
)
B&W NUCLEAR TECHNOLOGIES 8-4
TAJ!tLE 8-1 (A)
RESULTS OF TUBE PULL EXAMINATIONS ON STP-1 1993 H/L TSP INTERSECTIONS l
4 (d)
DP J
M reviations DSI = Distorted support signal IGA = Intergrannular Attack N/A = Not Inspected NDD = No detectable degradation SAI = Single Axial Indication SVI = Single Volumetric Indication B&W NUCLEAR TECHNOLOGIES 8-5
TABLE 8-1 (B)
RESULTS OF TUBE PULL EXAMINATIONS
'ON STP-1 1993 H/L TSP INTERSECTIONS
[
(d)
} BP Abbreviations DSI = Distorted support signal IGA = Intergrannular Attack N/A = Not Inspected NDD = No detectable degradation SAI = Single Axial Indication SVI = Single Volumetric Indication B&W NUCLEAR TECHNOLOGIES 8-6
r TABLE 8-1 (C)
RESULTS OF TUBE PULL EXAMINATIONS ON STP-1 1993 H/L TSP INTERSECTIONS
[
(d) d f
} BP Abbreviations DSI = Distorted support signal ICA = Intergrannular Attack N/A = Not Inspected NDD = No detectable degradation SAI = Single Axial Indication SVI = Single Volumetric Indication B&W NUCLEAR TECHNOLOGIES 8-7
TABLE 8-1'(D)
RESULTS OF TUBE PULL EXANINATIONS ON STP-1 1993 H/L TSP INTERSECTIONS
[
(d)
) BP Abbreviations DSI = Distorted support signal IGA = Intergrannular Attack N/A = Not Inspected NDD = No detectable degradation SAI = Single Axial Indication SVI = Single Volumetric Indication 1
B&W NUCLEAR TECHNOLOGIES 1
8-8
~. _
.4 t
9.O VOLTAGE DISTRIBUTIONS'AND PROJECTIONS
'9.1
' Introduction In order ~to support the calculation of the conditional-probability of burst and the total leak. rate during MSLB
' conditions, the voltage distributions, beginning:of cycle (BOC) and and of cycle (EOC), must both be developed as part of the ARC evaluation.
9.2 Distribution of Bobbin Indications as'a Function of Voltace-at BOC As stated.in the draft version of the' Generic ~ Letter,'the 1.
frequency distribution by voltage of bobbin indications actually found during inspection should be scaled upward'by-a factor of 1/ POD to account for non-detected cracks which could potentially leak or rupture during MSLB conditions during the next cycle of operation. - The probability of' detection (POD) reflects-the ability of the inspection method to detect all of the ODSCC flaws that exist in the steam generator tubing.
The adjusted frequency distribution minus the detected flaws that have been plugged or repaired constitutes, for the purposes of ARC tube integrity analysis, the assumed frequency distribution of bobbin indications as a function of voltage at-BOC.
This can also be expressed as:
Ng= (1/ POD) x (Nuf.g )-N, g g where Nuc assumed frequency distribution of bobbin
=
i indications at BOC probability of detection of ODSCC flaws POD
=
frequency distribution'of' indications No g.g
=
detected during-the inspection N',,a frequency distribution of repaired
=
indications l
B&W NUCLEAR TECHNOLOGIES 9-1 n
n.
a
+
7---
, -, + -
-y-
POD should have an assumed value of 0.6, or as-an alternative, an NRC approved POD function can be used if available.
STP Compliance with Requirement The STP-1 beginning of cycle voltage distribution has been-determined from the 1RE04 outage bobbin voltage inspection data.
The bobbin coil data obtained during the 1RE04 outage was re-sized prior to this evaluation in order to meet the requirements of and to be consistent with the Eddy Current Inspection Guidelines for the 1REOS inspection.
This reference requires that the bobbin voltage indications be sized on a peak-to-peak voltage basis, not by methods previously used (Maxrate) to size axial ODSCC indications at STP-1.
Table 9-1 contains the "as-found" voltage calls as.
vell as the re-sized bobbin voltage calls for STP-1 1993 inspections.
Table 9-2 contains the STP-2 "as found" bobbin calls only.
The STP-2 data will be re-sized prior to its next inspection outage.
The total number of ODSCC flaws found in all STP-1 steam generators during the last inspection was (d]", while (d]" flaws were found in Unit 2.
These previous inspections included 100% of the tube population in both units.
Evaluations for ARC probability of burst and leakage calculations, include separate analyses for each steam generator.
The approach taken for determining the BOC voltage distribution is consistent with the method outlined above.
The "as-found" flaws were scaled upward by a factor of 1/ POD, to account for the undetected cracks potentially not found during the ECT inspection.
This scaling factor will have an assumed value of 0.6, per the Generic Letter, until a more realistic value of POD has been approved for use.
After scaling the "as-found" voltage distribution, the flaws that were removed from service were subtracted from scaled number, to give the assumed BOC voltage distribution, N,oc to be used in the burst and leakage analyses.
B&W NUCLEAR TECHNOLOGIES 9-2
l l
1 Note that Norma includes all. flaw indications detected by bobbin coil, regardless of MRPC confirmation.
At the time of this submittal, itL&P does not have a plant specific adjustment factor for excluding those bobbin calls that were inspected with MRPC and were determined to have NDD.
Per Reference 12, such a methodology can be implemented upon review and approval by the NRC.
The BOC 5 voltage distribution of the STP-1
'C' steam generator is shown in Figure 9-1.
(
(e)
)nr 9.3 Voltaae Growth Due to Defect Proaression Potential voltage growth rates during the next inspection cycle should be based on growth rates observed during the, last one or two inspection cycles.
For a given inspection, previous results at TSP intersections currently exhibiting a bobbin indication should be re-evaluated consistent with the data analysis guidelines in Section 3 of Reference 1.
In cases where data acquisition guidelines utilized during previous inspections differ from those in Section 3 of Reference 1, adjustments to the previous data should be made to compensate for the differences.
Voltage growth rates should be evaluated for TSP intersections where bobbin indications can be identified at two successive outages.
The distribution of observed voltage growth rates should be determined for each of the last one or two inspection cycles.
When the current or the current and the previous inspections employed data acquisition guidelines similar to those in Section 3 of Reference 1, only the growth rate distribution for the previous cycle should be used to estimate the voltage growth rate expected for the next inspection cycle.
If the Reference 1, Section 3 guidelines were used in both of the two previous inspections, the most limiting of the two growth rates should be used to estimate the voltage growth for the next cycle.
The two distributions should be combined if one or both is based on B&W NUCLEAR TECHNOLOGIES 9-3
~
~]
7
,.j ;
'l
- i k
b
../
1
.i a minimal' number.of indications (i.e., < 200).'
Per-l
-Reference 17, if fewer than 200 ODSCC indications were' present in prior inspections, the'use of a bounding growth rate distribution based on experience 1at similarly designed.
j and operating units is acceptable.
l i
It 1s acceptable to use'a statistical model~ fit.of the
~
observed growth rate distribution:as'part'of'the tube.
[
integrity analysis.
I*.'is also acceptable that'the. voltage j
. growth distribution be in terms of a volts rather than j
. percent A volts, as.long as the conservatism of this j
approach is supported by operating experience.
Negative l
growth rates should be included as zero growth' rates in the-j assumed distribution.
{
STP Compliance with Requirement I
The above method'for evaluating the voltage growth due to
. defect progression was not utilized in the initicl application of ARC at STP (EOC 5 projections).
Since no l
ODSCC TSP defects were found previous to the last outages at j
STP-1 and STP-2, there is not enough data available to construct plant specific growth rates, as described above..
Therefore, the growth rate that will be. utilized for the-purposes of JJu: evaluations, until enough plant specific data is available to develop an accurate and conservative plant specific growth rate and meets the criteria specified above, is the cumulative distribution presented in Table 9-3, from Appendix C of Reference 6.
This growth table is expressed on a per EFPY basis, and the EFPY'used for the j
STP-1 Cycle 5 ARC analysis is [d]" EFPY.
The table is converted into a voltage growth distribution and is shown in Figure 9-2.
i'l i
B&W NUCLEAR TECHNOLOGIES 9-4
3.4 Eddy Current Voltace Measurement Uncertainty
?
Uncertainty in eddy current voltage measurements stems
-primarily from two sources:
1.
voltage response variability (test repeatability. error) resulting from probe wear, and i
2.
voltage measurement variability among data analysts (measurement repeatability error).
Each of these uncertainties should be quantified.
An acceptable characterization of these uncertainties is contained in Reference 6, with the exception that no distribution cutoff should be applied to the voltage measurement variability distribution.
The assumed []"
i cutoff for the voltage response variability distribution in Reference 6 is acceptable.
STP Compliance with Requirnment l
Accuisition and Analyst Errors For the purposes of the STP ARC analyses, the acquisition error that will be utilized will be sampled from (
)",as determined in Reference i
6.
[
(e)
) "'
The analyst error is addressed in a similar fashion as the acquisition error.
It is also represented by [
)" suggested in Reference 6.
[
B&W NUCLEAR TECHNOLOGIES 9-5 l
- q, '
i
~
i (e)-
f
'I
]"
i calibration Bias During the 1993 eddy current inspection of the STP-1 i
Westinghouse Series-E Steam Generators, a transfer standard l
was not used to normalize voltage response, as is normally
.l done during EC inspections.
Because all of the Model Boiler l
and. Pulled Tube voltages (3) were normalized with a transfer standard during the development of the burst and leak l
correlations, the STP data from the 1993 inspection-must be modified by a calibration bias-in order to use the.EPRI correlation lines in those models for the STP ARC analyses.
t Reference 6 states that if a calibration standard is not used, a calibration standard error term should~be included in the combined error due to acquisition and analysis.- 'For
[
7/8. inch tubing with possible differences in calibration hole dimensions of 0.001 inch, the calibration error term is
[
]".
Although no 3/4 inch tubing calibration error is given in Reference 6, a reasonable assumption for the STP-1 ARC analysis for the EOC 5, is to use-the ()"as a bias term which is applied to the BOC 5 voltages before they are
' grown' in the Monte Carlo simulation to the predicted EOC 5 voltages.
However, during future inspections of the'STP steam generator tubing, a transfer standard will be -
)
utilized, thus this error term will be not be needed in l
future EOC STP-1 ARC burst and leak rate analyses.
At that time, the actual error will be accounted for through the ratio in standards, and the ()" bias utilized in the initial STP-1 ARC analyses will no longer be necessary.
B&W NUCLEAR TECHNOLOGIES 9-6
9.5 Proiected End-of-Cycle (EOC) Voltaae Distribution As discussed above, the EOC voltage. distribution is required in order to calculate the conditional probability.of burst and leakage during a postulated MSLB.
In order to project an EOC voltage distribution from the BOC voltage' distribution determined in Section 9.2, the effects of voltage growth to account for defect progression (presented in 9.3), and eddy current voltage measurement uncertainty (presented in 9.4) must be considered.
Monte Carlo techniques are an acceptable means for sampling EC neasurement uncertainty and voltage growth distribution to determine the EOC voltage distribution.
STP Compliance with Requirement In order to project an EOC distribution, a Monte Carlo simulation was performed utilizing the BOC voltage distribution from Section 9.2, the voltage growth cumulative distribution table from Section 9.3, and the ECT uncertainties from Reference 7 and Section 9.4.
The EOC 5 predicted voltage distribution for the
'C' steam generator is shown in Figure 9-3.
B&W NUCLEAR TECHNOLOGIES 9-7
TABLE 9-1' STP-1 PIELD BOBBIN CALLS AND RE-SIEED CALLS FOR DSIs AT H/L TSPs 1993 INSPECTION
[
U i
P (d)
O i
i B&W NUCLEAR TECHNOLOGIES 9-8 i
5
_A TABLE 9-1 STP-1 FIELD BOBBIN CALLS AND RE-BIEED CALLS FOR-DSIs AT H/L TSPs 1993 INSPECTION s
1 (d) 5 NI B&W NUCLEAR TECHNOLOGIES 9-9
TABLE 9-1 STP-1 FIELD BOBBIN CALLS AND RE-SIZED CALLS FOR DSIs AT H/L TSPs 1993 INSPECTION
[
(d)
DP J
b i
c r
B&W NUCLEAR TECHNOLOGIES 9-10 i
... ~... ~...... ~.
... ~.
.,4 t:--
,,a:,
.i TABLE 9-2 STP-2' FIELD BOBBIN CALLS FOR DSIs AT H/L TSPs.
1993 INSPECTION
~
. i
.i I
e b
i (d)'
3 1
f i
f 5
?
l e
i
)
f a
P I
i.
i R
I d-i 4
l 1
I B&W NUCLEAR TECHNOLOGIES 9-11 l
r.ns, s
~..-.,_.n r,.
-. ~.,
i i
.L TABLE 9-3 [6]
BOUNDING VOLTAGE GROWTH DISTRIBUTION FOR DOMESTIC PLANTS
[
r i
(
J 4
i i
EP J
This Table from EPRI TR-100407, Revision 2A.
j B&W NUCLEAR TECHNOLOGIES 9-12
4 EJGURE 9-1 STP-1 C S/G BOC 5 VOLTAGE DISTRIBUTION B&W NUCLEAR TECHNOLOGIES 9-13
E.IGURE 9 _2_
EPRI VOLTAGE GROWTII DISTRIBUTION (FROM TABLE 9-3) l B&W NUCLEAR TECHNOLOGIES 9-14 I
i i
FIGURE 9-3 STP-1 C 8/G EOC 5 PREDICTED VOLTAGE DISTRIBUTION B&W NUCLEAR TECHNOLOGIES l
- ~- - -
i 1
\\
l i
10.0 PROBABILITY OF BURST-10.1 Introduction Calculation of conditional burst probability should be a
performed per the guidance of Section 2.a of Enclosure 1 of the Generic Letter.
This is a calculation to assess the voltage distribution of the indications left in~ service against a threshold.value..
Per the Generic Letter, Licensees should perform an evaluation prior to plant restart to. confirm that the steam generator tubes will retain adequate structural and leakage integrity until the next scheduled inspection.
The first portion of'this evaluation, referred to as the conditional burst probability calculation, assesses the voltage distribution of the axial ODSCC indications left in service 4
against a threshold value of 1 x-10 probability of' rupture under postulated main steam line break (MSLB) conditions.
(
The conditional burst probability calculation is intended to provide a conservative assessment of tube structural integrity during a postulated MSLB occurring at end-of -
cycle (EOC).
It is used to determine whether the NRC needs to focus additional attention on the particular voltage repair limit application.
If the calculated conditional burst probability exceeds 1 x 10, the' licensee should i
4 notify the NRC per the guidance'provided in Section 6 of the I
Generic Letter.
STP Compliance with Requirement i
10.1.1 STP Probability of Burst In compliance with Section 2.a of the Generic Letter, a conditional probability of burst'was l
calculated for South Texas Unit'l at EOC 05.
The most limiting probability of burst (POB) was [ (d) t
]" in the
'B' SG compared to the 1 x 10 l
4 threshold level identified by the staff as not requiring additional evaluation or justification.
i l
B&W NUCLEAR TECHNOLOGIES l
10-1 i
v
I
'I The value was determined in accordance with the methodology described in Section 10.2 which included the probability of detection (POD) factor of 0.6 per 2.b.1 of the Generic Letter.
Section 9 of this report discusses the method used in determining BOC for STP.
Two methods were used-for determining probability of burst.
The first method, discussed in Section 10.2.1, utilizes Monte Carlo simulations in accordance with 2.b.2 of the Generic Letter and EPRI (6) which accounted for growth and ECT uncertainties in arriving at the EOC distribution, as well as determining the POB.
The second method, discussed in Section 10.2.2, utilizes a look-up table for determining probability of burst.
10.2 Conditional Probability of Burst Durina MSLB For the Generic Letter, the conditional probability of burst refers to the probability that the burst pressure associated with 1 or more indications in the faulted steam generator will be less than the maximum pressure differential associated with a postulated MSLB assumed to occur at EOC.
A methodology should be submitted for NRC review and approval for calculating this conditional bur:i probability.
After the NRC approves a method for calculating conditional probability of burst, licensees may reference the approved method.
This methodology should involve (1) determining the distribution of indications as a function of their voltage response at beginning of cycle (BOC), as discussed in Section 2.b.1 of the Generic Letter, (2) projecting this BOC distribution to an EOC voltage distribution based on consideration of voltage growth due to defect progression between inspections, as discussed in Section 2.b.2(2) and voltage measurement uncertainty, as discussed in Section 2.b.2 (1), and (3) evaluating the conditional probability of burst for the projected EOC voltage distribution using the correlation between burst pressure and voltage discussed in Section 2.a.1.
The solution methodology should account for uncertainties in voltage measurement (Section 2.b.2(1)) the B&W NUCLEAR TECHNOLOGIES 10-2
i i
'l distribution of potential voltage growth rates applicable to each indication (Section 2.b.2(2)), and the distribution of I
potential burst pressure as a function of voltage (Section
- 2. a.1).
Monte Carlo simulations constitute an acceptable approach for accounting for these various sources of uncertainty.
STP Compliance with Requirement In compliance with 2a items (1) and (2) of the Generic Letter, a BOC was determined and this BOC was projected to an-EOC voltage distribution as discussed in Section 9 of this report.
In compliance with 2a, item (3), of the Generic Letter for developing a methodology of determining the probability of burst, two different methods were considered.
The first method uses a Monte Carlo simulation-for calculating the final probability of. burst.
The second method will be used on-site as a means of estimating the
. probability of burst as the eddy current inspection progresses.
These methods are discussed below.
10.2.1 Monte Carlo Method This method utilizes the EOC voltage distribution as discussed in Section 9 of this report.
The burst probability simulation then calculates a burst pressure for each EOC voltage by applying each voltage to the burst pressure vs. voltage model.
[
(e)
} DP B&W NUCLEAR TECHNOLOGIES 10-3 L
(
g Y
(.
-[
(e) t
)se The results of the STP-1 probability: of burst evaluation are sumitsrized in Table 10-1 below.
These results were obtained by running the Monte Carlo simulation for 1. x 10' trials.
TABLE 10-1 STP-1 EOC-5 PROBABILITY OF BURST RESULTS I
i l
(d) i
} DP B&W NUCLEAR TECHNOLOGIES 10-4
~
l l
10.2.2 Look-Up Table Method' l
A relationship between probability of burst at EOC and a' specific BOC voltage was developed [
i l
(c)~
Jar The simulator ran 1 million trials for each voltage range.
The advantage of this method is that the time-consuming Monte Carlo simulations can be run prior to the outage, thus avoiding delays during the outage.
The results from the Monte Carlo simulations were used to develop a table containing the probability of burst for each voltage range of interest.
The pr oability that one or more flaws will burst can tt 1 be calculated by applying the following i
eqt tion.
(e)
Eq. 10-1 BP J
B&W NUCLEAR TECHNOLOGIES 10-5
10.2.3 Comparison Of Both Methods For~ Calculating POB A comparison of the results obtained using the two methods was' performed to justify the results from the look-up table.
A BOC voltage distribution containing indications at_3.0, 4.0, and 5.0 volts as shown in Table 10-2 was used for this comparison.
For the purpose of this comparison, a cycle length of [(d)]8P EFPY was used.
TABLE 10-2 EXAMPLE FOR PROBABILITY OF BURST CALCULATION
[
(d)
] 8' Eq. 10-2 Using the BOC voltage distribution from Table 10-2 and a cycle length of [(d)]8PEFPY, the Monte Carlo method resulted in a [d}'Pupper confidence limit of
((d))nr Therefore, the results of the look-up table are approximately five percent higher than the results from the Monte Carlo analysis.
Despite the good correlation between these two methods, the look-up table will only be used on site as a quick method of estimating the POB during the outage.
The Monte Carlo method will be B&W NUCLEAR TECHNOLOGIES 10-6
i 1
l I
l used to calculate the final reportable results.
The results of the comparison between the two methods is_shown in Table 10-5.
10.3 Burst Pressure Versus Bobbin Voltaae Per the Generic Letter, an empirical model, for 3/4-inch diameter tubing, should be used to relate burst pressure to bobbin voltage response for purposes of estimating'the conditional probability of burst during a postulated MSLB.
This model should explicitly account for burst pressure uncertainty as indicated by scatter of the supporting test data and should also account for the parametric (i.e., slope and intercept) uncertainty of the regression fit of the data.
The supporting data sets for 3/4-inch diameter tubing should include all applicable data consistent with the industry recommendations.
Specifically, data excluded under criteria 2a and 2b should not be excluded pending staff review and approval of these criteria.
STP Compliance with Requirements 10.3.1 Burst Pressure - Voltage Correlation Database To comply with Section 2.a.1 of the Generic Letter a correlation between burst pressure and bobbin voltage amplitude was developed by EPRI(3].
The relationship between burst pressure and voltage is used to establish a voltage threshold to ensure that the structural requirements of Regulatory Guide 1.121 are satisfied during normal operating and postulated accident loading conditions.
The correlation for the 3/4-inch diameter tubing is based on[]" model boiler specimens and []BPpggggg, tube specimens from operating plants as summarized in Table 10-3.
The data from teuts of pulled tubes and model boiler specimens were combined to form an aggregate database which was then used to develop the burst pressure correlations described later.
B&W NUCLEAR TECHNOLOGIES 10-7
]
a-The database used forithe~ development of the' burst' correlation. (Burst Strength ins. Bobbin Coil
- Voltage Amplitude) for.3/4 inch diameter tubing is-derived from model boiler specimens and pulled tubes.
All of-the data were derived from Alloy 600 tubing with 3/4 inch OD and 0.043 inch nominal wall thickness.
The model boiler test results for 3/4 inch. tubing are described in detailtin 1
Reference 3.
on the basis of Reference 3, exclusion of data
-j criteria'l-3, all data were reviewed by EPRI for identification of data which were excluded from the databases supporting the ARC correlations for ODSCC.- Table 10-4 summarizes the data-removed.
{
from-the burst database by EPRI [3].
Per Reference 17, data excluded under criteria 1, 2a, and 2b are acceptable exclusion criteria and the Generic Letter will be revised accordingly..The data excluded from the EPRI database-was reviewed by.BWNT for this evaluation.
BWNT concurs that the excluded data meets the criteria. contained in Appendix C of Reference 3 and therefore was not included in this evaluation.
10.3.2' Burst Pressure Correlation for 3/4 Inch' Diameter-Tubing The bobbin coil voltage amplitude and burst pressure data of Table 10-3 were used to determine a correlation between burst pressure and bobbin voltage amplitude. The data considered are shown in Figure 10-1 along with the results of the correlation analyses.
N B&W NUCLEAR TECHNOLOGIES 10-8
m
. ~
\\)l ~
m
'i From the; methodology discussed in Section 6 of.
o Reference'3, the correlation line for the burst -
voltage correlation is:given by:
[
.i i
3" l
'Eq10-3 (3) where the burst pressure i:s measured in kai and I
the bobbin amplitude is measured in volts.
The correlation line from the equation above is shown.
in Figure 10-1.
Per the Generic Letter, Section 2.a.1, the burst pressure'model should account for data scatter of the' linear regression fit.
Therefore, the residual values were plotted against the predicted-burst pressures.
The results, shown in Figure 10-2, indicate ~a scatter about'a mean of zero for the full range of predicted burst pressures.
Per
]
Reference 3, this scatter for the data' indicates l'
an acceptable set. of. data. points and au) acceptable correlation.
A cumulative probability plot of the ordered residuals was also prepared and is shown in Figure 10-3.
Verification offthe regression is obtained
[
by plotting the ordered residuals on normal probability paper.
Since the data form a straight line, it has been shown that the distribution of l
the residuals is normal, as proven in Reference 3.-
In order to' determine the voltage structural limit (Vg)for STP, the ()" prediction bound was reduced to a level corresponding to the [
]"
lower tolerance limit (LTL) for the material properties i
?
~
B&W NUCLEAR TECHNOLOGIES 10-9 i
d of the tubes.
The estimated standard deviation of the residuals, i.e.,.the. error of the estimate, S,,
of the burst pressure was [ )" ksi.
The lower
()" prediction curve adjusted for-[
)" lower tolerance limit on material properties, is defined by the equation:
E Eq. 10-4
[6]
)"
A Monte Carlo simulator is used to predict EOC voltages as an input to the probability of burst model.
[
(e)
B&W NUCLEAR TECHNOLOGIES 10-10
m
]"' The predicted probability of burst is then
-calculated from the simulation.
In order to determine the repair limit (Section 6) for STP, the voltage structural limit at MSLB conditions is calculated utilizing Equation 10-4.
A second order fit of bobbin amplitude to differential pressure for the ()" prediction curve, as adjusted by the LTL material flow stress, (Equation 10-4) was performed for the purpose of determining lower bound voltage amplitudes as a function of the applied pressure differential.
[
)"
Eq. 10-5 (3)
Where:
Vut = Voltage at Lower Tolerance Limit P = Pressure at MSLB Using this Equation, the Vst corresponding to 1.43 times faulted differential pressure, or 3.790 ksi, is ((d))s' volts.
The 3.790 ksi pressure differential is 1.43 x APun3, where APun3 is equal to 2.650 ksi.
The lower ()" prediction bound for LTL material property is shown in Figure 10-1.
B&W NUCLEAR TECHNOLOGIES 10-11
l
~
10.4 South Texas Project Unit.1 Data from Pulled Tube During the 1RE04 refueling outage, STP-1 pulled one tube 4
containing an axial ODSCC-type indication.in the TSP region.
The testing procedure and pulled tube burst test data from South Texas is described in detail in Reference 4.
The burst test result from the STP-1 pulled tube is shown in
~
Figure 10-1.
The re-sized voltage call of [ (d) ]BP. volts is' adjusted to ((d)]a' volts based on the calibration standard error'of ()", as discussed in Section 9.4 The'STP-1 pulled tube containing the TSP ODSCC axial indication had a yield strength.(Sy) of (d)" ksi and an ultimate tensile strength (Su) of [d)"ksi (Su+sy=[d)"ksi).
Therefore, STP-1 pulled tube burst pressure was normalized to an the EPRI model by using an average tensile property of
[]" kai for this evaluation. The 1993 STP-1 pulled ' tube voltage was also normalized for this evaluation using the standard calibration error of ()" as discussed above.
As shown in Figure 10-1, the pulled tube data from STP-1' falls within the correlation of the model boiler data and pulled tube results from EPRI [3), with*a normalized burst pressure of (d)" psi, corresponding to a normalized voltage of [d)"
volts.
B&W NUCLEAR TECHNOLOGIES 10-12
-l l
J TABLE 10-3f3,61 VOLTAGE-BURST-LEAK RATE DATA TABLE I
JEP B&W NUCLFAR TECHNOLOGIES 10-13
TABLE 10-3[3,6]
VOLTAGE-B'URST-LEAK RATE DATA TABLE I
EP B&W NUCLEAR TECHNOLOGIES 10-14
TABLE 10-3[3,6]
VOLTAGE BURST-LEAK RATE DATA TABLE I
EP B&W NUCLEAR TECHNOLOGIES 10-15
TABLE _1MI3,6]
VOLTAGE-BURST-LEAK RATE DATA TABLE I
1 e
EP B&W NUCLEAR TECHNOLOGIES 10-16
4
./'
9 ;;; ? -
TABLE 10-4 [3]
BASIS FOR EXCLUDING DATA FROM THE 3/4 INCH V-BURST CORRELATION l
b A
O e
,)--
t p
i l
B&W NUCLEAR TECHNOLOGIES 10-17
4_.l_'
b i
I TABLE 1H [3]:-
BASIS FOR EXCLUDING DATA FRGM THE 3/4 INCH BURST CORRELATION I
I i
i L
b p
3.
t 1
jte This Table Om EPRI NP-7480-L, Vo:ume 2.
B&W NUCLEAR TECHNOLOGIES 10-18 v
r a
4
% i TABLE 10-5 COMPARISON OF SAMPLE POB CALCULATIONS USING MONTE CARLO AND LOOK-UP TABLE METHODS i
(d)
} DP i
P B&W NUCLEAR TECHNOLOGIES 10-19
FIGURE 10-1 [3]
BURST PRESSURE vs.-BOBBIN AMPLITUDE FOR 3/4 INCH ALLOY 600 S/G TUBE MODEL BOILER AND FIELD DATA
[
(d)
I
)se Data in this Figure derived from EPRI NP 7480-L, Volume 2, with exception of STP-1 pulled tube data point.
l l
B&W NUCLEAR TECIINOLOGIES 10-20 o
1 p-FIGURE 10-2'[3]
BURST PREESURE vs. BOBBIN AMPLITUDE RESIDUALS vs. PREDICTED VALUES l
t t
t i
L Data in this Figure derived from EPRI NP 7480-L, Volume 2.
i B&W NUCLEAR TECHNOLOGIES i
10-21
1 4
FIGCTE 10-3 [3]
BURST PRESSURE vs. BOBBIN AMPLITUDE ACTUAL vs. EXPECTED CUMULATIVE PROBABILITY
[
4 r
4 3"
Data in this Figure derived from EPRI NP 7480-L, Volume 2.
1 r
t B&W NUCLEAR TECHNOLOGIES 10-22
1
=,
11.O EVALUATION ~OF LEAIAGE 2
l 11.1' Introduction
. Calculation'of leakage should be performed per the guidance-of-Section 2.b of Enclosure 1 of the Generic Letter..This calculation, in conjunction with the use of licensing basis assumptions for calculating offsite releases,-enables licensees to demonstrate that the applicable limits of 10-l CFR 100 continue to be met.
l 5
Per the Generic Letter, a methodology should be submitted for calculating the total primary-to-secondary leak rate in the faulted steam generator during a postulated MSLB assumed l
to occur-at EOC.
This methodology involves' (1). determining '
the distribution of indications as a function of their voltage response at beginning of cycle (BOC) as discussed in Section 2.b.1, (2) projecting this BOC distribution to an EOC voltage distribution based on consideration of voltage l
growth due to defect progression between inspections as discussed in Section 2.b.2(2) and' voltage measurement uncertainty as discussed in. Section 2.b.2 (1), and (3) evaluating the total leak rate model as discussed in Section f
2.b.3(2).
The solution methodology should account for uncertainties in voltage measurement (Section 2.b.2 (1)), the distribution of potential voltage growth rates applicable to each indications (Section 2.b.2(2)), the uncertainties in the probability of leakage as a function.of voltage (Section 2.b.3(1)), and the distribution of potential conditional
[
leak rates as a function of voltage (Section 2.b.3(2)).
Monte Carlo simulations are an acceptable method for accounting for these sources of uncertainty provided that the calculated total leak rate reflects an upper 95%
quantile value.
This portion of the tube integrity evaluation is intended to-assure that the total leak rate from the affected steam generator (SG) during a postulated MSLB occurring at EOC would be less than that which could lead to radiological releases in excess of the licensing basis for the plant.
If B&W NUCLEAR TECHNOLOGIES 11-1 n
~
,n_.
?
i calculated' leakage exceeds the allowable limit determined by the licensing basis dose calculation, licensees canfeither repair tubes, beginning with the largest voltage indications until the leak limit is met, or. reduce reactor coolant system specific iodine activity.
i 11.2 Calculation of Proiected MSLB Leakaae Using the projected EOC voltage distribution as calculated from the method presented in Section 9,-
the' leakage for the postulated MSLB is calculated utilizing.the EOC voltage distribution and the use of two models: _(1)- the probability of leakage model and (2) the conditional leak rate model.
As'previously discussed in Section 2.b of the Generic Letter, Monte Carlo Techniques are an acceptable approach for accounting for uncertainties implicit in these models.-
bTP Compliance with Requirements 11.2.1 Calculation of Projected MSLB Leakage for STP In compliance with Section 2.b.3 of the Generic Letter, the probability of leakage correlation and the conditional leak rate correlation provide the basis for a best-estimate ODSCC leak rate model.
The methodology described in Section 11.3 and 11.4 of this report is consistent with that prescribed by EPRI (6).
The probability of leakage (POL) model for STP was developed using field and laboratory data provided by EPRI (3,6).
Due'to a number of uncertainties in the. input variables to the ODSCC leak rate model, the leak rate for-an individual tube may deviate from the value predicted by this correlation.
I i
+
b b
B&W NUCLEAR TECHNOLOGIES 11-2 t
I
~.
/
i A Monte Carlo simulator is_used to predict EOC voltages.as an input into the probability'of leak model; accounting for its; uncertainty as' treated in Appendix D of Reference 6.
[
i 9
(e) p
] 8' 1
[
F (e) i k
J a' An estimate of the total leak rate for a steam generator is obtained for each Monte Carlo trial.
Each total leak rate is written to a file and is later.used to establish the desired statistical limit for the total faulted steam generator leakage.
At the end of the last Monte' Carlo trial, the data in the file'containing the individual total leak l
rates for the steam generator isEsorted, and a
[(d)] one-sided upper-tolerance limit is determined on a distribution-free basis.
ThisLis j
done to ensure a conservative' measure of the leak l
l rate for the faulted steam' generator under MSLB l
conditions.
B&W NUCLEAR TECHNOLOGIES 11-3 t
l
+--
I 1
~
The [(d)]ar one-sided' upper-tolerance limit for' j
leakage calculated during MSLB at EOC.05 at STP-1 is approximately.[
(d)
]
for steam-
~
~
generator _(SG
'C').
For an accident, this leakage must be evaluated in the form of offsite dose release per 2.b.4 of the Generic Letter and is addressed in Section 11.5 of this report.
11.3 Probability of Leakaae as a Functionlof Voltaae I
The probability of leakage (POL) model should utilize the log-logistic functional form [1].
This model should explicitly account for. parameter. uncertainty of the POL functional fit of the data.
The supporting data sets for 3/4-inch diameter tubing should include all applicable data consistent with industry recommendations, with certain exceptions.
Namely, data excluded under criteria 2a and 2b should not be excluded pending staff review and approval of these criteria.
STP Compliance with Reguirements 11.3.1 Probability of Leakage Model Per the Generic Letter, once the EOC distribution has been determined, the probability of leakage (POL) needs to be evaluated.
Probability of leakage is determined from the magnitude of the bobbin coil voltage measurement for a specific TSP intersection.
For a number of TSP intersections containing equal amounts of degradation, based on voltage, a proportion is statically predicted to be leakers, while the remaining. proportion is
~
predicted to be non-leaking.
Based on industry test data, the logistic function best represents the probability of leakage (POL) i and therefore was used for the leak rate model
[6).
This probability distribution-function is f
I appropriate for binary-type variables, such as the B&W NUCLEAR TECHNOLOGIES 11-4 a
leak /no-leak designation.- The POL for an ODSCC.
produced voltage measurement at a specific TSP
" ^
location is.therefore determined by:
(
Eq. 11-2 l
(6)
-i i
~
1
]"
l The parameters of the logistic function are.
determined from an iterative maximum likelihood procedure which is applied to the leak /no-leak data.
i l-A probability of leak model was developed using l
the method described in Appendix D of Reference 6.
A logistic regression fitting procedure'from Reference 24 was applied to the data mt shown in Table D-1 of Appendix D of Reference 6.
l On the basis of the Reference 3 exclusion criteria 1-3, all data documented in-Reference 3 were reviewed by EPRI for identification of data which were excluded from the databases supporting the ARC correlations for ODSCC..BWNT excluded data on C
the' basis of Criteria 1 and 2 in this analysis
)
based on review of the data and the requirements of the Generic Letter and Reference 17.
The data excluded from the probability of leakage correlation are summarized in Table 11-2 of this report.
B&W NUCLEAR TECHNOLOGIES 11-5
>w w
m
-m
-e-a,-
e
,n--
-,-r-
--r-e
o.
t.
F The resulting. parameter estimates for the POL model are depicted in Table 11-1, below.
The-factors needed to estimate the uncertainty for the POL model were computed using-the method shown in
~
Reference 6.,-Appendix D.
The values for the
^
variance-covariance estimates are also listed.
below.
These. values compare well,with those calculated by EPRI in Appendix D of Reference 6.
I TABLE 11-1 PROBABILITY OF LEAKAGE PARAMETER 8 E
i e
(c)
) pr i.
where:
no = coefficient determined from analysis 3:
' coefficient determined'from analysis t
=
Pn = variance-covariance value determined-l from the analysis Pn = variance-covariance value determined l
from the analysis Pu = variance-covariance value determined from the analysis P
11.4 conditional Leakace Rate under MSLB Conditions s
The conditional leak rate model should incorporate a linear regression fit to the log of the leak rate data for 3/4-inch diameter tubing, as a function of the log of the bobbin voltage and should account for both data scatter and 3
parameter uncertainty of the linear regression fit.
Use'of B&W NUCLEAR TECHNOLOGIES 11-6
n this approach is subject to' demonstrating that the linear regression fit is valid at the 5% level with a "p-value" test.' 'If this. condition is not satisfied, the. linear regression fit should be assumed to have zero slope-(i.e.,
the linear regression fit should'be assumed to be constant 1
with_ voltage).
The supporting data set for 3/4-inch diameter tubing should
-include all applicable data consistent with the industry recommendations, respectively with certain exceptions.
Specifically, data excluded under criteria 2a, 2b,H3a, 3b, and 3c should not be excluded pending staff review and approval of these criteria.
In addition, an MSLB leak rate of 2496 1/hr should be utilized for the data point obtained from V.C.
Summer tube R28C41'pending staff review.and approval of the revised leakage estimate for this tube.
STP Compilance with Requirements i
11.4.1 Conditional Leak Rate versus Voltage Model Database j
a' In compliance with 2.b.3(2) of the Generic Letter, the database-used for the development of the leak rate correlation (MSLB Leak Rate versus-Bobbin Coil Voltage Amplitude) for 3/4 inch diameter tubing is derived from model boiler specimens and pulled tubes.
All of the data were derived'from Alloy 600 tubing with 3/4 inch OD and 0.043 inch nominal wall thickness.
The model boiler test results for 3/4 inch tubing are described in detail in Reference 3.
The database used for the Leak Rate - Voltage correlation is shown in Table 10-3 of this report.
i
]" The Generic Letter l
B&W NUCLEAR TECHNOLOGIES 11-7 S 3 mr e--
v
--a-m-
u-,,-
\\"- --- -- ------- -- ^
(Section;2.b.3(2)) states that a MSLB' leak rate of' 2496 1/hr should be utilized for the data point
~
obtainedLfrom V.C. Summer Tube R28C41.
For the STP leakage analysis, the MSLB leak rate of'2496 1/hr for Plant S Tube R28C41 was included in the correlations.
On the. basis,of the Reference 3 exclusion criteria 1-3, all data documented in Reference 3 were
-reviewed by.EPRI for identification of data which
'were excluded from the databases supporting the ARC correlations for ODSCC.
BWNT' excluded data on the basis of criteria 1 and 2 in this analysis based on review of the data and the requirements of the Generic Letter.
The data. point excluded by-EPRI in Reference 3 on the basis of Criteria 3
[
)" for the leak' rate.
correlation was not excluded from this evaluation-based on the requirements of the Generic Letter and Reference 17.
Table 11-3 summarizes the data removed from the leak rate correlation database by.
f EPRI [3] with the exclusion of [
')u l
11.4.2 Leak Rate Versus Voltage Correlation i
t The bobbin coil voltage amplitude and leak rate data of Table 10-3 were used to determine a correlation between leak rate and bobbin voltage.
The data' considered are shown in Figure 11-1 along l
with the results of the correlation analyses.
I From the methodology discussed in Section 7 of Reference 3, the correlation between leak rate and bobbin voltage is achieved by considering the log of the voltage as the independent variable and the log of the leak rate as the dependent variable, The results of our evaluation are shown below in e
Table 11-4 for the correlation with 41 data points.
i l
]
B&W-NUCLEAR TECHNOLOGIES 11-8 i
i l.
..I
...u..
I TABLE 11-4 l
(c)
] DP The results from the index of determination suggests treating the voltage on a logarithmic scale and the leak rate on a logarithmic scale.
Therefore, from this analysis, the correlation line for the leak rate-voltage correlation for [d] data points is given by:
[
]"
Therefore:
[
(d)
) BP Eq. 11-3 The error of the estimate from the evaluation, S,,
was [
]"
A []" prediction band for individual values of leak rate, Q, as a function of voltage 3
was also calculated per for following equation:
[
Eq. 11-4
[3]
B&W NUCLEAR TECHNOLOGIES 11-9
1 2-3p A plot of the expected leak rate is provided in Figure 11-1.
9 11.4.3 Analysis of Residuals Per the Generic Letter Section 2.b.3(2), the leak rate model should account for data scatter of the linear regression fit.*
Figure 11-2 shows the scatter. plot of the log (Q) residuals as a function of the prediction log (Q) for the MSLB pressure of 2650 psi.
The arrangement of the points is non-descript, indicating no apparent correlation between the residuals and the predicted values.
The cumulative probability plot prepared for MSLB differential pressure of 2650 psi is shown in Figure 11-3.
A straight line is approximated, typical of the behavior of normally distributed residuals.
1 Given the results of the residuals scatter plot and the normal probability plot, it is appropriate to use the regression curve and statistics can be used for the prediction of leak rate as a function of bobbin amplitude, and for the establishment of statistical inference bounds.
B&W NUCLEAR TECHNOLOGIES 11-10
11.4.4 Conditional Leak Rate Under Accident Conditions The experimentally obtained= leak rate' data for 3/4 inch tubes indicates a bi-logarithmic relationship between measured leak rate and measured bobbin coil voltage.
These data have been used to-develop the conditional leak rate correlation.
The correlation is used to predict the leak rate for an individual tube (at a TSP' location with confirmed ODSCC) from the measured bobbin coil voltage.
The relationship between individual letk rate and measure voltage can be represented by:
[
Eq. 11-5 (6) 3" Taking the logarithms of both sides of Eq. 11-5, yields the equivalent linear form:
I
[
l Eq. 11-6 (6) ja Correlations presented in Reference 6 are not used j
for the STP ARC analyses due to restrictions stated in the Generic Letter, and the additional B&W NUCLEAR TECHNOLOGIES 11-11 i
leak. data used in the' development of the leak rate correlation for STP.
Therefore, the correlation used in the STP leak rate analysis is given by:
[
(e)
) pr The predicted calculated conditional leak rate is based on the randomization of leak rate correlation slope, intercept and residual error values, as described in Reference 6.
The predicted conditional leak rate at EOC is calculated from Eq. 11-7.
The calculated conditional leak rate result from Equation 11-7 is then multiplied by the probability of leakage result from Equation 11-2 for each predicted EOC voltage to obtain the predicted leak rate.
The total process is repeated until all predicted leak rate results are calculated for the predicted EOC voltages.
11.5 Calculations of Offsite and Control Room Doses For the MSLB leak rate calculated above, offsite and control room doses should be calculated utilizing currently accepted licensing basis assumptions.
Licensees should note that of the Generic Letter provides example Technical Specification (TS) pages for reducing reactor coolant system B&W NUCLEAR TECHNOLOGIES 11-12
I I
e i
specific iodine activity limits.
Reactor Coolant system lodine activities maylbe reduced to.35 microcurie:perLgram dose equivalent'I-131.
Licensees; wishing ~to reduce iodine
-activities'below this level-should provide'a. justification supporting the request that addresses the release rate' data. -
-Reduction'of reactor' coolant iodine activity is'an acceptable means for accepting higher _ projected leakage
= rates and still meeting the applicable limits of 10 CFR'100 utilizing licensing basis assumptions.
STP Compliance with Requirements 11.5.1 Leakage Evaluation Per 10 CFR 100 The maximum allowable end of cycle primary-to-secondary steam generator leak-rate will be-
, evaluated to determine whether the radiological consequences will remain within the-limits of'10 CFR 100 and GDC 19 design criteria for STP during a postulated main steam line break.
The evaluation will be based on an acceptance criteria of 30 rem thyroid dose at the Exclusion Area Boundary per the Standard Review Plant (NUREG 0800) Section 15.1.5, Appendix A.
For the pre-accident spike, the initial primary coolant activity was 60 yCi/gm dose equivalent Iodine 131 (I 131), and for the concurrent spike the activity was 1 yCi/gm.
The secondary. coolant activity was 0.1 yCi/gm I 131.
The leak rate in the three intact steam generators was assumed to be the proposed Technical Specification limit of 150 gallons per day (about 0.1 gpm) in each generator.
The leak rate from the reactor coolant system was assumed to be 1 gpm also per the technical specification.
The activity released to the environment due to a main steam line break was analyzed in two distinct releases:
1.
the release of the iodine activity B&W NUCLEAR TECHNOLOGIES 11-13 l
s that has been established in.the secondary coolant prior to the accident, and 2.
the release of the
. primary coolant iodine activity due.to tube l
leakage.
The bounding leak. rate will be.
determined'and maintained as part of the design
~
basis.
The STP-1 EOC 5 predicted leakage is.( (d),]ar for; the faulted steam generator (SG 'C')'.which isi much less than the current design basis leakage of 1 gpm.
Therefore, it is reasonable that the 150 gpd--
proposed technical specification limit per steam generator is sufficient to preclude unexpected-
~
crack propagation leakages which would result in a MSLB dose release in excess of 10 CFR 100 or GDC 19.
i 1
4
+
0
)
t a
d B&W NUCLEAR TECHNOLOGIES 11-14
TABLE 11-2 [3]
l I
This Table from EPRI NP 7480-L, Volume 2.
B&W NUCLEAR TECHNOLOGIES 11-15
TABLE 11-3
[
l 1
)lr This Table from EPRI NP 7480-L, Volume 2.
B&W NUCLEAR TECHNOLOGIES 11-16
4 FIGURE 11-1 [3]
2650 PSI MSLB LEAK RATE VS. BOBBIN AMPLITUDE 3/4" TUBES, MODEL BOILER AND FIELD DATA
[
(c) j i
i y ur i
Data in this Figure derived from EPRI NP 7480-L, Volume 2.
B&W NUCLEAR TECHNOLOGIES j
11-17 l
FIGURE 11-2 [3]
RESIDUALS VS. PREDICTED LOG OF LEAK RATES 3/4" TUBES, MODEL BOILER AND FIELD DATA g.
(c)
Data in this Figure derived from EPRI NP 7480-L, Volume 2.
B&W NUCLEAR TECHNOLOGIES 11-18
-FIGURE 11-3 [3]
CUMULATIVE PROBABILITY OF RESIDUAL LEAK RATES
[
i (c)
)w
-Data in this Figure derived from EPRI NP 7480-L, Volume 2.
B&W NUCLEAR TECHNOLOGIES 11-19 l...
d 12.O~ OPERATIONAL LEAKAGE LIMITS 12.1 Introduction The operational leak limit is a defense-in-depth measure that provides a means for identifying. leaks during operation to enable repair before such leaks result in tube failure.
Review of leakage monitoring measures includes the procedures for timely detection, trending, and response to rapidly increasing leaks.
The objective is-to ensure that should a significant leak be experienced in service, it will be detected and the plant shutdown in a timely manner to reduce the likelihood of potential tube rupture.
12.2 gggrational Leakaae Limits Per the Generic Letter, the operational. leakage' limit should be reduced to 150 gallons per day (gpd) through each steam generator.
Licensees should review their plant specific leakage monitoring measures to. ensure that if a significant in-service leak occurs, it will be detected and the plant shutdown in a timely manner to reduce.the likelihood of tube rupture.
Specifically, the effectiveness'of the plants' procedures to ensure timely detection, trending, and response to rapidly increasing leaks should be assessed.
Alarm setpoints on primary-to-secondary leakage detection instrumentation and the various criteria for specific operator actions in response to detected leakage must also be evaluated.
Steam generator tubes with known leaks should be repaired prior to returning the steam generators to service following a steam generator inspection outage.
B&W NUCLEAR TECHNOLOGIES 12-1
STP Compliance with Requirement HL&P will, in its Revised Technical Specification submittal, commit-to an operational leakage limit of 150 gpd per steam generator,.for each of the STP units.
HL&P has reviewed STP plant specific leakage monitoring procedures and actions to ensure that any leaks that develop during an operational cycle will be detected,-trended, and the plant shutdown in a timely manner.
HL&P has committed to repair all tubes with known leaks prior to returning steam generators to service following an inspection outage.
I B&W NUCLEAR TECHNOLOGIES 12-2
~
~.... - -
4 4
L13.0 REPORTING REQUIREMENTS 13.1 Introduction Per Section 6 of the Generic Letter, documentation reporting the EOC voltage distribution, cycle growth rate.
distribution, voltage distribution.for EOC-repaired indications-(indications confirmed and unconfirmed by MRPC) and NDE' uncertainty distribution in predicting the next EOC distribution shall be submitted to the NRC.
i 4
13.2 Threshold Criteria'for Reauirina Prior Staff Anoroval to Continue with'Voltaae-Based Criteria This guidance allows licensees to implement the voltage-based repair' criteria on a continuing basis after the NRC staff has approved the initial TS amendment.
However, there-are several situations for which the NRC staff must receive prior notification before a licensee can continue'with the implementation of the voltage-based repair criteria:
13.2.a If the actual measured voltage distribution would have resulted in an estimated leakage during the previous operating cycle greater than the leakage limit (determined from the licensing basis calculation), then the licensee should notify the NRC of this occurrence and provide an assessment r
of this significance prior to returning the steam generators to service.
13.2.b If (1) indications are identified that extend beyond the confines of the TSP cnr (2) indications are identified that appear to be1circumferential in nature, then the NRC staff should be notified prior to returning the steam generators to service.
B&W NUCLEAR TECHNOLOGIES 13-1
F 1
2 0-a Lp b
13.2.c.
If the-calculated conditional probability of burst.
~
EC
~ based on the projected EOC voltage l distribution; 1
4 exceeds 1 x;10,Llicensees.should notify NRC and provide an assessment of the significance of-this occurrence prior.to returning the. steam generators-to service.
This assessment ~should address the f
safety significance of the calculated; conditional, probability.
k STP chapliance with Requirements L
13.2.1 STP Threshold Criteria for Requiring Prior Staff Approval to continue with Voltage-Based Criteria Per Section 6.a of the Generic Letter STP will notify the NRC if the actual measured voltage-distribution results in an estimated leakage during the previous operating cycle greater than the leakage limit determined from Section 11 and
-12 of this report.
STP will notify the NRC if indications are identified that extend beyond the TSP or indications that: appear to be p
circumferential in nature prior to returning the l
SGs to service.
STP will notify the NRC if the
~
calculated conditional probability of burst based on the EOC voltage distribution.is greater.than 1
]
l' x 10 prior to returning the SGs to service.
4 13.3 Information To Be Provided Followina'Ilach Restart The following information should be submitted to the NRC staff within 90 days of each restart following a steam generator inspection:
(a)
The results of metallurgical examinations performed for tube intersections removed from the steau generator.
B&W NUCLEAR TECHNOLOGIES 13-2
,q in i
h
?
1" l
1
..The-following distributions should;be previded.in'both i
(b)
-+
This information is to tabular and.grsphical form..
f enable the staff to assess'the effectiveness of the' methodology,' determine whether,the degradation is l
changing significantly,'dctermine whether the data.
supports higher voltage repair limits, and to perform confirmatory calculations:
-(i)
As.found EOC voltage' distribution - all:
indications found.during the inspection-regardless of MRPC confirmation
-}
(ii) cycle voltage. growth rate distribution (i.e.,
(iii) volt. age distribution for EOC repaired I
indications - distribution of indications j
presented in (i) above.that were repaired 5
[
(i.e., plugged or sleeved)
(iv) voltage distribution for indications left in service at the'beginning of the next i
operating cycle regardless of MRPC confirmation -'obtained from (i) and (iii) above
]
(v) voltage distribution for-indications left in service at the beginning of the next operating cycle that were confirmed by MRPC to be crack-like or not MRPC inspected (vi) non-destructive examination uncertainty distribution used in predicting the EOC (for the next cycle of operation) voltage distribution 1
1 i
B&W NUCLEAR TECHNOLOGIES 13-3 1
21, (c)
The results of the tube integrity evaluation described in Section 2 of the Generic Letter and discussed in Section 10.0 of th'.s report.
Note that these calculations must be completed prior to restart to ensure that an adequate number of tubes have been repaired to meet the leakage limit and ensure continued tube integrity.
STP Compilance with Requirements 13.3.1 Information To Be Provided by STP Following Restart Consistent with the Generic Letter requirements discussed in Section 13.3 of this report, the following actions shall be taken to ensure that the NRC is aware of the on-going status of STP ARC implementation.
Beginning with the upcoming Refueling Outage 6 data for Unit 1 and subsequent outages; (1)
All indications found during the inspection regardless of MRPC confirmation will be reported to the NRC.
(2)
The cycle growth rate distribution will be evaluated to determine whether the growth rate assumed remains bounding.
The results of the growth rate distribution will be reported to the NRC.
(3)
Voltage distribution for EOC repaired indications that were repaired will be reported to the NRC.
l (4)
Voltage distribution for indications left in j
service at the beginning of the next operating cycle regardless of MRPC confirmation will be reported to the NRC.
l l
B&W NUCLEAR TECHNOLOGIES 13-4
(S)
Voltage-distribution for indications left in l.[
service at the beginning of the next operating cycle that were confirmed by MRPC or not MRPC inspected will be reported to the NRC.
(6)
The results of the pulled tube data per Section 8.0 shall be reviewed against the burst correlation data for continued' i
applicability or adjustment.
The-metallurgical examinacion and testing results shall be reported.
The inspection data shall be reviewed along with destructive
)
examination results to verify that the morphology remains consistent with this submittal.
Any indication of changing morphology or observation of cracks extending beyond the confines of the. TSP shall be reported.
I 1
l B&W NUCLEAR TECHNOLOGIES 13-5
r o
The loads were based on.a larger attachment line break (14" diameter schedule 140 versus.the' actual STP 12" diameter schedule:140 line size).
o-A stress-strain curve based on the ASME Code minimum yield and tensile strength properties for.
~
the support plate was used.
It is highly.likely.
that the actual material has greater than Code minimum properties.
i..
o The increase in yield strength due to the rapidly applied load (high strain rate) was neglected.
o It was assumed that the tube deformation.is equal-to the hole ID deformation in the finite element
(
analysis even though a gap may exist.
Also,..the TSP stiffness neglected any contribution provided by the tubing.
1 I
o It was assumed that the irterface between the support plate and wedges is frictionless even though the wedges were snugly installed and are securely welded to wrapper support blocks..
o It was assumed that the entire LOCA rarefaction load is reacted out at the top support plate only.
I The resulting load on the TSP due to the rarefaction wave was [
(d)
)s' for STP.
The associated shaking load was [
(d)
) s'.
These loads were conservatively added and a dynamic load factor of (d]ar was applied, yielding a total LOCA load of (
(d)
) 8' [15).
Seismic The seismic loads considered result from ground motion t
during an earthquake.
This motion causes an excitation i
of the steam generators in the form of acceleration
[
response at the steam generator supports.
The I
acceleration response is converted to a time history B&W NUCLEAR TECHNOLOGIES 14-4 j
_ _ =
t i.i response to determine the load contribution within the
' tubes.and support plates.
The loading evaluation was
~
previously performed and the-results are contained in j
the stress report-[15) as in-plane loading of the tube support plates.
The SSE load was determined to be [(d) l
)".
The total TSP load was determined to'be [
L (d).
] [15).
14.2.2 Analysis Methodology The steam generator upper tube support plate was input into an inelastic ANSYS finite element model to evaluate the two wedge groups for one quadrant 7
(symmetry) as the bounding case for all wedge groups at the top support plate.
An overall illustration of the model is provided in Figure 14-1.
A detailed' view of
{
the elements modeled is provided in' Figures-14-2 and 14-3.
The total load of ( (d)]" was applied to the' 4
finite element model.
The loading and deformation j
results are provided in Reference 11.
Each of the j
lower support plates were conservatively assumed to exhibit this worst case loading as well.
However since the wedge groups are vertically aligned, the number of I
tubes affected is minimized.
A summary of the excluded tubes is provided in Section 14.3.
14.3 Snamarv of Excluded Tube Locations t
Figure 14-4 provides a sketch of the tube bundle showing the TSP locations and designations for identifying the elevation of the wedge group locations.
Figures 14-5 and 14-6 show the circumferential locations of the wedge groups for the tube support plates.
The corresponding nomenclature from South Texas Project is also provided for additional information.
The tube locations which deformed greater than the [
(d)
Ja' criterion as described in Section 5.3 were characterized through the analysis (11] as unacceptable.
B&W NUCLEAR TECHNOLOGIES 14-5
The lower TSP wedge groups at the 37 degree reference location were conservatively assumed to exhibit the same exclusion region size as those at the 32 degree reference location.
Figures 14-7 through 14-10 summarize the exclusion tubes.
Table 14-1 provides a summary of all excluded SG tubes due to deformation in the vicinity of the wedge locations.
The total number of tubes excluded per steam generator'is-
[ (d) )"'.
The most limiting wedge group location (32 degree reference) has [d]n' tubes excluded due to tube deformation.
This large number of tubes excluded is mainly the result of the previously described conservatisms used in developing and evaluating the analysis results.
A submittal for another plant shows the limiting wedge group exclusion region to be approximately 30 tubes for a similar steam generator with slightly higher loading conditions from a large primary pipe break (16].
The exclusion region for STP is nearly [ (d)
]"' as large, for the limiting wedge group.
Thus, the number of tubes excluded at STP is representative.
1 B&W NUCLEAR TECHNOLOGIES 14-6
FIGURE 14-1 TSP MODEL
[
(d) j ne I
B&W NUCLEAR TECHNOLOGIES 14-7
..j- -
5
'#IGURE-14-2 MODEL ELEMENTS 32 DEGREE WEDGE LOCATION l
i c
f f
)
f (d) a i
i 4
UP B&W NUCLEAR TECHNOLOGIES 14-8 i
Y l
i FIGURE 14-3
.t i
NODEL ELEMENTS 16 DEGREE. WEDGE LOCATION l'-
i 9
F 1
4 9
(d) i 6
4 h
k
{
y ar i
i B&W NUCLEAR TECHNOLOGIES 14-9 r
+ -
+
3
~i i
i FIGURE 14-4 1
-TUBE BUNDLE SHOWING TSP ELEVATIONS FOR IDENTIFYING WEDGE GROUPS
[
I (d) 1 i
) ttP i
I B&W NUCLEAR TECHNOLOGIES 14-10
- l;.
~
s i
FIGURE 14-5
-WEDGE GROUP LOCATIONS TSPs 1-11 f
~
-I.
I
[
t r
i b
i i
t (d) i l
3 ar B&W NUCLEAR TECHNOLOGIES 14-11
~.
t
,;s
-t
}
FIGURE 14-6
- WEDGE GROUP LOCATIONS:-T8P.12 I
i
~ $
I
~
e i
>I i
s I
r i
t (d)
)
1 e
+
i 6
9 f
-f I
UE B&W NUCLEAR TECHNOLOGIES 14-12 i
1 I
m i
---r-
--+r e
I. l.
l FIGURE 14-7 EXCLUDED TUBE REGION: TSPs 1-11 (NOEELE SIDE)
I (d) 3 ar Note:
This drawing is a depiction of the tubes contained in this exclusion area.
Only an actual tubesheet map should be used to show the exact tube exclusion locations.
B&W NUCLEAR TECHNOLOGIES 14-13
Y t
FIGURE 14-8 1
EXCLUDED TUBE-REGION: TSPs 1-11 (MANWAY SIDE)
[
~
(d) t BJr Note:
This drawing is a depiction of the tubes contained in this exclusion area.
Only an actual.tubesheet map should be used to show the exact tube exclusion locations.
B&W NUCLEAR TECHNOLOGIES 14-14
FIGURE 14-9 EXCLUDED TUBE REGION: TSP 12 (NO22LE~ SIDE)
I 3
l 1
A J
Y
'i h
(d)
+
t l
b
)se Note:This drawing is a depiction of the tubes contained in this exclusion area.
Only an actual tubesheet map should be used to show the exact tube exclusion locations.-
1 B&W NUCLEAR TECHNOLOGIES 14-15
FIGURE 14-10 EXCLUDED TUBE REGION: TSP 12
.(MANWAY SIDE) 4 a
(d)
]er Note:
This drawing is a depiction of the tubes contained in this exclusion area.
Only an actual tubesheet map should be used to show the exact tube exclusion locations.
B&W NUCLEAR TECHNOLOGIES 14-16 1..
.. ~.
.m.
r i
14.
I; i
TABLE 14-1 F
SU3OULRY OF TUBES TO BE EXCLUDED - FROM ARC 4
(;
1 4
5 5
i 4'
?
L e
b f
e t
(d) 4 s
i t
9 e
f 5
jer d
4 9
l b
i i
B&W NUCLEAR TECHNOLOGIES 14-17
M
~
15.0 CONCLUSION
S
)
This. assessment demonstrates that a correlation relating tube burst pressure to bobbin. voltage. and mainL ateam line break (MSLB) leakage to bobbin voltage can be used to conservatively :
satisfy.the Reg Guide 1.121_ guidelines for tube integrity at i
South Texas Project Units 1'& 2.
Application of the generic ODSCC ARC methodology' developed-through EPRI is appropriate for South Texas Units l'& 2 ODSCC flaws,' satisfies the NRC Generic Letter on ODSCC ARC, and is consistent with other. approved ODSCC ARC methodologies.
l The Generic Letter consists of six sections of. requirements for a licensee to include in a-proposed program to implement l
ARC.
STP complies with these as follows:
(1)
Confirmation that the degradation mechanism is predominately axial ODSCC confined to the TSP.
s STP Compliance with Requirement During 1RE04, STP pulled 4 t u b'e s to verify ODSCC is th'e
~*
dominate degradation mechanism at the TSP. The indications at the TSP were burst tested to demonstrate that the. dominant, mechanism affecting the burst and leakage. properties of the tube is axially oriented ODSCC.
(2)
Confirmation that the steam generator tubes will retain j
adequate structural and leakage integrity until the next scheduled inspection.
j STP Compliance with Requirement
{
Section 10 of this report discusses the methodology used to calculate the probability of burst for ' STP EOC-5.
The probability of. burst for SG
'B' is [ (d) Ja' for EOC-5, which is well within the threshold value of 1 x 102 provided by the NRC in the Generic Letter.
Section 11 of this report discusses the methodology used to calculate the predicted B&W NUCLEAR TECHNOLOGIES 15-1 i
1 4
conditional MSLB' leak rate of the. steam generators-at STP.
For STP-1, EOC-5 leak rates were calculated'for each steam generator.
Steam generator:'C'_had the largest leak rate of
[d)"' gpd which is well within the current design basis of[d]*"
gpd..
(3)
Inspection scope, data acquisition, and data analysis should-. be performed in a manner consistent. with the methodology utilized to develop the voltage limits.
i STP Compliance with Requirement Section 7 and' Appendix A of this report address the NDE inspection criteria and ECT analysis requirements to be followed during 1RE05 and future outages at STP in accordance with the requirements provided by the Generic Letter.
(4)
Implementation of voltage-based plugging criteria should include a program of tube removals for testing and examination as described in Section 4 of the Generic Letter.
STP Compilance with Reguirement i
Section 8 discusses the requirements for tube removal and examination at STP during 1RE05 and future outages..
In compliance with the Generic Letter and to diagnose other possible corrosion problem areas, STP will pull 3 tubes during'-
1RE05 for Unit 1.
These pulled tubes will be burst'and leak tested to verify that axial ODSCC is the dominate degradation mechanism at the TSPs in addition to the 12 intersections pulled from 1993.
B&W hUCLEAR TECHNOLOGIES 15-2 j
i
(5).The-operational-leakage limit should be. reduced to 150 gpd through each steam generator BTP Compliance with Requirement Section 12 of this report addresses operational leakage limits at STP-1.
In compliance with the Generic Letter, STP will review leakage monitoring measures 'to ensure that-a significant leak will be detected.
amendment to the Technical (6)
Reporting Requirements Specifications STP Compliance with Requirement STP will submit the amendment to the TS with the submittal of.
this report to the NRC.
1 4
B&W NUCLEAR TECHNOLOGIES 15-3
)
s
16.0 REFERENCES
- 1..
NRC Generic Letter 94-xx:
" Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Genera' tor Tubes Affected by Outside Diameter Stress Corrosion Cracking", undated.
2.
[
je 3.
[
]"
4.
[
(c) 3" 5.
NRC Regulatory Guide 1.121 " Bases for Plugging Degraded PWR Steam Generator Tubes", August 1976.
i 6.
[
ja 7.
(
(c) 3" 8.
USNRC to HL&P Letter, "NRC Bulletin 88-11, " Pressurizer Surge Line Thermal Stratification - South Texas Project, Units 1 and 2 (TAC No. 72168)", September 17, 1990.
9.
"ASME Boiler and Pressure Vessel Code",Section III, Subsection NB and Division I Appendices, 1989 Edition.
l j
B&W NUCLEAR TECHNOLOGIES 16-1
(.
...3
.j V
10.
[
(c)'
)"
l 1
11.
[
(c) 3" x
l 12.
NRC letter from Allen R. Johnson to Robert C. Macredy
" Application of Leak-Before-Break Technology, R.E..
Ginna Nuclear Power Plant (TAC No. M86376)", Docket No.
50-244, 1993.
13.
[
(c)
)w 14.
(
(c)
)w 15.
HL&P Document No. 120 (1) 00019-CWN, "Model E2 Steam Generator Stress Report" and addendum.
Reference (15) 1:s not available for. entry into the-BWNT Records Center, but may be referenced for use on Task 1101 of BWNT Contract 1010277.
Use of this reference is permitted by BWNT Procedure, BWNT-0402-01, Appendix 2.
16.
NRC Letter from George F. Dick, Jr. to D.L. Farrar
" Issuance of Amendments.(TAC Nos. M90052 and M90053),
dated October 24, 1994; Amendment No. 66, Docket STN 50-454 p.15.
17.
Meeting Minutes with Industry 1/18/95, Resolution of Public Comments NRC Draft GL 94-XX.
18.
NUREG-0781, " Safety Evaluation Report related to the 5
operation of South Texas Project, Units 1 &
2",
Supplement No.
2, January 1987.
i B&W NUCLEAR TECHNOLOGIES 16-2
l t
19 NUREG-0781, " Safety Evaluation Report related to the operation of South Texas Project, Units 1 & 2",
Supplement No.
4, July.1987.
20.
HLEP to USNRC Letter ST-HL-AE-3016, " Pressurizer Surge Line Thermal Stratification", March 14,'1989.
21.
NUREG-1477, " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes", June, 1993..
22.
NUREG-0800, " Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR, Rev.
2, July, 1981.
23.
WCAP-13523, "V.C. Summer Steam Generator Tube Plugging Criteria for Indications at Tube Support Plate, Westinghouse Non-Proprietary, January, 1993.
24.
BMDP Solo, Version 4, BMDP Statintical Software, Inc.,
12121 Wilshire Blvd, Los Angeles, CA 90025.
25.
HL&P Letter, PFN M18.05.02, ST-HS-2U-0009, Mechanical Properties of Unit 1 and Unit 2 Steam Generator Tubing, dated February 18, 1994.
l l
B&W NUCLEAR TECHNOLOGIES 16-3 j
1 4
APPENDIX 3p MDB DATA ACQUISITION AND ANALYSIS REQUIREMENTS POR OD8CC'AT TSP ARC 1
t i
A.1 - INTRODUCTION-Th'is appendix documents required techniques-for'the inspection of South Texas Project Units 1&2 steam generator A
4 tubes related to the identification of ODSCC at.the tube support plate regions.
This appendix contains requirements which provide direction in applying the ODSCC alternate repair criteria (ARC) described in this report.
The procedures for eddy current testing using bobbin coil- (BC) and rotating pancake coil
(
(RPC) techniques are summarized.
The procedures given apply to the bobbin coil inspection, except as explicitly noted
[
for MRPC inspection.
The following sections define specific acquisition and analysis parameters and methods to be used for the inspection of steam generator tubing.
+.
A.2 DATA ACQUISITION The following guidelines are specified for non-destructive examination of the tubes within TSP at South Texas Project Unit 1 and Unit 2.
A.2.1 Instrumentation Eddy current equipment shall be the Zetec MIZ-18 t
or engineering approved equivalent.
+
B&W NUCLEAR TECHNOLOGIES A-1
- w g-f
, f, t
a
~
l 4
A.2.2 Probes A.2.2.1 Bobbin Coil' Probes
'To maximize consistency withLlaboratory ARC data, differential' probes with'the following parameters shall' be used for examination of tube support' plate-intersections:
0.610 outer diameter' two bobbin coils, each 60 mils long x 60 mils deep,.with 60 mils between coils.(coil centers separated by 120 mils) l In addition, the probe design must incorporate centering features that provide.for minimum probe wobble.and offset; 'the centering features must
-maintain constant' probe center to tube ID offset for nominal diameter tubing.
For locations-which must be inspected with smaller than nominal diameter probes, it is essential that the reduced diameter probe be calibrated to the.
reference normalization (Section A.2.6.1 and A.2.6.2) and that the centering feature permit constant probe center to tube ID offset.
A.2.2.2 Rotating Pancake Coil Probes Pancake coils designs (vertical dipole moment) with_a coil diameter d, where d
)
is 0.060" $ d 5 0.125", shall be used.
While other multi-coil (i.e.,
1, 2, or 3-coil) probes can be utilized, it is recommended that if a 3-coil single B&W NUCLEAR TECHNOLOGIES A-2
~
, pancake probe is used, any voltage measurements should be made with the probe's pancake coil rather than its
-circumferential or axial coil.
A.2.3 Calibration Standards A.2.3.1 Bobbin Coil Standards The bobbin coil calibration standards' contain the following items:
Voltaae Normalization Standard:
One'O.052" diameter 100% through wall hole Four 0.028" diameter through wall holes, 90 degrees apart in a single 7
plane around the tube circumference; the hole diameter tolerance shall be +/- 0.001" One 0.109" diameter flat bottom hole, 60% through from the OD Four 0.187" diameter flat bottom holes, 20% thrcugh from the OD, spaced 90 degrees apart in a single plane around the tube circumference.
The tolerance on hole diameter and depth shall be
+/- 0.001".
A simulated support ring, 0.75" long, comprised of SA-285 Grade C carbon steel or equivalent for Unit 1 and comprised of SA-240 type 405 stainless steel or equivalent for Unit 2.
If mix residuals are shown B&W NUCLEAR TECHNOLOGIES A-3
~
to be equivalent, then carbon steel can be used for both units.
All holes shall be machined using a mechanical drilling technique.
This calibration standard will need to be
~
calibrated against the reference i
standard used for the ARC laboratory work by direct testing or through the use of a transfer standard.
Probe Wear Standard A probe wear standard is used for monitoring the degradation of. probe centering devices leading to off-center coil positioning and potential variations in flaw amplitude responses.
This standard shall include four 0.052" +/-
0.001" diameter through-wall holes, spaced 90 degrees apart around the tube circumference with an axial spacing such that signals can be clearly distinguished from one another.
See Figure A-1.
A.2.3.2 Rotating Probe Standard A satisfactory MRPC standard may j
I contain*
)
Two axial EDM notches, located at the same axial position but 180 degrees apart circumferential, each 0.006" wide and 0.5" long, one 80%
and one 100% through wall from the OD.
B&W NUCLEAR TECHNOLOGIES A-4
4 i
Two axial EDM notches, located'at i
the same axial position but-180~
3 degrees' apart circumferentially, each'0.006" wide and 0.5" long,-one 60% and one 40% through-wall from f
/
the OD.-
Two circumferential EDM notches, one 50% through wall from the OD vith a 75 degree (0.49") arc lengthi and one 100% through walli l
with a 26 degree-(0.17") erc length, with both notches.0.006"'
wide.
t A simulated support segment.270 degrees in circumferential extent ~,
O.75" thick, comprised of SA-285 Grade C carbon steel or~ equivalent for Unit 1 and comprised of SA-240 type 405 stainless steel or.
equivalent for Unit 2.
Similar configurations which satisfy.the-intent of calibrating MRPC probes for'OD axial and circumferential' cracking are i
satisfactory.
The center to center distance between the support plate simulation and the nearest slot shall be-at least 1.25".
The center to center distance between the EDM notches shall be at least 1.0".
The tolerance for widths and depths of the notches shall j
be 0.001".
The tolerance for the. slot lengths shall be 0.010".
B&W NUCLEAR TECHNOLOGIES A-5
.-,,. a --
.a
r A.2.4 Application of Bobbin Coil Wear Standard A calibration standard has been designed to monitor bobbin coil probe wear.
During steam generator examination, the bobbin probe is inserted into the wear monitoring standard; the initial (new probe) amplitude response from each of the four holes is determined and compared on an individual basis with subsequent measurements.
Signal amp]itudes from the individual holes -
compared with their initial amplitudes - must remain within 15% of their initial amplitude (i.e., {(worn-new)/new)) for an acceptable probe wear condition.
If this condition is not satisfied, then the probe must be replaced.
Two enticis exist for addressing the tubes inspected bl 6Ae worn probe since the last calibration.
1)
If any of the last probe wear standard signal amplitudes prior to probe replacement exceeds the 15% limit, say by a variable value, xt, then indications measured since the last acceptable probe wear measurement that are within x% of the rcxpair limit must be re-inspected with the new probe.
2)
Obtain all preliminary voltages with no probe wear standard.
Then reinspect the preliminary indications that are above 75% of the repair limit voltage with a new probe that is subject to probe wear guidelines.
Wear measurements should be taken before and after each inspection.
A.2.4.1 Bobbin Coil Wear Standard Placement Under ideal circumstances, the incorporation of a wear standard in line with the conduit and guide tube configuration would provide continuous monitoring of the behavior of bcbbin B&W NUCLEAR TECHNOLOGIES A-6
e s
e
. probe wear.
However, the curvature of
[
the channelhead places restrictions on the length on in line tubing inserts which'can be accommodated.
The spacing.
of the ASME Section XI holes and tho wear standard results in a'langth of tubing which cannot be-freely positioned within the restricted space available.
The flexible conduit sections inside the channelhead, together with the guide tube, limit the space available for i
additional in line components.- Voltage responses for the wear standards are sensitive to bending of the leads, and-mock up tests have shown sensitivity to the robot end effector position in the tubesheet, even when the wear standard is placed on the bottom of the channelhead.
Wear standard measurements i
must permit some optimization of positions for the measurement and this should be a periodic measurement for inspection efficiency.
The pre-existing requirement to check calibration using the ASME tubing standard is satisfied by I
periodic probing at the beginning and end'of each probe's use as well as at four hour intervals.
This frequency is adequate for wear standard' purposes as well.
Evaluating the probe wear under uncontrollable circumstances would' present variability.in response due to channelhead orientations rather than changes in the probe itself.
i B&W NUCLEAR TECHNOLOGIES A-7
f' n
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A.2.5 Acquisition Parameters 1
The following parameters' apply to bobbin coil. data:
acquisition and should be' incorporated-in the applicable inspection procedures.to supplement, i
(not necessarily replace) the parameters normally.
l used..
6 A.2.5.1-Test Frequencies This technique requires'the use of bobbin coil 550 kHz and 130 kHz test frequencies in the differential mode.
It is recommended that the absolute mode also be used, at' test frequencies of 130 kHz and 10 kHz.
The' low frequency.-(10 kHz) channel should be recorded to
)
i provide a means of verifying tube support plate edge detection for flaw-location purposes.
The 550/130. kHz mix or the 550 kHz differential channel is used to access changes in signal amplitude.for the probe wear standard as well as for flaw detection.
.1 i
MRPC frequencies should include channels adequate for detection of OD degradation' in the range of 100 kHz to 550 kHz, as well as a low frequency channel'to provide support location of the TSP edges.
l 4
A.2.5.2 Digitizing Rate A minimum digitizing rate of 30' samples:
7 per inch-should be used for both bobbin j
and MRPC.
Combinations of probe speeds and instrument sample rates should be j
chosen such that:
i B&W NUCLEAR TECHNOLOGIES A-8
Samole Rate (samoles/sec.).;t 3 0 (samples /in.)
Probe Speed (in./sec.)
j A.2.6 Analysis-Parameters This section discusses the methodology for establishing bobbin ~ coil data analysis variables such as spans, rotations, mixes, voltage scales, and calibration curves.
Although indicated depth measurement may not-be required to support an alternative repair limit,.the-methodology for establishing the calibration curves is presented.
The use of these curves is recommended for consistency in reporting and to provide compatibility.
of results with subsequent inspections of-the same steam generator and for comparison with other steam generators and/or plants.
A.2.6.1 Bobbin Coil 550 kHz Differential Channel Rotations: The signal from the 100%'through-wall hole should be set to 40' (+/- 1 degree) with the initial signal excursion down and to the right during probe withdrawal.
Voltaae scale: The peak-to-peak signal amplitude of the signal from the four 20%
through-wall holes should be set to produce a voltage equivalent to that obtained from the ARC lab standard.
The laturatory standard normalization voltage is 4.0 volts at 550 kHz.
The transfer / field standard will be calibrated against the laboratory standard using a reference laboratory probe to establish voltages for the field standard that are equivalent to the above laboratory standard.
These equivalent voltages are then set on the field standard to establish calibration voltages for any other standard.
B&W NUCLEAR TECHNOLOGIES A-9
Voltage normalization to the' standard-calibration voltages at 550 kHz is the preferred normalization to minimize analyst sensitivity in establishing the mix.
However, if the bobbin probts used result in a 550/130 kHz mix'to 550 kHz voltage ratios differing from the laboratory standard ratio -
of 0.69 by more than 5% (0.66 to 0.72), the 550/130 kHz mix calibration voltage should be used for voltage normalization.
-Once the probe has been calibrated on the 20%
through-wall holes, the voltage response of new bobbin coil probes for the 20-80% ASME through-wall holes should not differ from the nominal voltage by more than i 10%.
As an alternative, probes can be supplied with certification of meeting the variability requirements upon shipment from the vendor.
Calibration curve: Establish a phase versus depth calibration curve using measured signal phase angles in combination with the "as-built" flaw depths for the 100%, 60%, and 20%
holes.
A.2.6.2 Bobbin Coil 550/130 kHz Differential Mix Channel Rotations: Probe motion is set horizontal with the initial excursion of the signal from the single 100% through-wall hole going down and to the right during probe withdrawal.
Voltace Scalg: The peak-to-peak signal-amplitude of the signal from the four 20%
I through-wall holes should be set to produce a voltage equivalent to that obtained from the ARC lab standard.
The laboratory standard normalization voltage is 2.75 volts for the B&W NUCLEAR TECHNOLOGIES A-10 4
l
'l i
~-
j 550/130 kHz mix.
Calibration Curve:. Mix l'is a 550/130 kHz differential support mix; mix on ASME-fatandard-support ring.
Set 3-point phase angle-depth calibration curve using.ASME j
100%, 60%, and 20% drill hole signals.. Mix 1 j
is the primary channel for reportingL
~
indications at support structures.-
A.2.6.3 Rotating Pancake Coil Inspection Rotations: Probe motion is set horizontal.
(+/- 5 degrees) with the initial excursion of the signal from the 100% through-wall ~ notch directed upwards during probe withdrawal.
Voltaae Scale: The MRPC amplitude will be-referenced to 20 volts for a 0.5" long 100%
through wall notch at'300 kHz.
Each cha'nnel shall be set individually to the desired amplitude for the EDM notches on the plant standards.
A.2.7 Analysis Methodology Bobbin coil indications at support plates attributable to ODSCC are. quantified using the Mix 1 (550 kHz/130 kHz) data channel.
This is illustrated with the example shown in Figure A-2.
The 550/130 kHz mix channel-or other channels appropriate for flaw detection'(550 kHz, 300 kHz,.or 130 kHz)-may be used to locate the indications of interest within the support-plate signal.
The largest amplitude portion of the' Lissajous signal representing the flaw should then be measured using the 550/130 kHz Mix 1 channel to establish the peak-to-peak voltage as shown in Figure
~
A-2.
Initial placement of the dots for identification of the flaw location may be performed as shown in Figures A-3 and A-4, but the final peak-to-peak measurements must be performed on the Mix 1 Lissajous B&W NUCLEAR TECHNOLOGIES A-11
4 b
signal to' include the full. flaw segment of the signal.
~
It may be necessary to iterate.the positions of.the dots between the identifying frequency and the 550/130 kHz mix to obtain proper placement.
As can be'seen in Figure A-4, failure to do so can reduce the voltage i
measurements.of Mix.1 by as much as 65% to.70% due to the interference of the support plate. signal in the raw frequencies.
The voltage as measured from Mix l'is then entered as the analysis of record for comparison with the repair limit voltage.
j To support the uncertainty allowances maintained in the
't ARC, the difference in amplitude' measurements for each-indication will be limited to 20%.
If the voltage values called by the independent analysts deviate by more than 20% and one or both of the calls exceeds 1.0 volts, analysis by the resolution analyst will be performed.
These triplicate analyses result in assurance that the voltage reported departs from the correct call by no more than 204.
There is no industry recognized method for measuring the eddy current test signal-to-noise ratio to t
determine which data is too noisy and should be re-acquired.
However, the EPRI Steam Generator Management Program has been tasked with' identifying or developing l
such methods.
Until such methods are identified, electrical noise in excess of 0.3 volts peak-to-peak on channel I will be rejected and the data will be re-l acquired.
l i
B&W NUCLEAR TECHNOLOGIES A-12
a
{.
A.2.8 Reporting Guidelines LThe reporting requirements identified below are in addition to any other. reporting requirements specified by the user.-
A.2.8.1 Minimum Requirements All bobbin coil flaw indications-in the-550/130 kHz mix. channel at the tube support I
plate intersections'regardless of'the peak--
to-peak signal amplitude must be. reported.
All TSP locations with indications exceeding-1.0 volt must be examined with MRPC probes.-
A.2.8.2' Additional Requirements For each reported indication,=the following information should also be recorded:-
Tube identification (row, column)
Signal-amplitude (volts)
Signal phase angle (degrees)
Test channel (ch#)
Arial position of tube (location)-
Extent'of test (extent)
}
MRPC reporting requirements should include as a minimum: type of degradation (axial, circumferential, or other), maximum voltage and location of the' center of the crack within the TSP.. The crack axial center to edge need not coincide with the position of-the maximum amplitude.
Locations which do not exhibit flaw-like indications in--the MRPC isometric plots may continue in service, except that all intersections exhibiting i
flaw-like bobbin behavior and bobbin amplitudes in excess of the repair limit voltage must be repaired, notwithstanding the MRPC analyses.
MRPC isometrics should be interpreted by the analyst to characterize-l B&W NUCLEAR TECHNOLOGIES A-13 I
l
~
5.
)
the signals l observed; only featureless isometrics are to be reported as.NDD.
Signals not interpreted as flaws include _
' dents, liftoff, deposits, copper, magnetite, j
etc.
MRPC indications with circumferential cracks or cracks extending outside the TSP 4
must be repaired.
i'
~
A.3 DATA EVALUATION.
A.3.1 Use of 550/130 Differential: Mix for Extracting the Bobbin, Flaw Signal
?
In order to identify a-discontinuity in.the composite j
signal as an indication of a flaw in the tube wall, a simple signal processing procedure of mixing-the data from the two test frequencies is_used which reduces the j
interference from the support plate signal by approximately one order of magnitude.. The test' frequencies most often used for this signal processing l
are 550 kHz and 130 kHz for 43. mil wall Alloy 600 tubing.
Any of the differential data channels j
including the mix channel may be used for flaw detection (though the 130 kHz for 43 mil' wall Alloy 600 l
tubing is often subject to the influence from many i
different effects), but the final evaluation of signal j
detection, amplitude and phase angle will be made.from 3
the 550/130 kHz differential mix channel.
Upon detection of a flaw signal in the differential mix channel, confirmation from other raw channels is not i
required; all such signals must be reported as indications of possible ODSCC.
The voltage scale for the 550/130 kHz differential channel should be normalized as described in Section A.2.6.1 and A.2.6.2..
i The present evaluation procedure requires that there is no minimum voltage-for flaw detection purposes and that j
all flaw signals, however small, be identified.
The intersections with flaw signals 2 1.0 volt will be inspected with MRPC, unless the tube is to be plugged l
B&W NUCLEAR TECHNOLOGIES A-14 j
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'.f or sleeved.
Although the: signal voltage is not a measure of flaw depth, it'is an indicator of the-tube 1
burst pressure when-the flaw is identified as axial.
ODSCC withLor without minor IGA. RIf-an indication is not' confirmed by MRPC,.no action is required and-the tube may remain in service.
+
A.3.2 Amplitude Variability i
It has been observed that voltage-measurements'taken from the same data by different analysts may. vary, even.
when using identical analysis guidelines.
This is
~
largely due to differences in the analyst interpretation of where to place the dots on the Lissajous figure for the peak-to-peak amplitude measurement.
Figures A-5 and A-6 show the correct placement of the dots on the Mix 1 Lissajous figures for the peak-to-peak voltage amplitude' measurements for.
a two tubes from Plant S.
In Figure A-5,-the placement is quite obvious.
In Figure A-6, the. placement requires slightly more of a judgement call.
Figure A-7 and A-8 show these same two tubes with peak-to-peak i
measurements being made, but in both cases the dots
~
have been placed at locations where the normal max-rate-
~
dots would be located.
The reduction in the voltage' amplitude measurement is 19.3% in Figure A-7 and-16.3%
in Figure A-8.. While this is an accepted method-of analysis for phase-angle measurements, it is not-appropriate for the voltage amplitude measurements required.
In Figures A-5 and A-6, the locations of the dots for the peak-to-peak measurements being performed from Mix 1 show the corresponding dots on the 550 kHz raw frequencies as also being located at the peak or maximum point of the flaw portion of the Lissajous figure.
In no case should the dots to measure the 1
voltage amplitude be at locations less than the maximum points of the flaw portion of the 550 kHz raw frequency.
B&W NUCLEAR TECHNOLOGIES A-15 1
.. ~.
m Figure A-9'is'an example of where the dots have-been placed on the transition region of the 550 kHz raw j
frequency data Lissajous figure that this.does not correspond to.the maximum voltage-measurement.
The correct placement on the Mix 1 Lissajous figure is shown in Figure A-10..
This placement'also corresponds i
to the maximum, voltage measurement on the 550 kHz raw frequency data channel.
D In some cases, it'will be found that little'if any.
definitive. help.is available from the use of.the raw frequencies.
Such an example is shown in Figure A-ll, where there are no significantly sharp transitions in any of the. raw. frequencies.
Consequently, the placement of the measurement dots must'be mado completely on the basis of the Mix 1 channel Lissajous figure as shown in the. upper.left of the graphic.
An-even more difficult example is shown in Figure A-12.
The' logic behind the placement of the dots in the Mix 1 is that~ sharp transitions in the residual support plate.
signals can.be observed-at the locations of both dots.
In the following graphic,. Figure A-13, somewhat the
~
same logic could be applied in determining the flaw-like portion of the signal from the Mix 1'Lissajous pattern.
However, inasmuch as there is no sharp, clearly defined transition, coupled with the fact that the entry-lobe into the support plate is distorted on all of the raw frequencies, the dots should be placed as shown in Figure A-14.
This is a conservative approach and should be taken whenever a degree of doubt as to the dot placement exists.
It ie.noted that.by utilizing these techniques, identification of flaws is improved and that conservative amplitude measurements are promoted.
The Mix 1 traces which result from this approach confirms 1 the model of tsp ODSCC which represents the degradation' as a series of microcrack segments axially integrated by the bobbin coil; i.e.,
short segments of changing phase angle direction represent changes in average B&W NUCLEAR TECHNOLOGIES A-16 l
\\.
~.
d
/
depth with changing axial _ position.
This procedure may not yield;the maximum bobbin depth call. - If maximum depth is desired for information purposes, shorted I
segments of the overall crack:may have.to be evaluated to obtain'the' maximum depth estimate.
However,'the
~
peak-to-peak voltages as described herein'must be reported, even if a different segment is used for the
' depth call.
A.3.3 Alloy Property Changes This. signal manifests itself as part of.the support plate " mix ~ residual" in both the differential and
' absolute mix channels.
It'has~often been confused with copper deposit as the cause.
Such signals are often
^
i found at support plate intersections of operating plants, as well as in'some'model boiler test samples, and are not necessarily indicative of tube wall degradation.
Six support plate intersections from 1
I Plant A, judged as free of. tube wall degradation on'the basis of the mixed differential channel using the l
guidelines given in Section A.2.7 of this document, l
were pulled in 1989.: Examples of the bobbin coil field data are shown in Figure A-15 (inspection data from a plant with 7/8" diameter tubing).. The mix residual for this example is approximately 3 volts in'the
' {
differential mix channel and no discontinuity
]
suggestive of a flaw can be found in this channel.
An-offset in the absolute mix channel which cou'ld be.
confused as a possible indication is also present.
These signals persisted without any significant~ change even after chemically cleaning the OD and ID of the tubes.
The destructive examination of these intersections showed very, minor'or no tube wall degradation.
Thus, the overall " residuals" of-both the differential and the absolute mix channels were not indications of tube wall degradation.
One needs to examine the detailed structure of the " mix residual" (as outlined in Section A.2.7) in order to assess the
. j possibility that a flaw signal is present in the i
B&W NUCLEAR TECHNOLOGIES A-17 i
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J residual composite.
Verification of the integrity of TSP intersections exhibiting alloy property or artifact signals is accomplished by MRPC testing of a f
representative sample of such signals.
A.3.4 Denting and Copper Influences The South Texas Project Units 1 & 2 have not experienced significant corrosion-assisted denting nor do they have reported indications-indicative of copper deposits.
A.3.4.1 Dent Interference
[
(d)
I
} BP B&W NUCLEAR TECHNOLOGIES A-18
3 p-,
[
(d)
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w-r i
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(d)
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B&W NUCLEAR TECHNOLOGIES A-20
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A.3.5 MRPC Flaw Characterization The MRPC inspection of some support plate intersections with bobbin coil indications > 1.0 volts is required in order to verify the applicability of the alternate repair limit.
This is based on establishing the presence of ODSCC with minor IGA as-the cause of.the bobbin indications.
The signal voltage for MRPC data evaluation will be based on 20 volts.for the 100% throughwall 0.5" long EDM notch at all frequencies.
P The nature of the degradation and.its orientation (axial, or circumferential) will be determined from careful examination of the isometric plots of.the MRPC data.
The presence of axial ODSCC at the support plates has been well documented, but the presence of circumferential indications related to ODSCC at support.
B&W NUCLEAR TECHNOLOGIES A-21
~
i plate' intersections has also been established by tube-pulls at two. plants.. Figure-A-16 to A-18 show examples of single and multiple axial ODSCC'from' Plant S.
Figure A-19 is an example of a circumferential indication'related to ODSCC at a tube support plate-location from another~ plant.. If circumferential involvement results from circumferential cracks as i
opposed to multiple axial crack, discrimination between axial and circumferential1y oriented cracking can be
~
generally established'for affected arc lengths of about
-j 45 degrees-to 60 degrees or larger. -Axial cracking has
~
-been found by pulled tube exams for MRPC arcs of 150 i
degrees when the axial extent is significant, such as >
0.2 inch.
Pancake coil resolution is considered adequate for-l separation between circumferential and axial. cracks.
This can be supplemented by' careful interpretation of 3-coil results.
Since denting has not occurred at the South Texas Project Units 1&2 units, circumferential cracking is not expected to happen.
l The presence of IGA as a local effect directly adjacent to crack faces is expected to be indistinguishable from the crack responses.and as such of no structural consequence. 'When IGA exists as a general phenomenon,:
- the eddy current response is proportional to the volume of affected tube material, with phase angle j
corresponding to depth of penetration and-amplitude relatively larger than that expected for.small cracks.
The presence of distributed cracking, e.g.,
cellular SCC, may produce responses from microcracks of sufficient individual dimensions to be detected-but not' resolved by the MRPC, resulting in volumetric responses similar to three-dimensional degradation.
For hot let TSP locations, there is little. industry experience on the basis of tube pulls for volumetric degradation, i.e.,
actual wall loss or general IGA.
B&W NUCLEAR TECHNOLOGIES A-22 1
e n--w.-
,e
L f
For cold leg TSP locations, considerable experience'isL available for volumetric degradation-in the' form:of=
thinning of. peripheral tubes, favoring the;1ower TSP elevations.
Therefore, in the' absence.of confirmed pulled' tube experience to the contrary,: volumetric OD indications at hot-leg tube support plates should be considered to represent ODSCC.
4 A.3'6 Confinement of ODSCC/ IGA within the support Plate Region The measurements of axial crack lengths.from MRPC' isometrics can be determined using the'following analysis practices.
For the location of' interest, the low frequency channel (e.g., 10 kHz) is used to set'a local scale for measurement.
By establishing the midpoint of the support plate response, a reference point for indication location is established.
Calibration of the distance scale is accomplished by setting the displacement between the 10 kHz absolute; upper and lower support plate transitions equal to 0.75 inch.
A.3.7 Length Determination with MRPC Probes t
At the analysis frequency, 300 kHz, the ends of the crack are located using the slope-intercept method; i.e.,
the leading and trailing edges of the signal pattern are extrapolated to cross the null baseline (See Figure A-26).
The difference between these two-positions is the crack length estimate. _ Alternately,.
- the number of scan lines. indicating the presence of the flaw times the pitch of the rotating probe.provides a-conservative estimate of crack length which may then be corrected for beam spread.
B&W NUCLEAR TECHNOLOGIES A-23 s
+w-
~,
..w.
m,w..
.s..
< ~ - - -
I 1
l
. A.3.8 MRPC Inspection Plan The MRPC inspection plan will include the following I
upon' implementation of the ARCLrepair' limits:
Bobbin voltage indications > 1 volt, Large residuals,.
and' Dents >-5 volts A representative sample of 100 TSP intersections based on the following:
1)
Artifact signals (alloy property changes) spanning the' range of amplitudes observed during bobbini coil examination.
2)
Dented tubes at. TSP intersections with bobbin dent-voltages exceeding 5 volts.
3)
Bobbin indications less than 1 volt for-justification of these indications as typical-of-i ODSCC.
The 100 TSP intersections for MRPC inspection would be targeted toward a distribution on the order of 40 i
dents, 40 artifacts, artifacts, and 20 indications with bobbin voltages < l.0 volts; this distribution will be i
adjusted to reflect field observations as appropriate.
Consideration for expansion of the MRPC inspection l
program would be based on identifying unusual or unexpected indications such as clear circumferential
{
cracks.
In this case, structural assessments of the j
significance of the indications would be used to guide i
the need for further MRPC inspection.
A.3.8.1 3-Coil MRPC Usage f
?
It is Houston Lighting & Power's standard practice to use 3-coil MRPC probes,
[
incorporating a pancake coil, an axial B&W NUCLEAR TECHNOLOGIES A-24 e
t-
,.n y r g,,
s 8
i i
f preference coil,'and'a circumferential~
+
preference coil.
Comparisons for-oDSCC with Lbobbin-amplitudes exceeding:1.0 volts have shown that the pancake coll fulfills the need-for discrimination between' axial and circumferential indications, when compared-
=;
against the outputs of the preferred--
direction coils.
Pancake coils have been the basis for. reporting MRPC voltages for model boiler and pulled tube. indications in the ARC database; these-data permit semi-quantitative judgements on thenpotential significance ~of MRPC indications.
The requirement:for a
~
pancake coil is: satisfied by the single coili 2-coil, and 3-coil probes in common use for.
MRPC' inspections.
1 A.3.9 Noise Criteria
- i Quantitative noise criteria (resulting from electrical-noise,' tube noise, or calibration standard noise) should-be included in the data analysis procedures..
Actions should be taken to correct the data'by re-performing the calibration or re-inspecting;the-affected tube (s).
t Eddy current data acquired from active, tubes and calibrations standards shall be reviewed for the' i
presence of electrical and tube noise.
General eddy current data quality shall be monitored to ensure-that
~ ;
a minimum 3:1 signal-to-noise ratio (S/N) is' maintained.
This value of S/N is a commonly accepted industry value for data quality ensuring reliable-
[
signal detection.
)
i i
b B&W NUCLEAR TECHNOLOGIES A-25
..,_,m,-.
--e
A.3.9.1 ID. Chatter or Pilgering Noise.
Tubes identified with noise associated with ID chatter or pilgering in excess of 5 volts-peak-to-peak on Channel 1 shall also-be screened'using Channel 5.
A.3.9.2 Probe Noise Electrical noise due to a failing or' intermittent probe is readily recognizable as the noise signal often assumes the shape of a random square wave modulating the eddy
~
current signal.
Electrical noise in excess'of 0.3 volts peak-to-peak on Channel 1 will be rejected by the-analyst and the tube re-examined.
e 9
6
}
-1 B&W NUCLEAR TECHNOLOGIES A-26 J
a-p~
FIGURE A-1 PROBE WEAR STANDARD SCHEMATIC
[
i i
i I
h EJr Figure derived from EPRI TR-100407, Rev. 2A, Appendix B.
B&W NUCLEAR TECHNOLOGIES A-27
.rr--+
w
..c-.o
,s
,--i-
j FJGURE A-2 BOBBIN COIL AMPLITUDE ANALYSIS OF ODSCC AT TSP 36 061 V MIR 1 V 4.19 1:5 file 1 12 4.31
$50 Ehr CH 1 13 I
i m
s w %-
_M ses --unir-- ounn ]
5C i 2M l +e.15 l sl llIN
-~
EX7DfT TTM l TEC l SPED 19.83l In/see IRAIN 1C 12C 13C Vpp 1.21 M 37 444 Vpp 1.42 M 75 838 t
i t
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4
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I seterase essa s
5 t
I
(
5 5 Y
si 4
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ran a
a s
s s
u s
b b
~
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Figure derived from WCAP 13523 (23].
B&W NUCLEAR TECHNOLOGIES A-28
l FIGURE A-3 BOBBIN COIL AMPLITUDE ANALYSIS OF ODSCC INDICATION AT TSP-IMPROPER IDENTIFICATION OF FULL FLAW SEGMENT RESULTING IN REDUCED VOLTAGE MEASUREMENT WHEN COMPARED WITH FIGURE A-2 36 061 W TEC n!E 1 V 4.19 1:5 mix 1 12 4.31 554 I:ha og 1 13 SGEM ROW CK.
[$
RELF5E 015E SIDE y
S/E E EUHlima. DUTLET l 5C C =
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. g 82TDIT tot i TEc lj IOC SPED 19.83lIn/see ITRAIN
-~
11C 1R g
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Figure derived from WCAP-13523 (23),
i B&W NUCLEAR TECHNOLOGIES A-29 L
J
FIGURE A-4 BOBBIN COIL AMPLITUDE ANALYSIS OF ODSCC INDICATION AT TSP-IMPROPER IDENTIFICATION OF FULL FLAW SEGMENT RESULTING IN REDUCED VOLTAGE MEASUREMENT WHEN COMPARED TO FIGURE A-2 36 Of 1 V MIX 1 V 4.13 til MIX 1 12 4.31 558KhaCHI 13 1h
$/G m:MUNITem OUTIIT l 5
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EC 2M l +4.87 l el HIN CITENT TEN l TEC l]
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I
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B&W NUCLEAR TECHNOLOGIES A-30
FIGURE A-5 CORRECT PLACEMENT OF VOLTAGE SET POINTS ON HIX 1 LISSAJOUS TRACES FOR R18C103 31 niu a y os a v 4.es ssa che os is ts.se see ou of 2 st te (c
- l " " l'l llIN se l
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spun 1s.12JIn/ e Fin tec
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'4 tac VPP 2.0 M Es 164 Vpp 6.s4 E sa 7sz.
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Figure derived from WCAP-13523 (23).
B&W NUCLEAR TECHNOLOGIES A-31
FIGURE A-6 CORRECT PLACEMENT OF VECTOR DOTS ON HIX 1 LISSAJOUS TRACES FOR R22C40 x cn s y nts a v 4.4e ~ sis nix :
3 7.si sse ou cx :
2
-} d, I
2 E
EE-a2n i.e. = ssi itin E
Tec cxTo T tot I tre lj sPero 22.stlinn.e Iruin sac 12C 13C vpp t.23 EEI sts sa y,,
2.s2 M in an l
14C os s
e
/
N A-
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T, u-
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23.15 3ee mz CM 3 333 23.49 13e ou CM s 13e 11H 9
m e
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l Figure derived from WCAP-13523 (23).
i B&W NUCLEAR TECHNOLOGIES A-32 1
1 t
FIGURE A-7 INCORRECT PLACEMENT OF VECTOR DOTS ON MIX 1 LISSAJOUS TRACES FOR R18C103 35 cx y nix,
4.34 sse ex ch is i m.as see the cu a ss a
g b
era y
y i
E 5:
2-1/G maIEUNiimm OUTLET W
an i.e.es I l una
'J ExtexT tra i tre 1 7 J
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tac
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eam
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b rene s
a sls s
s i 1s i
i 35 i
. e i
e Figure derived from WCAP-13523 [23),
i B&W NUCLEAR TECHNOLOGIES A-33
i FIGURE A-8 INCORRECT PLACEMENT OF VECTOR DOTS ON HIX 1 LISSAJOUS TRACES FOR R22C40 3s cM v niz : v 4.4e ans nix ses 7.s sse che cu :
su b
f,c EE ce 2H l +e.as l l ll n.
s y3 (zTtxt ros l Ytc l sec setze 22.eslan/me raann 11e 2C ix vpp
.03 M 32, 84 vpp
.s2 M is3 22 14C p J6
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l
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d f
k W
N 2-s-
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n.s ses cna en a su n.<s 33e cnr cx s nae e
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as
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s.3: e en son
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/
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Q
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s' W
W y-3 7
a.
a o
Figure derived from WCAP-13523 [23].
B&W NUCLEAR TECHNOLOGIES A-34
l J
FIGURE A-9 INCORRECT MAXIMUM VOLTAGE DERIVED FROM PLACEMENT OF VECTOR DOTS ON TRANSITION REGION OF 550 kHz RAW FREQUENCY DATA LISSAJOUS TRACE FOR R42C44 gy ggg g y 3.3e 1:5 MIX 1 18
'*U dCw
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=
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6 12n -e.esi i Ilia g
4 EXTDef TG l fit l 22.86lIn/see (]TeAIN f
11C 3 PEED 12C 13C
- -~
{
vpp 1.65 8 103 39t Ypp 1.01 M 122 ese f
h-N d
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C-y i
w 14H 13.53 300 thz Of 3 de 13.53 13e Diz 015 44 s
13H
[
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8 ID
~
L 338W38 ecm= e.. em, em Ypp E.te EB es sit L
Ype 3.3e ETE se 7e n.as astimiseone 1
e C0LL r
y.
s
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^
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=
I
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s x
Figure derived from WCAP-13523 (23].
B&W NUCLEAR TECHNOLOGIES A-35
ii j
o FIGURE A-10 CORRECT PLACi. MENT OF VECTOR DOTS ON MIX 1 LISSAJOUS FIGURE FOR R42C44 34 - Of 1 V n!K 17 3.30 1:5 pig 1 le 6.77 55e De 001 23 E
g 8 EEL
,g M:
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g
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EXText TEM i Trc I J teC
~~
5 PEED 22.eE{tn/sec lTRAgu j.
tre ac I
13" k
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j vp, 2.81 EIE ses 43:
vpp 2.te EIB 118 ssa I
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- I f
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c 14M 13.53 300 na Of 3 et 13.53 13e ghs CM 5 es 136 tan ttu y
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)
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)
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'i i
f Figure derived from WCAP-13523 [23).
l B&W NUCLEAR TECHNOLOGIES A-36 i
I
FIGURE A-11 PLACEMENT OF VECTOR DOTS BASED SOLELY ON MIZ 1 LISSAJOUS FIGURE (NO SIGNIFICANT SHARP TRANSITIONS IN ANY OF THE RAW FREQUENCIES) - R10C44 3s cn 1 v nix 1 v 3.24 1:s nix 1 le 3.as ssa na cn 1 le y, g 1, g NE E
E is E : ID e.
g N
s/s mir aunn
=
2H 1 8.se l:1 liin 5
cxtort tot I tre i lii iec secto 21.etisw.= lTm4:n 11c 12e
.r 13C l
,I Ype 1.2s M 124 213 Vpp 1.88 E 111 57s
)
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+
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=
=f vp, 1.47 e as its ve, 3.es e are en aae. easim>= ms
- y l
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Figure derived from WCAP-13523 (23).
B&W NUCLEAR TECHNOLOGIES A-37
l l
y_tg,URE A-12 PLACEMENT OF DOTS MARKING MIX 1 LISSAJOUS FIGURE FOR R16C26 l
30 Ot 1 V On 5 V 2.12 1:5 n!X 1 le 6.15 554 th 08 1 28 g==qp SC.e m m a-1m.......__,,.__
1....,___
i
- ~
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=
r
{
vpp s.52.
E e6 56a vpp s.s3 E B 184 set I
f s I
(
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}-
?
7~
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13.19 Joe Da CH 3 45 21.10 13e th: 01 5 71 13
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=
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ua nu.
4 )
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1 Figure derived from WCAP-13523 [23).
B&W NUCLEAR TECHNOLOGIES A-38 J
L
(.
FIGURE A-13 INCORRECT PLACEMENT OF VECTOR DOTS MARKIN'G MIX 1
~
LISSAJOUS FIGURE FOR R30C74 35 CM 1 V CH 5 W 2.91 til MIX 1 to 4.02 558 Da CM 1 13 3G u e a
r AEllei E 015K mm 880E m:u
$/C meUH!rmum OUTLET 9
m
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spero 21.lelivsee Irmalm
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a P.c g
11c -~
vpp e.92 G T5 654 vpp 1.24 E 65 s4:
12c p
q j
(
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tac 14C 1
13.24 300 Eh2 CM 3 45 21.1e 130 per Ot 5 11
/
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p 14H -f=.
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12H 11H seg g
setets scamas a serta ese n
gg Vpp 2.42 8 26 E54 Vpp 2.74 E 322 ft 8*e se, 5 eeretste sess !
I en I
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,a.
3 gg, 4
W M
t rrH T,
p
.p r
y z
i Figure derived from WCAP-13523 [23].
B&W NUCLEAR TECHNOLOGIES A-39 l
6 L
FIGURE A-14 CORRECT PLACEMENT OF DOTS TO EFFECT MAXIMUM VOLTAGE - R30C 35 CH 1 V Of 5 V 2.99 1:5 MIX 1 18 8.72 554 na CM i 21 i
dlMagga n-c.c l -.....,
lc trrorr itM i Ic l g
-3
,, m m.,it. rm i-n
7
..c 11C V.,
1.18 M 72 714 Ypp 1.96 e4 SE 1x m
s h
d i
d 1x m,-
7,
,.e
~
13.24 384 thz 04 3 45 21.18 13e Os CM S
?!
1.
13 y
y un vp, s.3s,e 2
sa v,,
s.es e 2n e,
s s.. s m esen: i m s
i m'
h 5"
N g'
,~
T, r
V
.x i
Figure derived from WCAP-13523 (23).
B&W NUCLEAR TECHNOLOGIES A-40
FIGURE A-15 EXAMPLE OF DOBBIN COIL FIELD DATA -
MIX RESIDUAL DUE TO ALLOY CHANGE 33 CH 1 V MIR 2 Y i 4.98 444 Ehr Qt 1 243 i 3.22 2:6 n!X 2 42 x
t sG=.-==. m Ca.
f 400/100 Absolute
'cD = ois= = = sicC - =
f 13/G memitem m]
Eis i. i,mg cEron scG i o. a SPEED, *****lin/sec~
W
[-
i vp, 3.31 OCG 13e et ve 2.1s DEG 2s!
es e
)
[
(-
/
M M
/W e
04 A
i i
I I
T r-P-
x e
~~
4.15 1:5 n!E 1 252 19.45 200 01 Of 3 359 (
[
400/100 Differential
.-=
l 1
M e
~
Mt4tAsi 7
Ypp 2.87 OCG 144 et Vpp 7.48 OCG 59 et u.4 a.,
s at: esser eses 8Caesag gifts 480 t.
f a
I
(
(
l
(
!ste a m a,ase....,
e i
s can
. p W
'ze m
g
- fe s 4
.a.
L s
1 1
1 1
3 Figure derived from WCAP-13523 [23).
i D&W NUCLEAR TECHNOLOGIES A-41
~...
FIGURE A-16 EXAMPLE OF MRPC DATA FOR SINGLE AXIAL INDICATION (SAI)
ATTRIBUTED TO ODSCC - PLANT S l 2e cu 4 v cu s v 4.as see na m 4 se g) 4 m
> g3 gug len at b7bI M J'
f SBftfles:
41 5/5 13:1T gm j
st. OF Icas Laut$s 25 misAus ett/sCass stW884(ItW8843 31 l g.36 ) l
[fgg 1
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6=
TD8 i TDs I lii O~
tatt!
-I lTIlGGERl l CIRC lAx!AL M S.31llMeet $ 1N s.
Vpp 1.21 DEC 32 et i
w w
4D 7
i
(.-
t 7
l Ax!E VfD1 I
CIRCUN D D(Ta t UTs 42 DEG 16 3
e.sa I
.j t
A m.s.ms.
/.3
-e.s7 ase Axse Taur axis. urs e.s3 in i
i!
I N
1 f
l Figure derived from WCAP-13523 (23].
B&W NUCLEAR TECIINOLOGIES A-42
FIGURE A-17 HRPC DATA FOR SINGLE AXIAL ODSCC INDICATION (SAI)
PLANT S L
3e cw 4 y al s v 4.4s see n a cu e es 3
M g
8 t]
- aa as 1
- estarias esters =.
as e
y s/s tart es. w sus uus n
assmus ris/sans 3 ysissageriass gg g 4,3y l,g llgg-71 carort ros i to #
y N
lTRIGGD]
kitClulg SPG9 8.34lIn/sec lTRALM cens Vpp 8.71 OCE 37 em
/
1 L,
b 3
/
J lg min via e
4
- p catcara m:4. tris 274 ots 1
e.42
,i In.e.c
-* 87 358
/
uru. ruct
~
MIAl. EXT 0.24 IN
=
j r
1 N
l
^
l I
Figure derived from WCAP-13523 (23).
B&W NUCLEAR TECHNOLOGIES A-43
FIGURE A-18 MRPC DATA FOR MULTIPLE AXIAL ODBCC INDICATIONS (MAI) PLANT S d
38 CH 4 V Of 3 V 2.as 300 Dur 014 383 Q
M g
b 3 El 88a4 I setsflats at 7 08f8f34Es 280
$/
g
- 88. er Stas Lluca 83 plamat tf5/9 tans 835/844 t135/188) 3pt l,0.41l l lllN
~{
>(
T DTDtT M i TDI l }
t catu I
liasocotl lctaclung se-e.asitar e rmtu v, e.55 DEG 57 se a
b 8
K 3se aan ssa CIRQsfUtDiTlE EXTs 31 DEG
-4.43 2 M.25 0.59 8
usn faux M14 EXT 8.44 IN -
f Figure derived from WCAP-13523 (23).
B&W NUCLEAR TECHNOLOGIES A-44 l
l
FIGURE A-19 MRPC DATA FOR.CIRCUMFERENTIAL ODSCC INDICATIONS AT DENTED UPPER AND LOWER TSP EDGES
- *' i
- 4
- * - i un ai m.
=a.
i.
on.
su a -...
s= -.e a merto,. ma ces 1
,, e.se Y
w
- <f
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~'
M
"-r
<ss,
s I
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ni e e.m i
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)
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oseen, se - a ps as au : so. se one-a
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- ER M
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=
1
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ss'
, 4.sa sam e as E NE ' !
! ""."., 'i n
- J,4!=, a 4
,u r
.it us
==
=
esta i
e sur a,ega Figure derived from hCAP-13523 [23).
l i
l B&W NUCLEAR TECHNOLOGIES A-45 i
i
FIGURE A-20 EXAMPLE OF BOBBIN COIL FIELD DATA - FLAW SIGNALS FOR ODBCC AT DENTED TSP INTERSECTION FROM PLANT A
- e.,atz, y ot ? V 1.M
- ee m m i de 5.M M msa y
E 2M f.e.ee g,l ggni Omti rtei m : 7
- UD l***ilW.se eraats
{
5 l
=
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i I
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W 9-1y ll,
~
Figure derived from WCAP-13523 [23].
B&W NUCLEAR TECHNOLOGIES A-46
FIGURE A-21 EXAMPLE OF BOBBIN COIL FIELD DATA - FLAW SIGNALS FOR ODSCC AT DENTED TSP INTERSECTION FROM PLANT A 1
3e nas I v CM 1 W 18.54 ees trts CM i et 1.13 15 nit t is
~
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w 7
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41 6
Wee 6.73-M 172 esti vee 4.49 M 144 cwt 8.e4 400 ms 08 8 241 4.76 1:1 RtX 1 245 t
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-.m-Figure derived from WCAP-13523 (23).
B&W NUCLEAR TECHNOLOGIES A-47
FIGURE A-22 EIAMPLE OF BOBBIN COIL FIELD DATA - FLAW SIGNALS FOR ODBCC AT DENTED TSP INTERSECTION PROM PLANT A
.e, nas a y os 7 v 1.15
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an 1 4.se til
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5
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4
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Figure derived from WCAP-13523 [23).
l 1
i B&W NUCLEAR TECHNOLOGIES A-48 l
i
i.
FIGURE A-23 EXAMPLE OF BOBBIN COIL FIELD DATA - FLAW SIGNALS FOR ODSCC AT DENTED TSP INTERSECTION FROM PLANT A e
44 sta 1 v O1y it.65 400 th: D1 275 5.00 1:5 alt I 2 T2 j
etIL ogsE - St0C 8
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i
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i Figure derived from WCAP-13523 (23).
B&W NUCLEAR TECHNOLOGIES A-49
1 l
FIGURE A-24 EXAMPLE OF BOBBIN COIL FIELD DATA - FLAW BIGNALS FOR ODBCC AT DENTED TSP INTERSECTION FROM PLANT A
{
se als t v CM 3 V 9.15 ese CN CM i s2 5.7e 5 at I
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==
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l m
e e
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Figure derived from WCAP-13523 [23).
t l
B&W NUCLEAR TECHNOLOGIES A-50 l
l i
FIGURE A-25 EXAMPLE OF BOBBIN COIL FIELD DATA - FLAW SIGNALS FOR ODSCC AT DENTED TSP INTERSECTION FROM PLANT A 38 1 RIE I v Os 7 y e.e4 ese De Ot t 34 4.M 8:s att t a
38 l ae.ee lel 4:se 9
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l
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l e
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l Figure derived from WCAP-13523 [23].
i B&W NUCLEAR TECHNOLOGIES A-51 l
FIGURE A-26 LOCATION OF ONE END OF AN INDICATION USING AN RPC PROBE
., CH 2 VUtf.'. CH 2HDit!*l CHMEL DC - 2 lfD G ROh SS CEL 3
XY D!ruW l
CHme - 2 FPEG 400 kHz
. ".l SPf84 156 g
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WM N
q........
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3 SPfW 155 1
- i
' ft0TRTION - 354 CEO
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-y n10w mne eest cume - aH
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FilEG 400 kHz SP584 156
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ROTRTICH - 554 DEO
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SYSTEM CDNElghggIlgl l f' MfM - CEFALLT 4 ef 05H= 4 12:23:29 ful JfM33RY 1 1984 f23:.Le FC '1 112 3 4 5 6 7 5 1,
i.
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i i 33
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l 4
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aute 347s i
Figure derived from WCAP-13523 [23).
B&W NUCLEAR TECHNOLOGIES A-52
-