ML023640319
| ML023640319 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/31/2002 |
| From: | Bamford W, Becker F, Kaihwa Hsu, Petsche J, Server W ATI Consulting, Electric Power Research Institute, Westinghouse |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| WCAP-15984-NP | |
| Download: ML023640319 (38) | |
Text
Westinghouse Non-Proprietary Class 3 WCAP-15984-NP December 2002 Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units I and 2
WESTINGHOUSE NON PROPRIETARY CLASS 3 WCAP-15984-NP Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2 ATI Consulting William Server Westinghouse Electric Company Warren Bamford K. Robert Hsu Joseph F. Petsche EPRI NDE Center F. L. Becker December 2002 Reviewer:
ill*i-/
P. L. Strauch Structural Mechanics Technology Approved:
S O-.
/S. A. Swamy, Manager Structural Mechanics Technology Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355
© 2002 Westinghouse Electric Company LLC All Rights Reserved 6121 -NonProp-120302
TABLE OF CONTENTS 1
IN TR O D U CTIO N.................................................................................................................
1-1 2
TECHNICAL APPROACH..................................................
2-1 3
FRACTURE ANALYSIS METHODS AND MATERIAL PROPERTIES................................
3-1 3.1 Stress Intensity Factor Calculations..................................
3-1 3.2 Fracture Toughness.......................................................................................
.... 3-1 3.3 Irradiation Effects................................................................
3-2 4
FLANGE INTEGRITY...............................................
4-1 5
ARE FLANGE REQUIREMENTS NECESSARY?......................
5-1 6
SAFETY IMPLICATIONS OF THE FLANGE REQUIREMENT........................
6-1 7
R EFE REN C E S...........................................................................................................................
7-1 APPENDIX A REACTOR PRESSURE VESSEL INSPECTION RELIABILITY...........................
A-i 6121-NonProp-120302 December 2002 December 2002 6121 -NonProp-120302
i-1 I
INTRODUCTION 10 CFR Part 50, Appendix G contains requirements for pressure-temperature limits for the primary system, and requirements for the metal temperature of the closure head flange and vessel flange regions The pressure-temperature limits are to be determined using the methodology of ASME Section XI, Appendix Q but the flange temperature requirements are specified in IOCFR50 Appendix CL This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the pre-service hydrostatic test pressure, which is 621 psig for a typical PWR, and 300 psig for a typical BWR.
This requirement was originally based on concerns about the fracture margin in the closure flange region.
During the boltup process, outside surface stresses in this region typically reach over 70 percent of the steady state stress, without being at steady state temperature. The margin of 120'F and the pressure limitation of 20 percent of hydrotest pressure were developed using the K13 fracture toughness, in the mid 1970s, to ensure that appropriate margins would be maintained.
Improved knowledge of fracture toughness and other issues which affect the integrity of the reactor vessel have led to the recent change to allow the use of KI, in'the development of pressure-temperature curves, as contained in ASME Code Case N640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division I".
Figure 1-1 illustrates the problem created by the flange requirements for a typical PWR heatup curve. It is easy to see that the heatup curve using Kic jýrovides for a much higher allowable pressure through the entire range of temperatures. For this plant, however, the benefit is negated at temperatures below RTNDT
+1 20'F because of the flange requirement of 10 CFR Part 50, Appendix G. The flange requirement of 10 CFR 50 was originally developed using the Kia fracture toughness, and this report will show that use of the newly accepted KI, fracture toughness for flange considerations leads to the conclusion that the flange requirement can be eliminated for Sequoyah Units I and 2.
Introduction 6121 -NonProp-120302 December 2002 -
2500 2250 200O 1750 1500 1250 1000 750 500 I -
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Figure 1-1 Illustration of the Impact of the Flange Requirement for a Typical PWR Plant 6121-NonProp-120302 1-2 December 2002
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TECHNICAL APPROACH The evaluation to be presented here is intended to cover the Sequoyah Units I and 2 reactor vessels Fracture evaluations have been performed on the closure head geometry specific to these units, and results will be tabulated and discussed The geometry of the closure head region for Sequoyah Units I and 2 is shown in Figure 2-1.
Stress analyses have been performed, and these stress results were used to perform fracture mechanics evaluations. The highest stress location in the closure head and vessel flange region is in the head, just above the bolting flange. This corresponds with the loc'ation of a weld The highest stressed location is near the outside surface of the head in that region, 'and so the fracture evaluations have assumed a flaw at this location.
The goal of the evaluation is to compare the integrity of the closure head during the boltup and the heatup and cooldown process, to the integrity during steady state operation. The question to be addressed is:
With the higher K1, fracture toughness now known to-be applicable, is there still a concern about the integrity of the closure head during boltup?
Technical Approach December 2002 Technical Approach 61!21 -NonProp-120302 December 2002
2-2 TOP HEAD DOME TORUS TO FLANGE WELD 7
C VESSEL FLANGE TO UPPER SHELL WELD D
UPPER HEAD REGION ID 170.88 NOTE: ALL DIMENSIONS ARE IN INCHES Figure 2-1 Geometry of the Upper Head/Flange Region of the Sequoyah Units 1 and 2 Reactor Vessels I echnical Approach 6121-NonProp-120302 December 2002 Sequoyah Units I and 2 A
88.1 B
6.89 C
29.72
3-1 3
FRACTURE ANALYSIS METHODS AND MATERIAL PROPERTIES The fracture evaluation was carried out using the approach suggested by Section XI Appendix G (Ref. 1)
A semi-elliptic surface flaw was postulated to exist in the highest stress region, which is at the outside surface of the closure flange. The flaw depth was assumed to encompass a range of depths into the wall thickness, and the shape was set at a length six times the depth.,
3.1 STRESS INTENSITY FACTOR CALCULATIONS One of the key elements of a fracture evaluation is the determination of the dnving force or stress intensity factor (KI). In most cases, the stress intensity factor for the integrity calculations utilized a representation of the actual stress profile rather than a linearization. The stress profile was represented by a cubic polynomial:
c°(x)=A 0 +A, x+ A 2 (t
+A 3 t (3-1) where:
x
=
is the coordinate distance into the wall, in.
t
=
wall thickness, in a
=
stress perpendicular to the plane of the crack, ksi A,
=
coefficients of the cubic fit For the surface flaw with length six times its depth, the stress intensity factor expression of Raju and Newman (Ref. 2) was used. The stress intensity factor K, (4) can be calculated anywhere along the crack front. The point of maximum crack depth is represented by 4 = 0, and this location was found to also be the point of maximum K, for the cases considered here. The following expression is used for calculating K, (0), where 4) is the angular location around the crack. The units ofKI (K ) are ksii~n.
K G (a/c, a/t, t/R, 0) Aj aJ (3-2) 1Q J=l The magnification factors G, (0)), G2 (4), G 3 (4) and G 4 (4)) are obtained by the procedure outlined in reference (2). The dimension "a" is the crack depth, and "c" is the crack length, while t is the wall thickness.
3.2 FRACTURE TOUGHNESS Another key element in a fracture evaluation is the fracture toughness of the material. The fracture toughness has been taken directly from the reference curves of Appendix A,Section XI. In the transition temperature region, these curves can be represented by the following equations:
KI = 33.2 + 20.734 exp. [0.02 (T-RTNDT)]
(3-3)
Fracture Analysis Methods and Material Properties 6121 -NonProp-120302 December 2002
3-2 KI, = 26 8 + 12.445 exp. [0.0145 (T-RTNDT)]
(3-4) where K], and K], are in ksiJ/.
The upper shelf temperature regime requires utilization of a shelf toughness which is not specified in the ASME Code. A value of 200 ksiJn has been used here. This value is consistent with general practice in such evaluations, as shown for example in reference 3, which provides the background and technical basis of Appendix A of Section X1.
The final key element in the determination of the fracture toughness is the value of RTNDT, which is a material parameter determined from Charpy V-notch and drop-weight tests.
The value of RTNDT for the closure flange region of the Sequoyah units was obtained from certified material test reports and the results are shown in Table 3-1. The highest value was 57F, and so this value was used for the illustrations to be discussed in Sections 4 and 5.
3.3 IRRADIATION EFFECTS Neutron irradiation has been shown to produce embrittlement which reduces the toughness properties of reactor vessel steels. The decrease in the toughness properties can be assessed by determining the shift to higher temperatures of the reference nil-ductility transition temperature, RTNDT.
The location of the closure flange region is such that the irradiation levels are very low and therefore the fracture toughness is not measurably affected a'c'e Fracture Analysis Methods and Material Properties 6121-NonProp-120302 December 2002 I
I-
+
I-3-2
4-1 4
FLANGE INTEGRITY The first step in evaluation of the closure head/flange region is to examine the stresses. The stresses which are affected by the boltup event are the axial, or meridional stresses, which are perpendicular to the nominal plane of the closure head to flange weld. The stresses in this region during steady state operation are summarized in Table 4-1.
The boltup is the key condition to review here, in comparison with steady state operation, since the flange requirement applies to boltup conditions. No other transients result in stresses in this region at low temperatures. One might suggest that the cooldown might be of similar concern, but the boltup is governing for a number of reasons:
I.
The heatup and cooldown transient is structured to ensure generous margins are maintained (SF = 2) for a large flaw in the irradiated beltline region. This is a more governing condition than the unirradiated flange region.
- 2.
The cooldown transient has much higher temperatures in the head region than the boltup, and
- 3.
The thermal stresses that are produced tend to counteract the boltup stresses; that is, they are tensile on the inside surface and compressive on the outside surface Table 4-1 provides a comparison of the stresses at boltup with those at steady state. It is easy to see that the stresses at boltup are mostly bending, with a very small membrane stress. As the vessel is pressurized, the membrane stresses increase. These results were taken from a finite element analysis of the heatup/cooldown process, and the boltup was determined to be the most limiting time step of the entire heatup/cooldown transient. The combination of the pressurization and the heatup process tends to reduce the stresses as the transient proceeds.
The relative impact of these stresses can best be addressed through a fracture evaluation. A semi-elliptic surface flaw was postulated at the outer surface of the closure head flange, and the stress intensity factor, K, (or crack driving force) was calculated. The results are shown for the boltup condition in Figure 4-1.
It can be seen that the applied stress intensity factor at boltup reaches a maximum for a flaw about half way through the head thickness, and then decreases as the flaw extends into the lower stress region near the inside surface of the head. The maximum value of the stress intensity factor was calculated to be 31 ksini, at a postulated flaw depth of 42 percent of the wall.
It will be useful to highlight the difference in the integrity story for the head region using the two values of fracture toughness. The boltup temperature for a typical PWR is 60'F, so if RTNDT = 5°F the ASME reference toughness values are Kia = 54.4 ksi-jlfin and K1, = 95.5 ksi-%ii Using the K1, toughness (which was the basis for the original flange requirements) it can be seen that the toughness exceeds the applied stress intensity factor by at least a factor of 1.75, for flaws of any depth in the head thickness. The smallest margin of 1.75 occurs for a flaw 42 percent of the wall thickness; for other flaws the margin is larger.
Flange Integrity December 2002 6121 -NonProp-120302
4-2 Using the K1, toughness, which has now been adopted by Section XI for P-T Curves, it can be seen that there is also a significant margin between the fracture toughness and the applied stress intensity factor at virtually all crack depths. In this case the margin exceeds 3, which is a very generous margin Another objective of the requirements in Appendix G is to assure that fracture margins are maintained to protect against service induced cracking due to environmental effects. Since the governing flaw is on the outside surface (the inside is in compression) where there are no environmental effects, there is even greater assurance of fracture margin. Therefore, it may be concluded that the integrity of the closure head/flange region is not a concern for the Sequoyah units using the KI, toughness There are two possible mechanisms of degradation for this region, thermal aging and fatigue.
Effect of Fatigue. The calculated design fatigue usage for this region is less than 0 1, so it may be concluded that flaws are unlikely to initiate in this region.
la c~e Flange Integrity December 2002 Flange Integrity 6121 -Non Prop-1 20302 December 2002
4-3 Table 4-1 Stress Distributions for the Closure Flange Region - Sequoyah Units I and 2 Distance Boltup Stress Steady State (xlt)
(ksi)
(2250 psi) 0 (ID)
-14.38 328 0.1
-1077 440 0.2
-7.83 5.70 0.3
-5.14 6.30 04
'-2.66 7.10 0.5
-0.26 8.10 0.6 2.16 8.60 0.7 4.72 9.70 0.8 7.54 10.60 0.9 11.24 120 1.0 (OD) 19.70 16.44 Flange Integrity 6121-NonProp-120302 December 2002
4-4 Figure 4-1 Crack Driving Force as a Function of Flaw Size: Outside Surface Flaw in the Closure tlead to Flange Region Weld for Sequoyah Units I and 2 (stress intensity factor units are ksi-,in) 62lange Integrity 6121 -NonProp-120302 W 4 LOOP REACTOR VESSEL CLOSURE HEAD/FLANGE WELD BOLT-UP OUTSIDE SURFACE STRESS INTENSITY FACTOR vs a/t 35 cc 30 0
25
_20o z
III S15 rn U) 5 0
006 012 018 024 03 036 042 048 a/t December 2002
5-1 5
ARE FLANGE REQUIREMENTS NECESSARY?
Using the K,, curve can support the elimination of the flange temperature requirement. This can be illustrated by examining the stress intensity factor change for a postulated flaw as the vessel is pressurized after boltup, progressing up to steady state operation.
The stresses at the region of interest are shown in Table 4-1, for steady state operation, as well as boltup.
Included here are the stress distributions through the wall, showing that the highest stress location for this region is the outer surface.
As the vessel is pressurized, the stresses in the closure flange region gradually change from mostly bending stresses to a combination of bending and membrane stresses. The stress intensity factor, or driving force, increases for a postulated flaw at the outside surface, as the vessel is pressurized.
A direct comparison between the original basis for the boltup requirement and the new KI, approach is provided in Table 5-1. This table provides calculated boltup requirements for all the designs, using a safety factor of 2, and a reference flaw depth of a/t = 0.10, which was used by Randall as the basis for the original requirement (Ref. 11) Before discussing the table, it will be helpful to discuss the basis for the reference flaw, in light of current technology, and using the results of the Performance Demonstration Initiative.
Basis for The Reference Flaw Size. Regulatory Guide 1. 150 stimulated improvement in examinations of the clad to base-metal interface. The same techniques have been used for more than 10 years for reactor vessel head examinations performed from the outside surface. Capability demonstrations for the clad to base-metal interface have been conducted at the EPRI NDE Center since 1983. These demonstrations were performed initially for the belt-line region. However, similar techniques are used for both the vessel belt-line and the reactor vessel head, although the head exams are done manually.
[
] a,c,e December 2002 Are Flange Requirements Necessary?
6121-NonProp-120302
5-2 Iace Are Flange Requirements Necessary?
December 2002 61!21!-Non Prop-120302
5-3 Table 5-1 Comparison of Various Plant Designs Boltup Requirements T -
RTNDT (°F)
T - RTNDT (OF)
K K with using KI, using Kia Plant (alt =.])
SF=2 (a/t =.10)
(a/t =.1O)
CE 30.0 60.0 13 68 B&W 39.4 79.8 41 100 W 4 Loop 197 39.4 0
1 W 3 Loop 194 38.8 0
25.1 50.2 0
43
- All units in ksiVn" Are Flange Requirements Necessary?
6121t-NonProp-120302 December 2002
5-4 Figure 5-1 Probability of Correct Rejection/Reporting (PCR) Considering Passed plus Failed Candidates, Appendix VIII from the Outside Surface. Reporting Criterion A' = 0.15 inch Are Flange Requirements Necessary?
December 2002 Are Flange Requrements Necessary?
61!2 l-NonProp-120302 December 2002
5-5 a c,e Figure 5-2 Probability of Correct Rejection/Reporting (PCR) Considering Only Passed Candidates, Appendix VIII from the Outside Surface. Reporting Criterion A' = 0.15 inch.
December 2002 Are Flange Requirements Necessary?
6121 -NonProp-120302
6-1 6
SAFETY IMPLICATIONS OF THE FLANGE REQUIREMENT There are important safety implications which are associated with the flange requirement, as illustrated by Figure 6-1. The safety concern is the narrow operating window at low temperatures forced by the flange requirement. The flange requirement sets a pressure limit of 621 psi for a PWR (20 percent of hydrotest pressure). Thus, no matter how good the toughness of the vessel, the P-T limit curve may be superceded by the flange requirement for temperatures below RTNDT + 120'F. This requirement was originally imposed to ensure the integrity of the flange region during boltup, but Section 4 has shown that this is no longer a concern.
The flange requirement can cause severe operational limitations when instrument uncertainties are added to the lower limit (621 psi), for the Low Temperature Overpressure Protection system of PWRs. The minimum pressure required to cool the seals of the main coolant pumps is 325 psi, so the operating window sometimes becomes very small, as shown schematically in Figure 6-1. If the operator allows the pressure to drop below the pump seal limit, the seals could fail, causing the equivalent of a small break LOCA, a significant safety problem. Elimination of the flange requirement will significantly widen the operating window for most PWRs.
An example will be provided to illustrate this situation for an operating PWR plant, Byron Unit 1. This is a forging-limited vessel at 12 EFPY, with a low leakage core, and low copper weld material in the core region. The vessel has excellent fracture toughness, which means that the flange notch is very prominent, as shown in the vessel heatup curve of Figure 6-2. As illustrated before in Figure 6-1, Byron has the LTOP setpoints significantly below the flange requirement of 621 psi, because of a relatively large instrument uncertainty. The setpoints of the two power operated relief valves are staggered by about 16 psi to prevent a simultaneous activation. The two PORVs have different instrument uncertainties, and for conservatism the higher uncertainty is used. A similar situation exists for cooldown, as shown in Figure 6-3.
Elimination of the flange requirement for Byron Unit I would mean that the PORV curve could become level at 604/587 psig, which are the leading/trailing setpoints to protect the PORV downstream piping, through the temperature range of the 350'F down to boltup at 60'F. The operating window between the leading PORV and the pump seal limit rises from 121 psig (446-325) to 262 psig (587-325). This change will make a significant improvement in plant safety by reducing the probability of a small LOCA, and easing the burden on the operators.
This is only one example of the impact of the flange requirement. Every operating PWR plant will have a different situation, but the operational safety level will certainly be generally improved by the elimination of this unnecessary requirement. The flange impact for Sequoyah Unit 2, for example, is shown in Figures 6-4 and 6-5 [13].
Safety Implications of the Flange Requirement December 2002 6121-NonProp-120302
/ '- Heatup Curve Instrument Uncertainty 621
\\
Operating Window Pump Seal Limrt 325 psI RTr&+1 20 Temperature Figure 6-1 Illustration of the Flange Requirement and its Effect on the Operating Window for a Typical Heatup Curve Safety Implications of the Flange Requirement December 2002 6121 -NonProp-120302 6-2
6-3 LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 5P-5933 (using murv. capuO(l data)
LIMITING ART VALUES AT 12 EFPY:
114T,.7OF° 3/4T, 600F 2500 o
I
_1 _______
2250
- 2000, fI I useats illI I1 I I II-litt I1 I
.1-4 111111 I jilt:
LEAI TZll LIXII I I I /:U -I 1500 I-t11-1-t 1111111 1 #
11 1 1 U NAC I TABLE H
073RATION-IZATP RATE ou TO ll 1/Er.
Ii fltli1il l I I I It ACCIPYABLI
,PORV LTOP SETPOITS ITgP rPrrifl RCP SEAL LMIT: 325 E CtIII¢ALIt-LIMIT? 35lll 0 1 IOiMlIUICI ET1101TATIC ?NO?
Il 1111PIIRAT l7 l1I r)
T111,ll 5i113lrlS Fall N
p 10012.
gP oi FPy ur 50 100 150
-200.
0 250 - 300 350
-400 450 500 Indicated Temperature (Deg.F)
Figure 6-2 Illustration of the Actual Operating Window for Heatup of Byron Unit 1, a Low Copper Plant at 12 EFPY December 2002 Safety Implications of the Flange Requirement 6121-NonProp-120302 1250 4 rn 4.)
1000 11750'-
500 "250 0
1YIJJYJ~L~f~
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l I I
I I l I I
I I
6-4 LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 5P-5933 (usig urv :=psule data)
LIMITING ART VALUES AT 12 EFPY:
2500 2250 t20O
-- 2000 I /5U 1/4T. 70*F 3/4T. 60*F
-I 9
z0118.1s s ssI III PUNACCEPTABL OPERATION
'i I
ACCEPTABLE OPiRATION
-1T7 7 1
1 1 1il til jil If 11
-- PORV ILTOP SETPOINTS LTOP OPERATING-I 1 1 1 RCP SEAL LIM*J: 323r Fal I I t I I I I I I I I I I I--T-[-
0 I I 1EEMNIUANLT.J 0
50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.F)
Figure 6-3 Illustration of the Actual Operating Window for Cooldown of Byron Unit 1, a Low Copper Plant at 12 EFPY December 2002 Safety Implcations of the Flange Requirement 6121-NonProp-120302 1500 1250 1000 rn rn C
C.)
COOL:OVN
- r/E!r 750 500 250 i
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2500 2250 4--- - -
2000 -
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S1250 CL CD
= 1000 750 500 250-0 Heatup Rate 60 Deg. FIHrI Heatup Rate 100 Deg. F/Hr Operim Version 5 1 Run 27850 Leak Test Limit
... Unacceptable Operation S....,
..... [Acceptable Operation "Critical Limit 60 Deg. F/Hr Critical Limit 100 Deg. F/Hr Criticality Limit based on q
inservice hydrostatic test temperature (203 F) for the service period up to 32 EFPY
-1 0
50 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg. F)
Figure 6-4 Illustration of the Flange Notch for Sequoyah Unit 2, Heatup Curve, without Instrument Uncertainties 1131 December 2002 Safety Implications of the Flange Requirement 6121 -NonProp-120302 6-5 550 FBolt.p empj
6-6 2500 Operlim Version 5 1 Run 27850]
2250 Acceptable I Unacceptable Operation 2000 Operation 1750 U)
S 1500 (V
LO (A
E 1250 125 Cooldown
"-0 Rates 2,
F/Hr steady-state
= 1000
-20
-40
-60
-100 750 500 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 6-5 Illustration of the Flange Notch for Sequoyah Unit 2, Cooldown Curves, without Instrument Uncertainties 1131 Safety Implications of the Flange Requirement 6121-NonProp-120302 December 2002
7-1 7
REFERENCES
- 1.
ASME Boiler and Pressure Vessel Code,Section XI, Appendix Q 1996 Addenda, ASME, New York.
- 2.
Raju, 1. S. and Newman, J. C. Jr., "Stress Intensity Factor Influence Coefficients for Internal and External Surface Cracks in Cylindrical Vessels," Trans. ASME, Journal of Pressure Vessel Technology, Vol. 104, pp 293-98, 1982.
- 3.
Marston, T. Ut-, ed., "Flaw Evaluation Procedures: ASME Section XI," Electric Power Research Institute Report EPRI-NP-719 SR, August 1978.
- 4.
Mitchell, M.A., "RPV P-T Limits and RPV Flange Requirements; Potential Exemptions from the Requirements of 10 CFR Part 50, Appendix G," presentation to ASME Boiler and Pressure Vessel Code,Section XI, Working Group on Operating Plant Criteria, Hollywood, FL, September 10, 2002.
- 5.
Nanstad, R.K., et al., Preliminary Review of Data Regarding Chemical Composition and Thermal Embrittlenment ofReactor Vessel Steels, ORNL/NRC/LTR-95/1, Oak Ridge, TN, January 1995.
- 6.
DeVan, M.J., Lowe, Jr., A.L., and Wade, S., "Evaluation of Thermally-'Aged Plates, Forgings, and Submerged Arc Weld Metals," Effects ofRadiation on Materials: 16th International Symposium, ASTM STP 1175, Philadelphia, PA, 1993.
- 7.
Kirk, M., "Revision of AT 30 Embrittlement Trend Curves," presented at the EPRI MRP/NRC PTS Re-Evaluation meeting in Rockville, MD, August 30, 2000.
- 8.
Charpy Embrittlement Correlations - Status of Combined Mechanistic and Statistical Bases for U.S. RPVSteels (MRP-45); PWR Materials Reliability Program (PWRMRP), EPRI, Palo Alto, CA: 2001, 1000705.
- 9.
ASTM E 90002, "Standard Guide for Predicting Radiation-Induced Transition Temperature Shift for Reactor Vessel Materials, E706 (IIF)," Annual Book ofASTM Standards, Vol. 12.02.
- 10.
Langer, R., et al., "A Survey of Results on Aging Experiments of Pressure Vessel Materials,"
presentation at the ATHENA Workshop, Madrid, September 2002.
- 11.
Randall, N., Abstract of Comments and Staff Response to Proposed Revision to 10 CFR Part 50, Appendices G and H, Published for Comment in the Federal Register, November 14, 1980.
- 12.
WCAP-1 5293, Revision 1, "Sequoyah Unit I Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," J. H. Ledger, April 2001.
- 13.
WCAP-15321, Revision 1, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," J. H. Ledger, et al., April 2001.
References December 2002 6121-NonProp-120302
A-I APPENDIX A REACTOR PRESSURE VESSEL INSPECTION RELIABILITY*
F. L. Becker EPRI Charlotte NC ABSTRACT
- Presented at the Joint EC-IAEA Technical Meeting on Improvements in Inservice Inspection Effectiveness, Pettan, The Netherlands, November 2002, to be published.
Appendix A _
6121 -NonProp-120302 December 2002
A-2 2
DETECTION I ace Appendix A 6121-NonProp-120302 December 2002
A-3 2.1 OUTSIDE SURFACE DEMONSTRATION II
] ace a,c,e Figure I Probability of Detection Performance For Passed And Passed Plus Failed Candidates for Appendix VIII Supplement 4, from the Outside Surface as a function of the flaw through wall extent (TWE). Both automated and manual techniques are included.
December 2002 Appendix A 6121-NonProp-120302
A-4 a,c,e Figure 2 POD for Inside Surface Examinations, Pass and Pass + Failed candidates, Passed and Pass Plus Failed Candidates are included.
F 7
a,c,e AflGIX A
December 2002 61ppendtx A 6121 -Non Prop-120302 December 2002
A-5 2.2 COMBINED ID AND OD DETECTION
[
]a c,e a,c,e Figure 3 Probability of Detection for Automated RPV Examinations Considering Both Inside and Outside Access. Passed and Passed Plus Failed Candidates are Shown.
Appendix A 6121 -NonProp-120302 December 2002
A-6 ac.c Figure 4 POD for Pass and Failed Candidates, Considering ID and OD Automated Demonstrations and Manual OD Demonstrations.
3 SIZING a-c,e Appendix A 6121 -NonProp-120302 December 2002
A-7 Figure 5 Histogram of Depth Successful Sizing Candidate Test Scores, Appendix VIII, Supplement 4. Examinations Were Performed Both From the Inside and Outside Surfaces.
[
a,c.e
]ace December 2002 Appendix A 6121 -NonProp-120302
lace Figure 6 Sizing Error Surface Model Figure 7 Plan View of Sizing Error Surface Model December 2002 Appendix A 0
6121 -NonProp-120302 A-8 a,c.e aic.e
A-9 4
ACCEPTABILITY EVALUATION I
]a,ce December 2002 Appendix A 6121-NonProp-120302 A
A-JO I
] ace Appendix A December 2002 Appendix A 6121-NonProp-120302 December 2002
A-I I A-ICe Figure 8 Probability of Correct Sizing for Passed Candidates, Appendix VIII Supplement 4.
Reporting Threshold A' = 0.15 inch.
[
]ace uecemoer LUUL Appendix A 6121 -NonProp-120302 D~ecember 2001
A-12 L
Figure 9 Probability of Correct Rejection/Reporting (PCR) for automated techniques, Considering Passed and Passed plus Failed Candidates, includes both inside and outside surface information. Reporting Criterion A' = 0.15 inch.
5
SUMMARY
[
lace 6
REFERENCES
]ace Appenazx A 6121 -NonProp-120302 December 2002 a,c.e
A-13
]a C.C Appendix A December 2002 6121 -NonProp-120302