ML031820670

To WCAP-15984-NP, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 & 2

ML031820670
Person / Time
Site:	Sequoyah
Issue date:	04/30/2003
From:	Strauch P, Swamy S ATI Consulting, Electric Power Research Institute, Westinghouse
To:	Office of Nuclear Reactor Regulation
References
WCAP-15984-NP, Rev 1
Download: ML031820670 (50)

Text

ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 WESTINGHOUSE ELECTRIC COMPANY WCAP-15984-NP, REVISION 1 (NON-PROPRIETARY VERSION)

E3-1

Westinghouse Non-Proprietary Class 3 WCAP-15984-NP Revision 1 April 2003 Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2 S

Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15984-NP Revision 1 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 April 2003 Reviewer:

c-4 P. L Strauch Structaral Mechanics Technology S.A.S amy, Mana 7

Strural Mechanics Technology Wesuigbouse Elecuic Company LLC P.O. Box 355 Pittsburgb, PA 15230.0355 0 2003 Westinghouse Eectric Company LLC AU Rights Resenred 6121 NPWoc.50103

i.i TABLE OF CONIENT I

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1-1 2

3 TECHNICAL APPROACH 2-1 FRACTURE ANALYSIS METHODS AND MATERIAL PROPERTIES.................................

1 3.1 Stress Intensity Factor Calculations 3-1 3.2 Fracture Toughness....

3-I 3.3 Irradiation Effects......................................................................................................

3-2 FLANGE INTEGRITY..............

4-1 ARE FLANGE REQUIREMENS NECESSARY?..........

.................... 1 SAFErY IMPCAT'IONS OF THE FLANGE REQIUIREMENT............................................6-1 REFERENCES.........

7-1 4

5 6

7 i

APENDIX A REACTOR PRESSURE VESSEL INSPECTION RELIABILITY........................... A-I APPENDIX B THERMAL AGING OF FERRlITC RPV STEELS AT REACTOR OPERATING TEMPERATURES B-1 APPENDIX C STRESS DISTRlBUTIONS IN THE CLOSURE HEAD REGION.............................. C-1 WCAP-159S4-NP 6l2H4NPioc-M10W3 Ap003 Ravision 1

1-1 1

INTRODUCTION 10 CFR Part 50, Appendix G contains requirements far pressur-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 deterrined using the methodology of ASME Section IM Appendix Q but the flange tenperature requireetus amr specified in IOCFR50 Appendix 0 This rule states that the metal temperatwe of the closure flange regions must exceed the material unirradiated RT,=r 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 hacture 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 stss, without being at steady state temperature. The margin of 120°F and the pressure limitation of 20 percent of bydrotest pressure were developed using the Kb 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 K1 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 1".

Figure 1-1 illustrates the problem created by the flange requirements for a typical PWR heatap curve. It is easy to sce that the heatup curve using Kk provides for a much higher allowable pressure through the entire range of temperatures. For this plant, however, the benefit is negated at temperatures below RrMT

+120°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 K, fracture toughness for flange considerations leads to the conclusion that the flange requirement can be eliminated for Sequoyah Units I and 2.

Revision I Revision I of this report was prepared to provide more details of the stress analysis performed, and to provide a detailed discussion of the effects of them aging on closure head materials.

WCAP-LS94-.N April2 0 6121-NPAOC4.00103 Revision I

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Figure 1-1 Musration of the Impact of the Flange Requirement for a Tpical PWR Plant WCAP-15984-NP 61214PTdoc-05003 April2003 Ravision 1

2-1 2

TECHNICALAPPROACH 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 1 and 2 is shown in Figure 2-1.

Stress analyses have been performed, and these stress results were used to perform fracture mechanics evaluations. Details of the finite element stress analysis are provided in Appendix C. The highest stress location in the closure head and vessel flange region is in the head, just above the bolting flange. This cormesponds with the location 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 bead 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 Ki. fracture toughness now known to be applicable, is there still a concern about the integrity of the closure head during boltup?

WCAP-15984-NP 4121-WAMcsol3 2Acril200 Revisio I

2-2 TOP HEAD DOME TORUS TO FLANGE WELD I-I-HEAD REGION VESSEL FLANGE TO UPPER SHELL WELD D

UPPER Sequoyah Units I and 2 A

88.1 B

6.89 C

29.72 D

170.88 NOTE: ALL DIMENSIONS ARE IN INCHES Figure 2-1 Geometry of the Upper eaBdtlnge Region of the Sequoyah Units 1 and 2 Reactor Vessels WCAP-15984-NP Apdl 2003 6121-NP4oc.050103 Revision 1 WCAP-15984-NP 6U1-WA0C-W0103 AprH 2003 Revisi= I

3-1 3

FRACTURE ANALYSIS METHODS AND MATERIAL PROPERTIES The fracture evaluation was carried out using the approach suggested by Section XI Appendix G (f.

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 wa1l 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 driving force or stress intensity factor (K,). 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:

o(x)=AD +A lx+A 2x2 +A 3x3 (3-1) where:

x

=

is the coordinate distance into the wall, in.

a

=

Stess 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 KI can be calculated anywhere along the crack front. The point of maximum crack depth is represented by 0 - 0, and this location was found to also be the point of maxinm KX for the cases considered here. The following expression is used for calculating Kg as a function of the angular location around the crack (). The units of K, are ksi/in.

0 3 E

[. 2 Gj (a/c, a/t, t/R, 4) Aj ai (3-2)

The boundazy correction factors Go. GI, G0, and 03 are obtained by the procedure outlined in referencc (2). The dimension 'a! is the crack depth, "c" is the crack half length, 't" is the wall thickness, "R" is the inside radius, and MQ" is the shape factor, approximated as I + 1464 (alc)lM4.

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:

K4, = 33.2 + 20.734 exp. [0.02 (r-RTmTr)]

(3-3)

Kk = 26.8 + 12A45 exp. [0.0145 (T-RTmwn)]

(3-4) where Kand K4.are in ksi/i.

WCAP-15984-NP il2003 6121-M7doc-00103 Rtvision 1

3-2 The upper shelf temperature regime reqwres utilization of a shelf toughness which is not specified in the ASME Code. A value of 200 ksiFinhas 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 XL The final key element in the determination of the fracture toughness is the value of RTIN, which is a material paramneter determined from Charpy V-notch and drop-weight tests.

The value of RTU for the closurc 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 5°F, and so this value was used for the illustradons to be discussed in Sections 4 and 5.

3.3 IRRADIATION EFFECTS Neutron irradiation has been shown to produce embriulement 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 nilductility transition temperature, RTNwr.

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.

alcle WCAP-15984-NP 61214NP~doc-050103 April 2003 Revision 1

4-1 4

FLANGE INTEGRITY The first step in evaluation of the closure head/flange region is to examine dhe stresses. The siresses which arm affected by te 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 the entire heatup and cooldown process arc summarized in Appendix C.

The boltup is the key condition to review here, in comparison with the heatup and cooldown operation, since the flange requirement applies tobolup conditions. No odter 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:

1.

The heatup and cooldown transient is strcturcd to ensure generous margins are maintained (SF = 2) for a large flaw in the irradiated betline 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 the governing time step of heatup and cooldown which is end of heatup. 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 compared with the most limiting time step of the entire heatuptcooldown transient.

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 bead flange, and the stress intensity factor, K, (or cack driving force) was calculated. The results are shown for the boltup condition in Figure 4-1, and for the. heatup and cooldown transient in Figure 4-2. For a semi-elliptic surface flaw with depth equal to 10 percent of the wall thickness postulated in the highest stress region of the head, the following values were determined for the stress intensity factor.

Boltup:

k = 20.0 ksi,/;

End of Heatup:

k = 54.64 ksi4s It will be useful to highlight the difference in the integrity story for the head region using the two values of fracture toughness. Mhe boltup temperature for a typical PWR is 60T, so if RTrur = 5SF the ASME reference toughness values are 1 = 54.4 ksi/i and 4 = 95.5 ksi-N. Using the KI. toughness (which was the basis for the original flange requirements) it can be seen that the toughness exceeds the applied stess intensity factor for boltup 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. For the beatup and cooldown transient, the coolant temperature at the governing time steps, near the end of heatup, is 547'.

The hacture toughness is therefore 200 ksi/i, so again the margin is very large.

WCAP-15984-NP 6121-lWAdoc-50103 April2003 Revisin 1

4-2 Using the Kk toughness, which has now been adopted by Section Xl for P-T Curves, it can be seen that them is also a significant margin between the fracture toughness and the applied stress intensity factor, for both the boltup and the heatup cooldown transient. 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 thee 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 Kk 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.

I WCAP-15984-NP 6121-NPAoc-050103 April 2003 Revsion I

4-3 Table 4-1 Stress Distributions for the Closure flange Region - Sequoyah Units 1 and 2 End of Heatup Distance Boltup Stress 344.2 minutes 2250 psi (ta (ks)i) 0 (ID)

-14.3S

-15.32 0.1

-10.77 0.2

-7.83

-3A2 0.3

-5.14 OA

-2.66 4.55 0O5 0.26 0.6 2.16 12.15 0.7 4.72 0.8 7.54 21.76 0.9 11.24 1.0 (OD) 19.70 38.77 WCAP-15934-NFP 6121-NP.doc-050103 April2003 Revision 1

44 W 4 LOOP REACTOR VESSEL CLOSURE HEAD/FLANGE WELD BOLT-UP OUTSIDE SURFACE STRESS MENSfTY FACTOR vs aft 25 a

C.06 0.12 0.18 0.24 CAS 0.36 0.42 0.48 SIn Figure 4-1 Crack Driving Force as a Function of Flaw Size: Outside Surface Flaw In the Closure Head to Flange Region Weld for Sequoyab Units 1 and 2 Boltup Condition (strs intensity factor units are kslf-)

WCAP-15984NlP 6121-NF.doc.050103 April 203 Revision 1

4-5 Sress Intensity Factors (K) for Clrcumferential Outside Surface Faw (Aspect Ration

• 6:1) 4-Loop top flange Reactor Vessel wlth Bdotup - Heatup and Cooldown Transient

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.E S0 A

0 0.1 02 0.3 04 0.5 0.6 0.7 0R 0

aft ratio (Ilaw depfh~all thIckness)

FIgure 4-2 Crack Driving Force as a Function of Flaw Size In the Closure Head to Flange Region Weld for an Outside Surface Flaw for Sequoyah Units 1 and 2: Heatup and Cooldown Transient (stress Intensity factor units are ksiFin)

WCAP-15984-NP 6121-NAIOC.-O103 April 2003 Revision I

5-1 5

ARE FLANGE REQUIREMENTS NECESSARY?

Using the Ks, curve can support th elimination of the flnge temperature requirement This can be illustrated by examining the stress intensity factor change for a postulated flaw as the vessel is pressurized afterboltup, progressing up to steady state operation.

The stresses at the region of interest are shown in Table 4-1, for the end of heatup, 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 sesses 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 Kk approach is provided in Table 5-i. This table provides calculated boltup requirements for all the designs, using a safety factor of 2, and a reference flaw depth of aft = 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 hea reference flaw, in light of current technology, and using the resulits of the Performance Demonstration Initiative.

Basis for the Reference Flaw Size. Regulatory Guide 1.150 stimulated improvement in examinations of the clad to base-mctal 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.

1"9 WCAP-15994-NP 6121-M&oc0350103 A

2003 Revision 1

5-2 II Ice WCAP-15984-NP 6121NADoc.050103 April 2003 Revision 1

5-3 Table 5-1 Comparison of Various Plant Designs Boltup Requirements T - RTmTD (OF T - RTN= (°li)

K Kwith UsingKi.

using KU Plant (aIt =.1)

SF=2 (aft =.10)

(an.1O)

CE 30.0 60.0 13 68 B&W 39.4 79.8 41 100 W4Loop 19.7 39.4 0

1 W3Loop 194 38.8 0

0 GE (CBI 251) 38.7 77.4 38 97 GE(B&W251'I) 48.0 96.0 56 118 GE (CE 21 'B) 25.1 50.2 0

43

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.1

5-4 alce Figure 5-1 Probability ef Correct RejectionuReporting (PCR) Considering Passed plus Failed Candidates, Appendix VIU rom ithe Outside Surface. Reporting Criterion A = 0.15 hich WCAP-15984-NP 6121-NPdol-O. 3 April2003 Revision I

5-5 Figure 5-2 Probability of Correct ReJectiontReporting (PCR) Considering Only Passed Candidates, Appendix VIII from the Outside Surface. Reporting Criterion A' = 0.15 nh WCAP-15984-NP 6121*ZPAIOo050103 April 2003 Revision I

6-1 6

SAFETY IMLICATIONS OF THE FLANGE REQUIREMENT There are imporant safety implications which are associated with the flange requirement, as illustrated by FiMre 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 Emit curve may be superceded by the flange requirement for temperatures below RThvr + 120F. 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 hearup 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 sezpoints 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 1 would mean that the PORV curve could become level at 6041587 psig, which are the leading/trailing setpoints to protect the PORKV downstream piping, through the temperature range of the 350F down to boltup at 60ME 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 casing 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 Z for example, is shown in Figures 6-4 and 6-5 [131.

WCAP-15984NP 6121.NPAIOC-O10I3 A2O03 Revision 1

6-2 aI. Heabtp Curve Ins~nuw~

Uncertainty 621 I

u Pump S"I Arh-Um325 PO RTM9 +120 Tumperutr FIgure 6-1 Ilustration of the Flange Requirement and Its Effect on the Operating Window for a Typical} eatup Curve WL.Ar-zlyAS-LNr 6121-NPAoc-050103 AprZ zWi Revision I

6-3 umMIIGmATERIAL On'ERMEDLDJ7 6HE.L FORGING P-5933 (usmx.cws*Awd)

LUMMTNG ART VALUES AT 12 EFPY:

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Figure 6&2 mustratl of wt Actul Opera1t EWindow for Heatup of Byron Unit 1, a Low Copper PlwAt at 12 EFPY WCAP-15984-NP 6121-NPA=ocO103 Apffl 2003 Revision I

64 UMrTNG MATERIAL 1NERMEDIATE SHELL. FORGNG 6P4W3 4manw.

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2500 2250 2000 1750 750 500 25D 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 6-4 Illustration of the Flange Notch for Sequoyah Unit 2, EIeatup Curve, without Instrument Uncertainties [13]

WCAP-15984-NP 6121-.doc450103 Api 2003 Revision 1

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Flgurc 6-5 Illustration of the Flange Notch for Sequoyah Unit 2, Cooldown Carves, without Instrument Uncertainties [131 WCAP-15984-NP 612-CW.doc-MSO!3 AprH 2MU RevWsOn I

7-1 7

REFERENCES

1.

ASME Boiler and Pressure Vessel Code,Section X, Appendix G. 1996 Addenda, ASME, New York.

2.

Raju, L 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. U., ed., "Flaw Evaluation Procedures: ASME Section XI' Electic Power Research Institute Report EPRI-NP-719 SR, August 1978.

4.

Mitchl, 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 XL Working Group on Operating Plant Criteria. Hollywood, FL, September 10, 2002.

5.

Nanstad, R. K., ct &d.,

Prelimnwy Review of Data Regarding Chemical Composition and Thermal Embritlement of Reactor Vessel Steels, ORNLINRC/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 Submered Arc Weld Metals:' Effects of Radiation on Materials: 16th International Symposium, ASTM S7P 1175, Philadelphia, PA, 1993.

7.

Kirk, M., "Revision of AT30 Embritacmcnt Trend Cwrvcs," presented at the EPRI MRPINRC PTS Re-Evaluation meeting in Rockville, MD, August 30,2000.

8.

Charpy Embfrirement Correlations - Staha of Combined Mechanistic and Statistical Basesfor U.S. RPV Steels (MRP-45); PWR Materials Reliability Program (PWRMRP), EPRL Palo Alto, CA: 2001, 1000705.

9.

ASTM E 900-02, "Standard Guide for Predicting Radiation-Induced Transition Temperature Shift for Reactor Vessel Materials, E706 ([I)," Annual Book of ASTM Standards, Vol. 12.02.

10.

Langer. R., ct al., "A Survey of Results on Aging Experiments of Pressure Vessel Materials:'

presentation at the ATHENA Wodcshop, Madrid, September2002.

11.

Randall, N., Abstract of Comments and Staff Response to Proposed Revision to 10 CFR Pan 50, Appendioes U and HL Published for Commcen in the Federal Register. November 14, 1980.

12.

WCAP-15293, 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

=TLR Support Documentation," J. IL Ledger, et al., April 2001.

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A-1 APPENDIX A REACTOR PRESSURE VESSEL INSPECTION RELIABILITY*

F. L Becker EPRI Charlotte NC AB3STRACT I

la dPgesenhd at theJorEC-IAEATedmnca1 Meeft~ an Improvemmts in Iaswe*eInspccuon~ ffciveness. PC=a. Mte Nedterhnds. November2002 to be publisbed.

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DETECTION i

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A-3 2.1 OUTSIDE SURFACE DEMONSTRATION i

Figure 1 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 (1WE).

Both automated and manual tedhniques are Included.

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&,c~e Figure 2 POD for Enside Surface Examinations, Pass and Pass + Failed Candidates, Passed and Pass Plus Failed Candidates are Included.

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A-5 2.2 COMBINED ID AND OD DETECTION I

a,ce Figure 3 Probability of Detection for Automated RPV Fxaminatons Considering Both Inside and Outside Access. Passed and Passed Plus Failed Candidates are shownL WCAP.15984-NP 611-NPdoc450103 April 2003 Revision I

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Figure 4 POD for Pass and Failed Candidlates, Considering ED and OD Automated Demonstrations and Manual OD Demonstrations.

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A-7 a,ce Figure S Histogram of Depth Successful Sizing Candidate Test Scares, Appendix VM, Supplement 4. Examinations Were Performed Both From the Inside and Outside.

Surfaces.

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1 rat mCe Figure 6 Sizing Error Surface Model Ftgre 7 Plan View of Sizing Error Surface Model xc,'

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ACCEPTABILITY EVALUATION I~~~~~~~~~~~~~~~~~

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Figure 8 Probability of Correct Sizing for Passed Candidates, Appendix Vm Supplement 4.

Reporting ThresholdA' = 0.15 Inch.

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A-12 ac,e Figure 9 Probability of Correct RejectionAReporting (PCR) for automated techniques, Considering Passed and Passed plus Failed Candidates, includes both Inside and outside surface information. Reporting Criterion A' - 0.15 inch.

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SUMMARY

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6 REFERENCES

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B-i APPENDIX B THERMAL AGING OF FERRITIC RPV STEELS AT REACTOR OPERATING TEMPERATURES

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c-1 APPENDIX C STRESS DIMTRBUTIONS [N THE CLOSURE HEAD REGION I

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