ML20101S940
| ML20101S940 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/31/1991 |
| From: | Branlund B, Caine T, Ranganath S GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20101S935 | List: |
| References | |
| DRF-B13-01445-1, DRF-B13-1445-1, SASR-88-24, SASR-88-24-R01, SASR-88-24-R1, NUDOCS 9207200158 | |
| Download: ML20101S940 (268) | |
Text
{{#Wiki_filter:SASR 88-24 , DRF-B13 01445-1 Rev. 1 C:'ss I betsmber 1991 PEACH BOTTOM ATOMIC POWER STATION UNIT 2 VESSEL SURVEILLANCE MATERIALS TESTING AND FRACTURE TOUCHNESS ANALYSIS Prepared byb b,_L , B. J. Branlund, Engineer Structural Analysis Services ar..i i Materials Monitoring l'
~
Verified by: .b. T. A. Caine, Senior Engineer Structural Analysis Services and Materials Monitoring , 9 Approved by: 6 \ S. Ranganath, Manager Structural Analysis Services and Materials Monitoring 9207200158 920710 PDR ADOCK 05000277- __
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IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the use of the Philadelphia Electric Company. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared. The only undertakings of the General Electric Compsny respecting information in this document are contained in the contract governing Philadelphia Electric Company Purchase Order No. PB999125-N and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately o'med rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information. 11
' CONTENTS EAxs ABSTRACT viii ACKNOWLEDCMENTS ix
- 1. INTRODUCTION 1-1
- 2.
SUMMARY
AND CONCLUSIONS 2-1 2.1 Summcry of Results 2-1 2.2 Conclusions 2-5
- 3. SURVEILLANCE PROGRAM EACKGROUND 3-1 3.1 Capsule Recovery 31 3.2 RPV Materials and Fabrication Background 3-1 3.2.1 Fabrication History 3-1 3.2.2 Material Properties of RPV at Fabrication 3-2 3.2.3 Specimen Chemical Composition 3-3 3.2.4 Initial Reference Temperature 3-3 3.3 Specimen Description 36 3.3.1 Charpy Specimens 3-7 ,
3.3.2 Tensile Specimens 3-8
- 4. PEAK RPV FLUENCE EVALUATION 41 4.1 Flux Wire Analysis 4-1 4.1.1 Procedure 41 4.1.2 Results 4-2
4.2 Determinat
of Lead Factors 4-3 4.2.1 ' Proc are 4-3 4.2.2 Results 4-4 4.3 Estimate of 32 EFPY Fluence 45
- 5. CHARPY V-NOTCH IMPACT TESTING 51
- 5.1 Impact Test Procedure 5-1 5.2 Impact Test Results 5-2 l
l 5.3 Irradiated Versus Unirradiated Charpy V-Notch Prog-rties 5-3 5.4 Comparison to Predicted Shift 5-4 l
- 6. TENSILE TESTING 6-1 6.1 Procedure 6 1-6.2 Results 6-2 6.3 Irradiated Versus Unirradiated Tensile Properties 6-3
- 7. DEVELOPMENT OF OPERATING LIMITS CURVES 7-1 7.1 Background 7-1
! 7.2 Non-Beltline Regions 7-1 7.3 Core Beltline-Region 7-2 l. 7.4 Closure Flange Region _ 7-2 i 7.5 Core Critical Operation Requirements of 10CFR50, Appendix G 7-3 I 111
CONTENTS EbLt 7,6 Evaluation of Radiation Effects 74 7.6.1 Measured Versus Predicted Surveillance Shift 7-5 7.6.2 ART Versus EFPY 7-5 7,6.3 Fracture Toughness Conditions at 32 EFPY 76 7.7 Operating Limits curves Valid to 32 EFPY 7-6 7.8 Reactor Operation Versus Operating Limits 77
- 8. REFERENCES 81 APPENDICES A. DETERMINATION OF ELECTROSLAG VELD PROPERTIES A-1 B. IRRADIATED CHARPY V-NOTCH FRACTURE SURFACE PHOTOCRAPHS B-1 l C. UNIRRADIATED CHARPY V-NOTCH FRACTURE SURFACE PHOTOCRAPHS C-1 l iv
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . - _ _ ]
- - ._- .. .-..~ ~... --.- - - . _
-TABLES Table Title Pa r.e 3-1 Chemical Composition of RPV Beltline Materials 39 32 Results of Fabrication Test Program for Selected RPV 3 10 Locations 33 Plasma Emission Spectrometry Chemical Analysis of RPV 3-11 Surveillance Plate and Weld Materials 34 Identification of Charpy and Tensile Specimens Removed 3-12 from Surveillance Capsule 4-1 Summary of Daily Paver History 4-6 4-2 Surveillance Capsule Location Flux and Fluence for 4-7 Irradiation 2/15/74 to 3/13/87 5-1 Qualification Test Results Using U.S. Army Watertown 5-5 Specimens (Tested in March 1988) 5-2 Charpy V-Notch Impact Test Results for Irradiated RPV 56 Materials 5-3 San Jose Tinius-Olson Charpy Machine Qnalification Test 5-7 Results using NIST Standard Reference Sps ' mens (Tested 5/14/90) 5-4 Charpy V-Notch Impact Test Results for Unirradiated RPV 5-8 Materials 5-5 Significant Results of Irradiated and Unirradiated Charpy 5-9 l l V Notch Data 6-1 Tensile Test Results for Irradiated RPV Materials 6-4 l-6-2 Comparison of Unirradiated and Irradiated 6-5 Tensile Properties at Room Temperature 7-1 Estimate of Upper Shelf Energy for Beltline Materials 7-8 l
V
.. . . _ . , . . . . , . . - . , . . - . . . . , . . . . - . . ~ . , . _ _ . . . _ . _ . - . . , . . , - . . . _ . . . _ . - . . ._..,__....___,..._-_.m..-,,_
. . - . . - . . . . -.. - . . . _ . . - .-e . -- . - ~ - - .- - . - . . .. . . . - ~ . . . . - -.--. . -
r ILLUSTRATIONS Firure Title Eagn 26 2-1 Minimum Temperature for Pressure Tests Such as Required by Section XI 2-2 Minimum Temperature fot Mechanical Heatup or Cooldown- 27 Following Nuclear Shutdown 23 Minimum Temperature for Core Operation (Criticality) 2-8 31 Surveillance Capsule Recovered from Peach Bottom 2 Reactor 3 13 32 Schematic of the RPV Showing Arrangement of Vessel 3-14 Plates and Welds 3-3 Fabrication Method for Base Metal Charpy Specimens 3 15 3-4 Fabrication Method for Veld Metal Charpy Specimens 3-16 35 F4'rication Method for MAZ Charpy Specimens 3 17 36 Fabrication Method for Base Metal Tensile Specimens. 3 18 3-7 Fabrication Method for Veld Metal Tensile Specimens 3 19 3-8 Fabrication Method for HAZ Tensile Specimens 3-20 4-1 Schematic of Model for Two-Dimensional Flux Distribution 4-8 Analysis 4-2 Relative Fast Neutron Flux Variation With Angular 4-9' Position at the Vessel 5-1 Trradiated Base Metal Inpact Energy 5-10 5-2 Irradiated Weld Metal Impact Energy 5 11 5-3 Irradiated HAZ Impact Energy 5-12 5-4 Irradiated Base Metal Lateral Expansion 5-13 5-5 Irradiated Weld Metal Lateral Expansion 5 5-6 Irradiated HAZ Lateral Expansion 5-11 5-7 Comparison of Unirradiated and Irradiated Base Metal 5-16 Impact Energy vi w ~ v ne a ew-d-, ' v er t ~ n um-s vm ' + c M - ' %ms '+m en += ke- w ns re e,o r s.o.~ eewe n,wer m e nmesm w w m m wm re w =, ow s- s ewmm v ow w* s e. +-sa m se wv a +.s-,u-me ne
ILLUSTRATIONS Firure Title Eagt 5-8 Comparison of Unirradiated and Irradiate. Weld Metal 5-17 Impact Energy 5-9 Comparison of Unirradiated and Irradiated Base Metal 5-18 l Lateral Expansion ) 5-10 Comparison of Unirradiated and Irradiated Weld Metal 5-19 Ieteral Expansion 6-1 Typical Engineering-Stress versu Percent Strain for 6-6 Irradiated RPV Materials 6-2 Strength versus Test Temperature for Irradiated Base, 6-/ Weld, and HAZ Tensile Specimens 6-3 Ductility versus Test Temperature for Irradiated Base, 6-8 Weld, and HAZ Tensile Specimens 6-4 Fracture Location, Necking Behavior, and Fracture 69 Appearance for Irradiated Base Metal Tensile Specimens 6-5 Fracture Location, Necking Behavior, and Fracture 6 10 Appearance for Irradiated Weld Metal Tensile Specimens 6-6 Fracture Location, Necking Behavior, and.7racture 6-11 Appearance for Irradiated HAZ hetal Tensile Specimens 7-1 Composents of Operating Limits Curve for Pressure Tests 7-9 (Curve A) 7-2 Components of Operating Limits Curve for Non-Nuclear. 7-10 Heatup/Cooldown (Curve B) 7-3 Components of Operating Limits Curve ;sr Core Critical 7-11 Operation (Curve C) for Cooper 7-4 Adjusted Reference Temperatures of Limiting Plate and 7-12 Weld 7-5 Minimum Tempsrature for Pressure Tests Such as- 7-13 Required by Section XI 7-6 Minimum Temperature for Mechanical Heatup or Cooldown 7 14 Following Nuclear Shutdown 7-7 Minimum Temperature for Core Operation (Criticality) 7-15 vii l ._ _ . . , . . . . . . _ , . . _ . _ . , . _ . , _ . _ . , , _ , _ , . . _ , _ -
ABSTRACT A surveillance capsule was removed from the Peach Bottom Atomic Power Station, Unit i reactor at the end of Fuel Cycle 7. The capsule contained flux wires for neutron fluence measurement and Charpy and tensile test specimens for a material property evaluation. A combination of flux wire testing and computer analysis was used to establish the vessel peak flux location and magnitude. Charpy V Notch impact testing and uniaxial tensile testing were performed to establish the material properties of the irradiated vessel beltline. The irradiation effects were projected, based on Regulatory Guide 1.99, Revision 2, to conditions for 32 EFPY of operation. The 32 EFPY conditions are predicted to be less severe than the limits requiring vessel thermal annealing. Pressure-temperature operating limits curves valid to 32 EFPY vere developed to July 1983 requirements of 10CFR50 Appendix G, accounting for irradiation shift per Regulatory Guide 1.99, Revision 2. The non-beltline limits are more severe than the beltline limits, even including predicted 32 EFPY shift. A revision was made to this report to include archive unirradiated l plate and weld specimens. The revisions will hereafter be called l
" Revision 1". Revision 1 additions to this report will be indicated by j bold print and a "l" in the right column. l In addition to surveillance capsule testing, Revision 1 includes l unirradiated plate and weld Charpy Specimens that were tested to establish l baseline curves. The irradiated Charpy data for the Unit 2 plate and veld l specimens were compared to the unirradiated data to determine the shif t in l Charpy curves due to irradiation. The measured shift for the plate l material is considerably less than both the mean and uppe.r bound prediction l of 1.99. The measured shift for the veld material nearly matches the mean l prediction and is considerably less than the upper bound prediction of l 1.99. l viii
_ . _ _ _ . _ ~ _ . _ _ , _ - . _ . _ . . _ _ _ _ _ __ .. _ _ ___ _-_._._. _ _ _ . . . _ . . . _ . . _ _ . . . . _ _ _ _ . . . ACKNOWLEDCMENTS
- Flus wire testing was performed by G. C. Martin. L. S. Burns provided 1
the evaluation of-flux distribution. The Charpy V Notch impact testing was done by G. P. Vozadio~and G. E. Dunning. S. B. Wisner and C. H. Hendersen i performed the tensile specimen testing. C. R. Judd performed the chemical 1 composition tes, ting. Y l l l l- ix l ___ ________ ._ ___ _ __ .._...-_.2 . . - _ . _ _ . _ _ . . , . - . . . . . . . - _ , . . . . . . . . , _ . , _ . , . , - , . . . . . _ . . . _ _ . , , . . - , . - , . . , , , , , , , , . , , . , , . _ , . , , , . . .
- 1. INTRODUCTION Part of the effort to assure reactor vessel integrity involves evaluacion of the fracture toughness of the vessel ferritic materials. The key values which characterize a material's fracture toughness are the reference temperature of nil-ductility transition (RTNDT) and the upper shelf energy (USE). These are defined in 10CFR50 Appendix G (Reference 1) and in Appendix G of the ASKE Boiler and Pressure Vessel Code, Section III (Reference 2). Thess documents contain requirements used to establish the pressure-temperature operating limits which must be met to avoid brittle fracture.
Appendix H of 10CFR50 (Reference 3) and ASTM E185 (Reference 4) establish the methods to be used for surveillance of the reactor vessel materials. In January,1988 one of the Peach Bottom 2 vessel surveillance specimen capsules required by Reference 3 was sent to General Electric for testing after exposure to seven fuel cycles of irradiation, or 7,53 effective full power years (EFPY) of operation. The surveillance capsule contained flux virea fos . tron flux monitoring and Charpy V Notch impact test specimens and mic, .1 tensile test specimens fabricated from the vessel materials nearest the core (beltline). The impact and tensile specimens were tested to establish material properties for the irradiated vessel materials. The results of surveillance specimen testing are presented in this report. The irradiated material properties are compared to available > unirradiated prope : ties from testing of archive -specimens. Predictions of l the RTNDT and USE at 32 EFPY are made for comparison with allowable values in Reference 1. Predictions of 32 EFPY properties were made based on Regulatory Guide 1.99, Revision 2 (Reference 5). Operating limits curves for the Peach Bottom 2 reactor vessel are-presented in this report. The. curves account for current requirements of References 1 and 2. Geometric discontinuities and highly stressed regions, 1-1
. J
such as the feedwater nozzles and the closure flanges, are evaluated separately-from the core beltline region. The operating limits developed consider- the anost limiting conditions of the discontinuity regions and the beltline region (including irradiation) to bound all operating conditions. The operating limits developed include irradiation shift in the beltline materials, based on the Reference 5 methods. i
?
F
- i. 1-2 i
,,e.,.......,=,,-...,,-e.-,,ws,ww.--,,mawm.,wyw..,,.m,...#.,,mn,,,w.,,..ww..y,,wrc.py--r,, e.e.-,,s.ye-c.c % ,, g c o ,. , w .w %, , , , - . --
- 2.
SUMMARY
AND Co..,.,LUSIONS 2.1
SUMMARY
OF PISULTS j Surveillance' capsule 2 was removed from the Peach Bottom 2 reactor at the end of Fuel Cycle 7 and shipped to the General Electric Vallecitos Nuclear Center. The flux wires, Charpy V-Notch and tensile test specimens removed from the capsule were tested according to ASTM-E185-82 (Reference 4). Revised operating limits curves were developed using the test results along with 10CFR50 Appendix G (Reference 1) and Appendix G of the ASME Code (Reference 2). The methods and-results of the fracture toughner.s evaluation are presented in this report as follows:
- a. Section 3: Surveillance Program Background
- b. Section 4: Peak Vessel Fluence Evaluation
- c. Section 5: Charpy V-Notch Impact Testing
- d. Section 6: Tensile Testin6
- e. Section 7: Operating Limits Curve Development Data used to determine the RTNDT of the electroslag velds in the beltline are documented in Appendix A. Photographs of fractured Charpy specimens are in Appendices B and C. The significant results of the evaluation are l summarized as follows: -
- a. Capsule 2 was removed from the 120' azimuth position of the reactor. The capsule contained 6 flux vires: 2 each of pure copper (Cu), irm (Fe), and nickel (N1). There were 24 Charpy .
V-Notch specimens: 8 each=of plate material, weld material and heat affected zone (HAZ) material. The 8 tensile specimens-- removed consisted of 3 plate, 3 weld and 2 HAZ metal specimens. x 24
i s. l All specimen materials were positively identified as being from the vessel beltline. 1
- b. The chemical compositions of the beltline materials were identified through a. combination of literature research and testing. The copper (Cu) and nickel (Ni) contents were determined for all heats of plate naterial. The values.for the limiting beltline plate are 0.12% Cu and 0.57% Ni, The values for the limiting beltline weld are 0.21% Cu and 0.21% Ni. i l
- c. Resulcs from the fabrication program materials certification testing were located and adjusted to be equivalent to test results done to current standards. The initial RTNDT values for locations of interest in the vessel were determined. They are
-6*F for the limiting beltline plate, -45'F for the limiting beltline veld, 10*F for the closure flange region and 52*F for the bottom head torus, which is the component with the hia; hest RTNDT in the non beltline regions.
- d. The flux wires were tetted to determine - the neutron flux at the surveillance capsule location. The fast flux (>1.0 MeV) measured was 7.5x108 n/cm2 -sec. Based on the flux wire data, the surveillance specimens had received a best estimate fluence'of 1.8x1017 n/cm2 ,e-,emoy,1,
- e. The vessel peak inuer surface and 1/4 T lead factors were established using an analysis that combines two dimensional and one-dimensional finite element computer analysis. The flux peak occurs at an azimuthal location 25.5* past the vessel quadrant references. The lead factors for the surveillance capsule are 0.95 to the peak vessel surface and 2. 38 to the peak 1/4 T depth location.
2 _ . _ . - _ . _ ._ . _ _ ,-.._... ... _ , . _ . . . _ _ . _ . . . . . _ . . . . . . . . _ . - . , _ , . . , . . . . - . . . _ . , _ . , . , _ _ . _ . , . _ ,
. - ~ _ . . . ... . . - .- - . _ . _ . .-- .. -_ .. .
l r
- f. The maximum accumulated nautron fluence at 32 EFPY was determined I at the peak 1/4 T location. The r.aximum 1/4 T vessel 32 EFPY I fluence is 5.5x10 17 n/cm 2 (best estimate) and 6.9x1017 r/cm2 (upper bound).
- g. The surveillance Charpy V-Notch specimens were impact tested at temperatures selected to define the transition of the fracture
;oughness curves of the plate, weld, and HAZ materials.
Measurements were taken of absorbe, energy, lateral expansion and percentage shear. Fracture surf :e photographs of each specimen are presented in Appendix B. From absorbed energy and lateral expansion results for the plate and weld materials the following j values are extracted: index temperatures for 30 ft-lb 50 ft lb, , and 35-mil lateral expansion (MLE) values and USE. Similar test l results were obtained for archive plate and weld Charpy Specimens l provided bv FECO. The unirradiated specimen frccture surface l photographs are presented in Appendix C. l
- h. The irradiated plate and weld impact energy curves are compared to the corresponding unirradiated data from fabrication records l-to establish the 30 ft-lb, 50 ft-lb, and 35 NLE index temperature l irradiation shifts and decreases in USE, 'The plate material l shows an estimated -5'F shift and 9% increase in USE. The weld l material estimated shift was 17'F vith an increase in USE of 3% -l (the negative shifts and the increases in USE are probably due to l the limited data for the statistical fit and the variable nature l of Charpy test resultn. l
- i. The irradiated tensile specimens were tested at room temperature (70*F), reactor operating temperature (550*F), and estimated onset to upper shelf temperature (130*F for plate material and 180'F for veld material). The'results tabulated for--each l
specimen include yield and ultimate tensile strength, uniform and total olongation, and reduction of area. 2-3
- - - - - . - . .-,,y m-,.rae-+= or - c< -
m-+r , = w-+c- -"e- **-wwww y i e vr- -%
- w -e --=~e- rev'-* w v-v-- w&
; 1 t \
I i 1 i j. The irradiated plate and weld tensile test results are compared i i j to unirradiated data from tha vessel fabrication test program. The materials show increased strengtn and generally decreased ductility, as expected for irradiation etabrittlement, j l k. As a part of the construction of the updated operating limits curves, the irradiation shift in RTNDT was based on predictions l calculated with Regulatory Guide 1.99, Revision 2 (Reference 5). For information purposes, the measured surveillance shifts for base and weld material were compared to the shift predicted by ] Reference 5. The surveillance test shirt of -5'T in plate base l ! material RTNDT for a fluence of 1.8x1017 n/cm2 (best i 6. stimate) is considerably lower than the predicted upper limit on j shift of 21.0'F. The surveillance test shift of 17'F in the weld l material RTNDT is also considerably lower than the predicted l upper limit on shift of 27.2*F. l l l
- 1. The USE at 32 EFPY is predicted using the methods in Reference 5.
The longitudinal veld USE is predicted to be 80 ft lb at 32 EFPY.
! The surveillance plate USE is 132 ft-lb longitudinal at 32 EFPY.
Branch Technical Position MTEB 5 2 (Reference 6) recommends 65% of the longitudinal USE as an estimate of transverse USE, so the 32 EFPY plate transverse USE would be 86 ft lb. Data needed to i l preoict USE for the other beltline plates and for the l i circumferential weld are not available. l
- m. Operating limits curves were constructed for three reactor conditions: hydrostatic pressure test.s, non nuclear hestup and I
cooldown, and core critica 1 oper . tion. The curves are valid up to 32 EFPY of operation. The limiting regions of the vessel affecting the curves' sha pe s are typically the core beltline (shifted to acco.rt for irradiation), the feedwater nozzle and CRD penetration discontinuities, and the ilosure flange region. The bolt preload and nfnimum permissible operating temperatures 24
on the curves of 70*T provide some additional margin in the closure flange region where a detectable flaw size of 0.24 inch is used instead of 1/4 T. The predicted irradiation shift for the beltline is low enough that the beltline is not predicted to be limiting through 32 EFPY of operation. The operating limits curves for Peach Bottom 2 are shown in Figures 2-1 through 2 3.
2.2 CONCLUSION
S The requirements of Reference 1 deal basically with vessel design life conditions and with limits of operation designed to prevent brittle fracture. Based on the evaluation of sv.rveillance testing, the following conclusions are dravn:
- a. The adjusted reference temperature for the plate material of
$1'F is the limiting beltline value at 32 EFPY. This is below the Reference 1 allowable limit of 200'F, above which annealing is required.
- b. The 32 EFPY values of USE for one beltline plate and the longitudinal weld materials aru 86 f t lb transverse and 80 f t lb, respectively. These are above the Reference 1 allowable of 50 ft lb, below which annealing is required. Data are not available for the other beltline plates-and the circumferential weld, but
. based on data for similar materials, and considering the low fluence, it is expected that all beltline materials will have USE above 50 f t lb through 32 EFPY.
- c. Based on (a) and (b) above, provisions for complete volumetric examination or annealing of the reactor vessel before completing 32 EFPY of operation need not ba considered.
- d. Examination of the normal and upset operating conditions expected for the reactor shows that the worst pressure temperature 2-5
____________________a
conditions oxpected from unplanned temperature transients are acceptable relative to the limits in Figures 2 1 throu6h 2 3. Therefore, the only operating conditions for which the operating limits aro a concern are those involving operator interaction, such as hydrotest and initiation of core criticality. 26
, i i ,
VALID TO 32 EFPY i I i k I_ i , , , , ,1
, , i i i t i i
1 e i , I e i i i
. 1 i
1400 g- l , I. i: ! ; EELTLINE 1/4 T FLAW l WITH 32 EFPY SH'FT l PER 1.99, REV. ?
' 15 LESS LIMITING
.1 l' l THAN NON BELTLINE 1200 ~
l , l - 1 I __ < - . . _ , I I - _ - _ _ _ i . . . _ f . _;
! -- J SAFE 1000 / CPERATING ~
I _ REGION ~.
- -I__-- .._.
/
[ I I &_ l . > Sor I
/ \ NON-BELTLINE FW N0ZZLE LIMITS,
- i
~
/ 1/4 i FLAW, RTNOT = 52 F 0
7 1 n ' ii e- -
,1 . .. . - -
-_=
_ - _. y F:0
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e. g ._
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~ ~ ~ ~ ~ ~
312 PSIG
=-. -
200 l . . l
,-g*
BOLT FRELOAD TEMPERATURE = 70 F 0 i- l i
~
~ ~ ~ -
0 ~~
.g
.j FLANGE REGION, RTNOT = 10 F ~~
y l
-~~ 1 ._.] . , __Q
.L . : _Z. .~..
i* 0 600 100 200 300 400 500 O
' RPV Metal Temperature (OF)
Figure 2-1. Minimum Temperature for Pressure Tests Such as Required by Section XI 2-7
! VALID TO 32 EFPY
[ t I e ] 8 i j a , 4 i e i l . i 4 ,. 4 i 1 i i e 8 e { i . i l l i . . . . 1400 , l . [ i _, BELTLINE 1/4 i FLAW l
' ,' l _, WITH 32 EFPY $HIFT
, 4 i PER 1. 9. REV. 2
. i , . . >
IS LES LI.w1 TING
. e l ' i THAN NON. BELTLINE
, 6 . , .
i ' i __ .1 i . 1200 . i l . i
' " ~
__ - SAFE toog OPERATING REGION 6 , , c C- . c , i H . -
? B00 l .. ...... ,
= . _ - _ _ . _ _..
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= . - - - - -
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, =
I - e . U 600 1 ._ __ _ - - - . . - . c i _ . 6 I __ _ _ , _ , , , ., . I c I*'s. f i x __., c -_. 6 I .N .
'- O NON. BELTLINE ---
/ -- FW N0ZZLE LIMITS, ---
0 400 , - 1/4 i FLAW, RTNDT = $2 F , I s e n t , i i
/ i - - _ -
1 . 1 - , 200 ..'._,. . . . . . . . .
/ ! ,
BOLT PRELOAD
' ' ' 0 ' 8 TEMPERATURE a 70 F 0
/ -[*--- FLANGE REGION, RTNOT = 10 F ,
{ T.Z l 7; , _ _ _ t 300 400 500 600 0 100 200 RPV Metal Temperature (OF) Figure 2 2. Minimum Temperature for Mechanical Heacup or Cooldown Following Nuclear Shutdown 2-8
.. ___. - -. _ . . - - - - _-. . _ - _ _ _ = _ _ - = - _ - . - . . . _ - . _ . - . _ _ = - . _ _ - - .
l l I l l l i . 4
' j i VALID TO 32 EFPY l . l l
. i H 5 +
8 'f f '1 1 if f j i
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_ m . + , 4 , , . _!____ . I 3'_ -i , , 4 I- , l , !
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. EELTLINE 1/4 T FL1'J l
! ' , . . WITH J2 EFPY $HIfi
. PER 1.99, REV, 2 IS LE$$ LIMITING THAN NCN. BELTLINE
. i_ '
1200 , , i i , , SAFE . OPERATING 1000 - REGION - 5 - _e., c. C h . E 800 . _ _ . _..._ m _ _ - - c _ j. .-- ___ _ _ _ . _ e -,I___
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- v. -
c-600 ,' _ u_ '
- $ I%
o I . . .%
$ .)- NON. BELTLINE FW N0ZILE LIMITS, 0 4g
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1/4 T FLAW, RTNOT = $2 F e I
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.1 200 j
_- .. .j l ,
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i BOLT PRELOAD 0 ;---- I
-.._.2 TEMPERATURE = 70 F j -___ f1 - - - -
FLANGE REGION, RTNOT = 10 f 0 . f r, p; ,
, . - -- g 0 600 0 100 200 300 400 500 RPV Metal Temperature (OF) t Figure 2-3. Minimum Temperature for core operation (criticality) 2-9
- 3. SURVEIU ANCE PROGRAM BACKCROUND 3.1 CAPSULE RECOVERY The Peach Bottom 2 reactor was shut down in March, 1987 for refueling and maintenance. The accumulated thermal power output was 9.06x10 6 MVd or 7.53 EFPY. The reactor pressure vessel (RPV) originally contained three surveillance capsules, at 30', 120' and 300' azimuths at the core midplane.
The specimen capsules are held against the RPV inside surface by a spring loaded specimen holder. Each capsule receives equal irradiation because of core symmetry. During the outage , Capsule 2 at 120' was removed. The capsule was cut from the holder assembly and shipped by a 200 Series cask to the General Electric Vallecitos Nuclear Center in Pleasanton, California, Upon arrival at Vallecitos, the capsule was examined for identifier 'on. The reactor code of 2* and the code for Capsule 2, shown on the CE drawing (Reference 7) were confirmed on the capsule, as shown in Figure 3-1. The capsule contained two Charpy specimen packets and four tensile specimen tubes. Each Charpy packet contained 12 Charpy specimens and 3 flux vires. The four tensile specimen tubes contained eight specimens. The specimen gage sections were covered by aluminum sleeves, and during removal of the sleeves, the gage section of one tensile specimen was nicked. The specimen was tested in spite of this, as discussed in Section 6. All other specimens were removed without incident. 3.2 RPV MATERIALS AND FABRICATION BACKGROUND < 3.2.1 Fabrication llistory The Peach Bottom 2 RPV is a 251 in. BVR/4. Construction was started by Babcock & Wilcox (B&W).to the Winter 1965 addenda of the 1965 ed!?.'.on of the ASME Code. Chicago Bridge & Iron (CB&I) completed the vessel, generally to the 1968 edition of the ASME i de. The shell and head plates 4 3-1
,y l
I I are ASME SA302, Grade B low alloy steel (LAS), modified per Code Case 1339. The nozzles and closure flanges are ASTM A508 Class 21AS, modified per Code Case 1332-2, and the closure flange boiting materials are ASTM AS40 Crade B24 1AS, modified by Code Cases 1335 2 and -4 The fabrication j l process employed double quench and temper heat treatment immediately after hot forming, then electroslag or submerged are welding and post weld heat t treatment. The post weld heat treatment was typically 30 hours at 1125'F i 25'F. The arrangement of plates and welds relative to the core beltline and various nozzles is shown in Figure 3 2. 3.2.2 linterial Properties of RPV at Fabrication i A search of General Electric Quality Assurance (QA) records was made to determine the chemical properties of the plates and welds in the RPV ' beltline. Table 3-1 shows the chemistry data obtained for the beltline materials. All data shown for the beltline plates were taken from QA , records. and has been incorporated into DRF B13-01445. Available data for the beltline longitudinal electroslag welds were obtained from B&W (Reference 8), and have been evaluated in Appendix A to determine conservative values representative for the beltline longitudinal welds. The data for the girth weld, made by CB61, were in QA records. The surveillance weld was fabricated with the same electroslag weld procedure as was used in the longitudinal seam welds, but records of the actual weld metal used in the surveillance weld were not found. Only one veld wire heat was used for the vessel electroslag welds, and the usual B&W practice was to use the same heat in the surveillance wold. However, the chemical analysis results in Subsection 3.2.3 do not appear to support the assumption that the surveillance and beltline welds are the same weld heat. f This is discussed in more detail in Subsection 3.2.3. l l A search of QA records was made to collect results of certification mechanical property tests performed during RPV fabrication, specifically tensile test, Charpy V Notch and dropweight impact test results. l 3-2
Properties of the beltline materials and other locations of interest are presented in Table 3 2. The Charpy data collected were used to establish the RTNDT valuea for each vessel component, as described in Subsection 3.2.4. Charpy data from Reference 8 for the electroslag welds have been evaluated in Appendix A to determino a mean RTNDT and standard deviation. 3.2.3 Srecimen Chemical composition Samples were taken irom tensile specimens 7K1 and 7JE (base) and from 7KK and 7KL (weld) after they were tested. The tensile specimens were fabricated from the same plates and weld as were the Charpy specimens, as detailed in Subsection 3.3, so the samples chosen are representative of all surveillance specimens. Chemical analyses were performed using a plasma emission spectrometer. Each sample was decomposed and dissolved, and a portion prepared for evaluation by the spectrometer. The spectrometer was calibrated with a standard solution containing 700 ppm Fe, 8 ppm Mn, 5 ppm Cu, 5 ppm Ni, 5 ppm Mo, 5 ppm Cr, and levels' of perchloric acid and lithium consistent with the test. The calibration for phosphorus was done by analyzing two National Bureau of Standards steels with known quantities of phosphorus. The chemical composition results are given in Table 3 3. The results for the plate are consistent with QA record information. However, the B&W suggestion that the surveillance weld wire heat is the same as in the vessel electroslag welds appears to be questionable. The rested Cu content of 0.101 is not within testing uncertainty of the 0.16% to 0.21% range in Appendix A. Similarly for Ni, the test value of 0.32% is not within uncertainty limits of the 0.14% to 0.21% range in Appendix A. 3.2.4 Initial Reference Temocratures The requirements applicable to establishing the vessel component RTNDT values in the ASME Code prior to 1972 can be summarized as follows: 3-3
.__ms_ .-
1 i l a. Test specimens shall be longitudinally oriented Charpy V Notch ) specimens.
- b. At the RTNDT, no impact test result shall be less than 25 ft lb, and the average of three test results shall be at least 30 ft lb.
1
- c. Pressure tests shall be conducted at a temperature at least 60'F l above the acceptable RTNDT for the vessel.
l The current requirements for establishing RTNDT are significantly l different. For plants constructed to the ASME Code after Summer 1972, the I requirements are as follows: 1 l a. Charpy V Notch specimens shall be oriented normal to the rolling direction (transverse). j b. RTNDT is defined as the higher of the dropweight NDT or 60*F j below tho temperature at which Charpy V Notch 50 ft lb energy and 35 mila lateral expansion are met. f c. Bolt-up in preparation for a pressure test or normal operation i shall be performed at or above the RTNDT or lowest service ] temperature (1.ST), whichever is greater. l 3 Reference 1 states that for vessels constructed to a version of the , ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses must be supplemented in an approved manner. General Electric j has developed methods for analytically converting fracture toughness data , for vessels ennstructed before 1972 to comply with current requirements, i i CE developed these methods from data.in VRC Bulletin 217 (Reference 9) and-l from data collected to respond to NRC questions on FSAR submittals in the late 1970s. These methods and example RTNDT calculations for vessel plate, weld, weld HAZ, forging, and bolting material'are summarized in the remainder of this subsection. Calculated RTNDT values for selected RPV locations are given in Table 3 2. 3-4 w r-
- - -w -we an e e ., -----en.: -,yer-, , , . -v.--,, - , _a-.r a. , ,- , - ..,.e--w ,,,---,e,,,a- ,,-,,,,,,..,,---,-e,.,--.m.-
1 l i I For vessel plate material, the f',rst step in calculating RTHDT 18 t' l establish the 50 f t lb transverse test temperature given longitudinal test 1 i specimen data. There are typically three energy values at a given test l temperature. The lowest energy Charpy value is adjusted by adding 2'F per l f t-lb energy to 50 f t lb. For example, for the six beltline plates the limiting combination of test temperature and charpy energy from Table 3-2 is 43 ft lb at +10*F for Heat C2873-1. The equivalent 50 ft lb l longitudinal test temperature is: i l T30L - 10'F + [(50 43) f t lb
- 2*F/f t-lb) - 24'F The transition from longitudinal data to transverse data is made by adding 30*F to the test temperature. In this case, the 50 ft lb transverse Charpy tast temperature is T50T - 54'F. The RTNDT is the greater of NDT or 4 (T50T
- 60*F). From Table 32, the NDT for Heat C2873 1 is 20'F.
. Therefore, the RTNDT for the core beltline plate is 6*F. For vessel weld material, the Charpy V-Notch results are usually liniting in establishing RTNDT. The 50 f t lb test temperature is established as for the plate material, but the 30'F adjustment to convert longitudinal data to transverse data is not applicable to weld metal. The limiting beltline veld Charpy V Notch a gy for the submerged arc weld in ! Table 3 2 is 41 fr lb at +10', so i 4 750T - 10* F + [(50 ' - 41) f t lb
- 2'F/f t 1b] - 28'F.
As shown in Table 3 2, there are no NDT data available for the weld metal. The CE procedure requires that, when no NDT is available, the resulting i RTNDT be greater than -50'F. In this example, (T50T 60'F) is greater than 50'F, so the RTNDT is -32'F. i The initial RTNDT for the longitudinal electroslag welds is determined I in Appendix A by fitting Charpy curves to nine sets of data for the veld heat used in the vessel electroslag welds. The T50T values were determined 35 l l l---.--a-..._____.._...,.,,_.4, ,.. , - , _ . . , _ _ _ _ . . . _ _ , _ _ , . _ _ _ _ _ ._ ,_g _. . _ _ _ _ _
by statistical fit for each curve and the T50T values were used to calculate the mean and standard deviation of the electroslag weld RTNDT-For the vessel weld HAZ material, the RTNDT is assumed the same as for the base material since ASHE Code veld procedure qualification test requirements and post weld heat treatment data indicate this assumption is valid. For vessel forging material, such as nozzles and closure flanges the i 1 method for establishing RTNDT is the same as for vessel plate material. Heat EV9934 in recirculation inlet nozzle N2F, listed in Table.3-2, has a calculated value of T50T - 110'F. The NDT is 40*F. Therefore, with the RTNDT being the greater of NDT or (T50T
- 60'F), the RTNDT is 50*F. This is the highest forging RTNDT in the vessel.
For bolting material, the current Code requirements define the LST as the temperature at which transverse Charpy V Notch energy c,f 45 ft lb and 25 mils lateral expansion (MLE) are achieved, If the. required Charpy results are not met, or are not reported, but the Charpy V-Notch energy reported is above 30 ft lb, the requirements of the ASME Code at construction are applied. As shown in Table 32, the limiting Charpy V Notch energy for bolting material is 35 ft-lb at +10'F. The current requirements are not met, so the construction Code requirements are used. The LST is defined as 60*F above the temperature at which 30 ft lb Charpy V-Notch energy is achieved. Therefore, the LST of the closure bolting material is 70*F. 3.3 SPECIMEN DESCRIPTION The surveillance capsule contained 24 Charpy specimens: base metal (8), weld metal (8), and HAZ (8). There were 8 tensile specimens: base metal (3), weld metal (3), and HAZ (2). The 6 flux wires recovered were iron (2), nickel (2) and copper (2). The chemistry and fabrication history for the Charpy and tensile specimens are described in this section. 36
3.3.1 Charov specimens The fabrication of the Charpy specimens is described in the Vessel Purchase Specification (Reference 10). All materials used for specimens were beltline materials from the lower intermediate shell course. The base metal specimens were cut from 11 eat C27612, a beltline plate. The chemical analysis of this heat is in Table 3-1, and confirmed by the results in Table 3 3. The test plate was double quench and tempered and then given a stress relief heat treatment for 30 hours at 1125'F i 25'F to simulate the post weld heat treatment of the vessel. The method used to machine the specimens from the test plate is shown in Figure 3-3. Specimans were machined from the 1/4 T and 3/4 T positions in the place, in the longitudinal orientation (long axis parallel to the rolling direction). The identifications of the base metal Charpy specimens recovered from the surveillance capsule are shown in Table 3-4. The veld metal and llAZ Charpy specimens were fabricated from trim off pieces of lleats C2761-2 and C2761-1 that were velded together by electroslag welding, using the same weld process for the longitudinal seam welds in the RPV beltline. The chemical analyses of the plates are given in Table 3 1. The weld metal chemistry from two specimens is presented in Table 3-3. As discussed in Subsection 3.2.3, the surveillance weld wire heat and vessel weld wire heat do not appear to be-the same. It may be possible for B&W to trace the veld wire heat in the surveillance veld, but this would require a B&W record search beyond the scope of the current effort. I l The welded test plate for the weld and llAZ Charpy specimens received a stress relief heat treatment at 1125'F 1 25*F for 30 hours to simulate the fabricated condition of the RPV. The veld specimenc and ilAZ specimens were fabricated as shown in Figures 3-4 and 3-5, respectively. The base metal orientation in the veld and liAZ specimens were longitudinal. Contained in 3-7
I Table 34 are the identifications of the veld metal and HAZ Charpy specimens from the surveillance capsule. 3.3.2 Tensile Snecimens Fabrication of the surveillance tensile specimens is described in Reference 9. The chemical composition and heat treatment for the base, weld and HAZ tensile specimens are the same as those for the corresponding j Charpy specimens. The identifications of the base, wcld, and HAZ tensile specimens recovered from the surveillance capsule are given in Table 3 4. A summary of the fabrication methods is presented in the remainder of this subsection. The base metal specimens were machined from material at the 1/4 T and 3/4 T depth from Heat C2761 2. The specimens, oriented along the plate rolling direction, were machined to the dimensions shown in Figure 3-6.- The ga5e section was tapered to a minimum diameter of 0.250 inch at the center. The weld metal tensile specimen material was cut from the welded test plate, as shown in Figure 3 7. The specimens were machined entirely from weld metal, scrapping material that might include base metal. The fabrication method for the HAZ tensile specimens is illustrated in rigure 3 8. The specimen blanks were cut from the welded test plate such . that the gage section minimum diameter was machined at the veld fusion line. The finished HAZ specimens are approximately half weld metal and half base metal oriented along the plate rol3ing direction. l l l l i t-l 3-8 1 I
Table 3-1 CHEMICAL COMPOSITION OF RPV BELTLINE MATERIALS Composition by Weicht Percent Identification Heat / Lot No. C Mn P 5 Si Ni Mo Cu Lower Plates: Mk-57 C2791-2 0.21 1.39 0.010 0.018 0.22 0.52 0.48 0.12
- C2761-1 0.22 1.30 0.011 0.015 0.23 0.54 0.47 0.11
- C2873-2 0.22 1.39 .0.010 0.018 0.22 0.57 0.52 0.12 Lower-Intermediate Plates:
Nk-58 02894-2 0.22 1.30 0.013 0.018 0.21 0.42 0.46 0.13
- C2873-1 0.22 1.39 0.010 0.018 0.22 0.57 0.52 0.12
- C2761-2 0.22 1.30 0.011 0.015 0.23 0.54 0.47 0.11 Longitudinal Welds:
B1,B2,B3
- Heat 37C065 0.17 1.41 'O.015 0.013 0.09 0.21 0.53 0.21 l
C1.C2,C3 Lower to Lower-Int. Cirth Weld: BC Adcom INMM 0.09 1.46 0.019 0.015 0.35 0.97 0.53 0.06 Heat S-3986 Linde 124 Flux. Lot 3876 a The basis for the electroslag veld chemistry determination is in Appendix A. -_ m
l I i I Table 3-2 RESULTS OF FABRICATION TEST PROCRAM FOR SELECTED RPV ' < '.ATIONS j , Tensile ' Total Area Test Charpy l Ident. Heat Yield UTS Elong Reduc. Temp. Energy NDT T50T-60 RTNDT ; (*F) (*F) Number Number (%) Location (ksi) (ksi) (2) (ft-Ib) (*F) (*F) L G. Beltline: l Lower Shell Plates Mk-57 C2791-2 67.6 86.3 28.2 73.6 10 61,55,44 -30 -8 -8 ! C2761-1 69.4 88.8 28.0 71.5 10 47,59,63 -30 -14 -14 j j " C2873-2 73.8 91.0 28.0. 73.1 10 60,69,64 -30 -20 -20 Lower-Intermediate Mk-58 C2894-2 68.0 85.0 28.0 72.1 10 57,52,54 -30 -20 -20 i Shell Plates - C2873-1 74.8 91.5 26.6 71.5 10 43,52,48 -20 -6 -6 , C2761-2 67.0 91.8 29.3 70.5 10 57,69,58 -20 20 -20 t ] Longitudinal Velds B1,B2.B3 37C065 65.5 83.4 26.8 65.4 --see Appendix A-- -45 a -45 ! { C1,C2 C3 ' Cirth Veld BC Ht. S-3986 74.7 91.2 25.5 66.0 10 41,45,46 n/a -32 -32 Linde 124 Flux ; Lot 3869
- Non-Beltline
Upper Shell Plate Mk-60 C2796-2 10 43,53,49 10 -6 10 Vessel Flange Mk-48 124V-201 10 111,123,141 10 -20 10 , Head Flange Mk-209 SP2744 10 57,78,96 10 -20 10 1 Top Head Torus Mk-202 C3131-3 10 75.46,91 10 -12 10 Bottom Head Torus Mk-2 A0931-2 40 29,65,50 40 52 52 , j Recirculation Inlet N2F EV9934 40 30,31,31 10 50 50 I Closure Bolts Mk-61 6720443 10 35,36,37 n/a LST - 70*F ! f *4s discussed in Appenoix A, the RTNDT value has a -tandard deviation of 16.44*F. i i
. . - . ~ . . ._ _ _ _ - _ _ _ .
. . ~ _ _ _ _ . . _- ._ . _ _ - _ _ . . . _ _ _ _ _ _ _ _ . . _ . . _ _
_ - _ _ _ . . _ _ . . _ . . . _ . . _ ~ . . - . _ _ _ . _ 4 1 1 4 3 j 1 1 i Table 3 3 r ll 1 I PLASMA EMISSION SPECTROMETRY CHEMICAL ANALYSIS OF RPV f SITRVEILLANCE PLATE AND WELD MATERIALS
- Base Metal Base Metal Weld Metal Veld Metal l Elernent Tensile 7K1 Tensile 7JE Tensile 7KK Tensile 7KL l
1
- Mn 1.24 1.28 1.43 1.46 1
! P 0.009- 0.011 0.010 0.012 i Cu 0.10 0.10 0.09 0.10 l l l i Ni 0.54 0.53 0.32 0.32 l
- Ho 0.48 0.49 0.49 0.49
.i Cr 0.09 0.09 0,08 0.08 ' f t l e e I h l I I I 3-11 i
. , - - ,.,-n, ,--r-n-,.,_ , - . . , , _ . . ---.,,.,n,..-n--,, m...,,,,,-,,,,-.,,._,...unn,.-n,.n.,w- -_ ,,,,as.,- -,n ~ _.
- 4 l
l Table 3 4 ' l i IDENTIFICATION OF CHARPY AND TENSILE SPECIMENS REMOVED FROM SURVEILIANCE CAPSULE Charov Seccinght a EJULt Ettl.d 116Z 75M 7AM 701 75P 7AT 7D3 765 7E3 7DK 76K 7B4 7DL 772 7BA 7DU 77A 7BL 7J1 77D 7BT 7J3 77K 7C1 7JA Tensile Specimens a EJuit Etld 13AZ 7K1 7KD 71A 7K3 7KK 7LE 7JE 7KL
- m a All specitnens-have two dots over the centar identification digit.
3-12
-..,+.~,.r_,. , - - - - - + - ,. _ , ,
, ,,%, , , 6-,..u s.ww-. , . . , ,.,c, , o,ywm. ,, ,,,,,g,. p. .,y ,.,,.,.,y,. . _, , , , . , ,,, ..y,y., ,
4 RU CTOR CODE
'e . 10 e e e 14 8 + 16 = 2 5 i e 4 2
g e e 1 1 s If
- J i
$7
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*A
,ez
~, ,,r. .
p , ~: aG' , 1
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4 (,, $ t ., O,t..'W L~ -- - t I CAPSULE CODE o 8 e 4 CAP 5ULE 2-
- 2
= 1 e
i 4 e Figure 3-1. Surveillance Capsule Recovered f rom Peach Bottom 2 Reactor 3-13
I E E 1 N, % [ TOP HFAD LNCLOSURE CLOSuoE - 1 FLANGE 7 3 REGION 7 ) UPPER SHELL UPPER INTERMEDIATE SHELL
]
6 g h MCDd
- SHELL LOWER HEAT b HLAT INTERMEDIATE C27612 C2894 2 s SHELL CORE
\ : HEA1 C 28731 .
/
B6ILINE LONGITUDINAL C'RCUF ERENTIAL REGION SEAM WELDS G1RTS WELD C1 C2,C3 BC
/
1 B1, B2, B3
/
( ~
.@ e @Ds9 HEAT ,
LCWER C2761 1 SHELL HEAT HEAT C2873 2 C27912 BOTTOM HEAD , . _ ENCLOSURE % .
~- -
- r. i Figure 3-2. Schemstic of the RPV Showing Arrangement of Vessel Plates and Welds 3-14 l
1
800
\
\\
g\\
. . _ _. \\\ HEAT
\\\
C2701 2
\\\ \
\\\\
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\ U q\\\ f"-
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- X '"%
= =.
9 y; / o . F 1.082 90 deg 2 0 deg to min # MARKED END I UNMARKED END l
= - 2.16 5 2 0.015 -+
22112 deg 0.010t0.001R 45 deg
+ + 0.3 94 Y
y W' -90 deg 2O ceg 10 min l 0.394 ALL CORNERS 0.315 h b Figure 3-3. Fabrication Method for Base Metal Charpy Specimens 3-15
1 i RE T ON ELECTROSLAG WELD 1
- PER WR.12 4. PO 1351 1 NN MM
\\ \\ HEAT C27612
\\ \\
g sg HEAT C2781 1
\S
\\ \\
\\ \\
\\ \\
xx gg \, s NN
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\ \. I % l
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\ \
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(- N N
\
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\
s
\
k l I i/ NN' i
)(
i-- s
\ / \
\ / \
2 3/8 in. L s s. . SPECIMENS MACHINED
\l_ _ ! ~! I AS IN FIGURE 3-3 l
F4gure 3-4. Fabrication Method for Weld Metal Charpy Specimens l l l l 3-16
l l ROLLINO DIRECTION HEAT C27612
\ g \\ El ECTROSLAG WELD PER WR 12 4, PO 1351 HEAT C2761 1
< \
\ 'g \
x \ \ NN s N N \\
\ N N NN s N N 'sN N ! '
i
\ t N \
I
/ \ \ i
/ N \ I N i \
\ 'x NN N \ l SCRAP
\ i
\ xx xw l
l
\
\
l D xx s A !
/
\ l *** SCRAP $PECIMENS MACHINED l
\ f k PER FIGURE 3 3 WITH NOTCH CENTERED ON WELD FUSION UNE l
l l Figure 3-5. Fabrication Method for HAZ Charpy Specimens 3-17 s,-m- m+ n---p. w g, ,-- e-
BON
\
\ - HEAT g\ C2761-2
\\g\
\ \\
\ \s\ \\\\
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\
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sg\ y~ ,"" " 3//4T
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= *-
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i. 1.000 t 0.005 GAGE LENGTH 7/16-14 UNC-2A '
~
DOTH ENDS 3/8R (TYP) 2 1/2 l
----u-Q _, GAGE MARKS j -7s
-i,D' I ', 0 T hD' - -
30 0EG TYP ,_ b r- 314 % 1/16--*= k1/2 I R-
= = 1f4 1 1/2-in. REDUCED -
SECTION
-c 321/16 $
NOTES:
- 1. D = 0.2 60 2 0.001 DIAM AT CENTER OF REDUCED SECTIOM
- 2. D' = ACTUAL "D" DIAM + 0.002 TO O.005 AT ENDS OF REDUCED SECTION, TAPERING TO CD" AT CENTER <
)
I 1 l Figure 3-6. Fabrication Method f or Base Metal Tensile Specimens f 1 3-18
m.
\
\ \
\ \
\ \
l I
\ \
t i l l
\ \
I
\
l I i \ O TEST P TE U
\ \ ELECTOSLAG
\ \ WELD
\ \
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\ \
\ g
\ \
\ \
\ \
\ \
g
\
\
t ('; % 3
\
l SPECIMAa 44, s \ \ l DIMEN J10NED \%;Qs \ g ! AS IN FIGUF<E 3-6 s(9 s \ i i l 1 i SCRAP x# o\W Figure 3-7. Fabrication Method for Weld Metal Tensile Specimens 3-19
. _ _ _ _ - _ _ _ _ ._m. .
ROLLING DIRECil0N C27012 HEAT ' C2761 1
's N N [
N
\
g
\ N
' FLECTROSLAG WELD N x 's l x
\ I n
\ l N \I N !
N / i O SPECIMENS DIMENSIONED AS IN FIGURE 3-6
\ -
3-~ s_ N s
'slA 'W i
j N v i , . L : 8
\
N
\ l l
j j $ CRAP
/
I \. l Figure 3-8. Nrication Method for HAZ Tensile Specimens 3-20
I i
- 4. PEAR RPV FLUENCE EVALUATION i
Flux wires were analyzed to determine flux and fluence ret.eived by the surveillance capsule. An analysis combining two dimensional and one dimensional flux distribution computer calculations was evaluated to l establish the location of peak vessel fluence and the lead factors of the surveillance capsule relative to the peak vessel location. l 4.1 FLUX VIRE ANALYSIS 4.1.1 Procedure The surveillance capsule contained six flux wires: two each of iron, copper, and nickel. Each wire was removed from the capsule, cleaned with dilute acid, weighed, mounted on a counting card, and analyzed for its radioactivity content by gamma spectrometry. Each iron wire was analyzed for Mn-54 conterit, each copper wire for Co 60 and each nickel wire for Co 58 at a calibrated 4 cm or 10 cm source-to detec*.or distance with 100 cc and 80 ce Ge(Li) detector systems. The gamma spectrometer was calibrated using NBS inaterial. , To properly predict the flux and fluence at the surveillance capsule from the activity of the flux wires, the periods-of full and partial power irradiation and the zero power decay periods were considered. Operating days for each fuel cycle and the reactor average power fraction are shown in Table 4 1. Zero power days between fuel cycles are listed as well. From the flux wire activity measurements and power history, reaction rates for Fe 54 (n.p) Mn 54, Cu 63 (n a) Co 60 and Ni 58 (n.p) Co 58 were calculated. The >l MeV fast flux reaction cross sections for the iron, copper and nickel wires were estimated to be 0.212 barn, 0.00374 barn and 0.276 barn, respectively. These values were obtained from measured cross section functions determined at Vallecitos from more than 65 spectral determinations for BWRs and for the General Electric Test Reactor using 4-1
. _ _ . - . . - . _ . _ . _ _ _ _ . . _ . _ . . _ - , _ _ _ _ . . _ _ _ _ _ , . . . _ . _ _ . , _ _ . - ~ . _ _ _ _ _ , . . - _
activation monitor and spectral unfolding techniques. These data functions are applied to BWR pressure vessel locations based on water gap (fuel to vessel wall) distances. The cross sections for >0.1 MeV flux were determined from the measured 1 to 0.1 MeV cross section ratio of 1.6 4.1.2 Results The measured activity, reaction rate and detet2ned full power flux results for the sur"eillance capsule are given in Table 4 2. The >l MeV and >0.1 MeV flux values of 7.5x108 and 1.2x109 n/cm .sec from the flux 2 wires were calculated by dividing the reaction rate measurement data by the appropriate cross sections. The corresponding fluence results, 1.8x1017 and 2.9x1017 n/cm2 for >l MeV and >0.1 MeV, respectively, wore obtained by multiplying the full power flux density values by the product of the total seconds irradiated (3.275x108 sec) and the full power fraction (0.726). Generally, for long term irradiations, dosimetry results from copper flux vires are considered the most accurate because of co 60's long half-life (5.27 years). The secondary iron flux monitor reaction yielding Mn 54 gave resulta very consistent with the copper reaction despite the shorter half life of- 312.5 days for Mn 54 Nickel wire results are least accurate, because of the 70.8 day half life of Co-58. In fact, Co-58 activity is usually so low by the time the ilux wires are test'd that the nickel wires are-not tested. Consistency in resules between t .e iron and copper wire results indicates an accurate power history evaluation and a consistent core radial power shape. The accuracies of the values in Table 4-2 for a 2s deviation are estimated to be: i 51 for dps/g (disintegratians per second per gram) 10% for dps/ nucleus (saturated) 25% for flux and fluence >l MeV i 35% for flux and fluence >0.1 MeV 4 4-2
i i A set of flux wires from-Peach Bottom 2 was evaluated by General Electric in 1978. The >l MeV flux was 1.05x109 n/cw2 .sec. Differences in the current test inputs and methods would decrease that number to about 7.9x108 n/cm2 .sec. "he result from this study of 7.5x10 8 n/cm2 -see is slightly lower than the adjusted flux value. Past expsriences with dosimeters removed after one fuel cycle have generally shown high results when compared to the ten-year dosimetry. The fi it cycle dosimetry reflects power cycling due to testing, and in some cases the initial cores had an atypical power shape in the first fuel cycle, . where the peripheral fuel bundles had unusually high power levels. The ten year results reflect a lon5 Period of typical operat!.on, so the results from the current test I are considered to be more accurate. 4 f 4.2 DETERMINATION OF LEAD FACTORS The flux vires detect flux at a single location. The wires will
- therefore reflact the power fluctuations associated with the operation of
! the plant. However, the flux wires are not nexessarily at the location of peak vessel flux. Lead factors are required to relate the flux at the
- wires' location to the peak flux. These lead factors are a function of the core and vessel geometry and af the distribution of bundl in the core.
Lead factors were generated for the Peach Bottom 2 geometry and operating history. The methods used to calculate the lead factors are discussed below.
- 4.2.1 Procedung 4 ,
Determination of the lead factors for the RPV inside wall and at 1/4 T depth was done using a combination of one-dimensional and two-dimensional finite difference computer analysis. The two-dimensional analysis established the relative fluence in the azimuthal direction at the vessel surface and 1/4 T depth. A series of one-dimensional analyses were done to determine the core height of the axial flux peak and its relationship to 4-3 u ...
the surveillance capsule height. The combination of azimuthal and axial distribution results provides the ratio of flux, or the lead factor, between the surveillance capsule location and the peak flux locations. The two-dimensional DOT IV computer program was used to solve the Boltzman transport equation using the discrete ordinate method on an (R,r) geometry, assuming a fixed source. Eighth core symmetry was used with periodic boundary conditions at 90' and 135'. Neutron cross sections were determined for 26 energy groups, with angular scattering approximated by a third-order Legendre expansion. A schematic of the two dimensional vessel model is shown in Figure 4 1. A total of 113 radial intervals and 45 azimuthal intervals were used. The model consists of an inner and outer core region, the shroud, water regions inside and outside the shroud, the vessel wall and an air region representing the drywell. Flux as a function of azimuth was calculated, establishing the azimuth of the peak flux and its magnitude relative to the flux at the wires' location of 120'. The one-dimensional computer code (SN1D) was used to calculate radial flux distribution at several core elevations at the azimuth angle of 115.5', where the azimuthal peak was determined to exist. The elevation of the peak flux was determined, as well as its magnitude relative to the flux at the surveillance capsule elevation. 4.2.2 Results The two-dimensional calculation indicated the ilux-to be a maximum 25.5' past the RPV quadrant references (0*, 90', etc.). The peak closest to the 120' location of Capsule 2 is at 115.5', as shown in Figure 4-2. The distribution calculations establish the lead factor between the surveillance capsule location and the peak location at the inner vessel wall. This lead factor is 0.95. The fracture toughness analysis done is based on a 1/4 7 depth flaw in the beltline region, so the attenuation of the flux to that depth is considered. The resultins lead factor from the
~
capsule to the 1/4 T depth at tne peak location is 1.38. 4 . . . . . . . .
4.3 ESTIMATE OF 32 EFPY FLUENCE
-The fluence at 32 EFPY is estimated by taking the best estimate or-upper bound of the measured flux from Table 4 2 and the 1/4 T lead factor.
The time period 32 EFPY is 1.01x10 9 secords. The resulting 32 EFPY fluence values are: Best Estimate: (7.5x10 8 n/cm2 -s)(1.00)(1.01x109 s)/1.38 - 5.5x1017 n/cm2 Upper Bound: (7.5x10 8 n/cm2 . )(1,23)(1,oixio9 s)/1.38 - 6.9x1017 n/cm2 , The upper bound fluence was used for analysis purposes in this report. 4 --
Table 4 1
SUMMARY
OF DAILY POWER HISTORY Opere:ing Percent of Days Between Cycle Cvele Dates Days Full-Power Cveles 1 2/15/74 - 3/27/76 772 0.634 88-2 6/24/76 - 4/27/77 308 0.742 140 3 9/15/77 - 9/8/78 359 0.815 30 4 10/18/78 - ?/21/80 521 0.862 146 5 8/15/80 - 2/19/82 554 0.771 132 6 7/2/82 - 4/28/84 667 0.646 440 7 7/13/85 - 3/13/87 121 0.715 3790 0.726 (average) 4-6 I
i Table 4-2 SURVEILIANCE CAPSULE LOCATION FLUX AND FLUENCE FOR IRRADIATION FROM 2/15/74 TO 3/13/87-i Wire dps/g Element Reaction Rate Full Power Flux
- Fluence !
{ Wire Weight (at end of [dps/ nucleus (n/cm2-s) (n/cm2) l (Element) __ (e) Irradiation) (saturated)1 >l MeV >0.1 MeV >1 MeV >0.1 MeV 3 Copper 64945 0.3742 8.36x10 2.80x10}f8 , Copper 64946 0.3706 8.49x10 2.84x10 8 9 1 17 Average - 2.82x10~ 7.5x10 1.2x10 1.8x10 2.9x10 ' i 0 Iron 64945 0.1017 5.74x104 1.55x10[f66 j p 1ron 64946 0.1041 6.13x10 1.66x10
-6 i Average - 1.60x10 7.6x10 4
i 6 Nickel 64945 0.2800 1.14x10 2.27x10[7 i Nickel 64946 0.2855 1.16x10 2.33x10 2
-16 8 Average - 2.30x10 8.3x10 l-i
!
- Full power of 3293 MW .
3
/ i 1 1 ff 1 1 1
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/ 2 1 1 1 1 1 1 1 1 1 2 2 2 1 1 1 1 1 1 1
, CORE INTERIOR:
14 INTERVALS ' 2 2 2 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 O 2 2 2 2 2 2 2 4 CORE EXTERIOR: 10 INTERVALS
%ATER REGION.
23 INTERVALS
) SHROUO: 7 INTERMS
+ CAPSULE 135' "
WATER REGION: 35 INTERVALS 45 INTERVALS IN AZlMUTHAL d } VESSELWALL 17 INTERVALS DIRECTION , DRYWELL AIR REGION: j 7 INTERVALS 90' I = CORE INTER!OR PJEL 2 = CORE EXTERIOR FUEL Figure 4-1. Schematic of Model for Two-Dimenszonal Flux Distribution Analysis 4-8
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m m 4 - 4 6--,A--,,----Ae. 4a--- .s a m+,,e,m,- p. n 4 a n: w - , 5. CHARPY V NOTCH IMPACT TESTING i ! The 24 Cherpy specimens recovered from the surveillance capsule were impect tested at temperatures selected to establish the toughress l transition and upper shelf of the irradiated RPV materiala. Testing was conducted in accordance with ASTM E23 82 (Reference 11), 5.1 IMPACT TEST PROCEDURE The testing machine used was a Riehle Model PL 2 impact machine, serial number R-89916. The pendulum has a maximum velocity of 15.44 f t/sec i l and a maximum available hammer energy of 240 ft-lb. The test apparatus and operator were qualified using U.S. Army Watertown standard specimens. The standards are designed to fail at 70.3 f t-lb and 15.0 f t lb at a test temperature of -40'F. According to Reference 11, the test apparatus averaged results must reproduce the Watertown design values within an accuracy of 15% or 11.0 f t-lb, whichever is greater. The successful ! qualificatian of the Riehle machine and operator is summarized in Table 5 1. Charpy V Notch tests were conducted at temperatures between -20'F and 300*F. For test, between 70*F and 212'F, the temperature conditioning riuid was water. Methanol was used at temperatures below 70*F. -Above 212*F, a silicone oil was used. Cooling of the conditioning fluids was done with liquid nitrogen, and heating by an immersion heater. The fluids were mechanically stirred to maintain uniform temperatures. The fluid temperature was measured with a calibrated chromel-alumel thermocouple. Once at test temperature, the specimens were manually transferred with centering ton 5s to the Riehle machine and impacted within 5 seconds. 5-1
For each Charpy V Notch specimen tested, test temperature, energy , absorbed, lateral expansion, and percent sherr were evaluated. Lateral expansion and percent shear were measured according to Reference 11 methods. Percent shear was determined with method one of Subsection 11.2.4.*J of Reference 11, which involves measuring the length and width of the cleavage surface and locating the percent shear value from Tables 1 or 2 of Reference 11. 5.2 IMPACT TEST RESULTS Eight Charpy V Netch specimens each of base actal, weld metal and HAZ were tested at temperatures selected to define the toughness transition and upper shelf portions of the fracture toughness curve. Absorbed energy, lateral expansion and percent shear data are listed in Table 5-2 for each material. Plots of absorbed energy data for base, weld, and HAZ metal are presented in Figures 5 1, 5 2 and 5-3, respectively. Lateral expansion plots for base, weld and HAZ metal are given in Figures 5-4, 5-5 and 5-6, respectively. The data sets are fit with the hyperbolic tangent function developed by Oldfield for the EPRI Irradiated Steel Handbook (Reference 12): Y - A + B
- TANH [( T To )/C),
where Y - impact energy or lat.aral e::pansion T - test temperature, and A, B, To and C are determined by non-linear re5ression. The TANH function is one of the few continuous functions with a shape characteristic of low alloy steel fracture toughner,s transition curves. Photographs were taken of the fracture surfaces for each specimen. The fracture surface photographs were used to evaluate percent shear. The photographs and a su= mary of test results for each specimen t.re contained in Appendix B. 5-2
(
)
i 1 i j 5.3 IRRADIATED VERSUS UNIRRADIATED CHARPY V-NOTCH PROPERTIES i As a part of the RPV fabrication test program, some Charpy V-Notch l i testing was performed. Data for the plat material specimens (Heat l C2761-2) were recovered from QA records, but consisted only of three j specimen tests at 10'F. For the surveillance veld. vositive identification l ) pf the weld wire heat was not obtained from BW. Therefore, unirradiated l 1 archive surveillance specimens were obtained from-PECO.and tested to l ]' establish the baseline needed to determine the irradiation shift. l I i f To minimize personnel exposure, the unirradiated specimens were tested l l on the Tinius-Olsen Charpy test apparatus in San Jose (serial number l 2 i 119o37), with available hammer energy of 264 ft-lb and impact velocity of l ! 16.8 ft/sec. The machine was calibrated with NIST specimens in May'1990. l , The calibration results, shown in Table 5-3, were within specifications. l Twelve each of plate and weld material specimens were tested at l t temperatures between -40'F and 300*F. . Impact energy, lateral expansion and l percent shear were determined with the same methods used for the irradiated l l specimens. The results are presented in Table'5-4. Photographs of l l specimen fracture surfaces are in Appendix C. l i The impact energy and MLE data for the unirradiated-materials were fit l
- with the TANH function in the same manner as the irradiated data. The l i
- results for plate and weld impact energy are plotted in Figures'5-7 and l 5-8, respectively, along with the corresponding irradiated data described l
! in Section 5.2. The results for lateral expansion are plotted in l Figures 5-9 and 5-10 for the plate and weld material, respectively. l l The irradiated and unirradiated Charpy V-Notch data were used to' j estimate the values given in Table 5-5: 30 ft-lb, 50 ft-lb, and 35 MLE l index temperatures, and the'USE for-both irradiated materials and l- . unirradiated materials, where-available. Table 5-5 values were not l determined for irradiated HAZ material because of the caount of scatter in l l i-l i l 5-3 ( m i w- w- r--- %a -h-%+m4"H 1 uwwr s YT-P *F'-4"".-wup gr Wga.---=V*O t19WV T%'T* * *
- Y'wN +'T'WTY***T*'9'WMN'r1"NY YD'-W T 'gt 'T' 't#'*T'***T-**-'N'G-'r'Pd'"--*T'-T'NW"4*M""
the Charpy data, which is typical for the material. Transition temperature ' shif t values are determined as the change in the temperature at which 50 l ft-lb impact energy is achieved, as required in Reference 4. l 5.4 COMPARISON TO PREDICTED SHIFT l The measured transition temperature shifts for the plate and weld l materials were compared to the predictions calculated according to l Regulatory Guide 1.99, Revision 2 (1.99). The 1.99 methods used to l calculate shift are described in Section 7.6. The inputs for the l surveillance plate and weld materials are as follows: l Plate: (based on Table 3-3 values) l Copper: 0.10% l Nickel: 0.54% l fluence: 1.8 x 1017 n/cm2 l Eeld: (based on Table 3-3 values) i Copper: 0.10% l Nickel: 0.32% l fluence: 1.8 x 1017 n/cm2 l , The chemistry factors (CF) are 65 for the piste and 84.2 for the veld. l The fluence factor is 0.1611. The predicted shifts are, therefore, as l follows: l ARTNDT ARTNDT + Margin Actual Test l Plate 10,5'F 21.0*F -5'F l Veld 13.6*F 27.2*F 17'F l The measured shift of -5*F for the plate material is considerably less l than both the mean and upper bound prediction of 1.99. The measured shift l of 17'F for the veld material is slightly more than the - mean prediction. l but is considerably less than the upper bound prediction. ' 5-4 1
- ..- . - . - .. . . . .- . .- . . . .-- - - = .-..
Table-5 1 QUALIFICATION TEST RESULTS USING U.S. ARMY WATERTOWN SPECIMENS (TESTED IN MARCH 1988) Energy Absorbed Qualification Test Test Temperature Mechanical Gage Snecimen Identificati2D ('F) (ft-lb) KK10-0681 -40 69.0 KK10 0525 72.5 KK10-0561 " 70.0 KK19-0527 68.3 KK10-0662 " 65.5 Average 69.1 Allowable -40 70.3 3.5 Acceptable LL5-0203 -40 14.3 LL5-0066 " 16.5 LLS-0493 " 15.0 LL5-0395
- 17.0 LL5 0162
- Average 15.7 Allowable -40 15.0 1.0 Acceptable I
5-5 l
-_ _ . . . . . - - . ~ . . . _ . , . . . . .._. _ .. .---_~-1..- . . _ , , - - - - _ . _ . ,
Table 5 2 CHARPY V-NOTCH IMPACT TEST RCSULTS FOR IRRADIATED RPV MATL4IALS Test Fracture Lateral Parcent Shear Specimen Temperature Energy Expansion (Method 1) Identification (*F) (ft-lb) (mils) (%) Base: 772 -20 22.0 24 6 765 0 47.0 38 18 76K 20 54.5 46 26 77A 40 58.0 50 34 77D 70 81.5 70 40 75M 140 131.0 98 100 75P 200 132.0 92 100 77K 300 136.0 90 100 Ueld: 7BT -20 15.0 16 1 7AT 20 34.0 32 9 7C1 40 31.0 34 8 7B3 60 53.5 48- 31 7BA 70 65.0 58 28 7AM 140 95.5 79 83 7BL 200 108.0 86 100 7B4 300 114.5 88 100 HAZ: 7D1 -20 24.0 21 4 7J3 20 47.0 45 H1'2 7DK 40 52.0 44 26-
- 7J1 60 75.0 64 50 7D3 70 55.0 49 25 7DU 140 102.5- 80 83 7JA 200 116.0 87 100 7DL 300 1152.0 87 100 5-6
__ _ _ - _ - -- _ .- . _ . - . ~ . - _ . -._ ... _ -. ~.. - -. .. ..__.- _ ... _-, __ _ ..-_. . ,._..~. _ -_.. -
, Table 5-3. l SAN JOSE TINIVS-OLSON CHARPY MACHINE l QUALIFICATION TEST RESULTS USING l NIST STANDARD REFERENCE SPECIMENS l (TESTED 5/14/90) l Test. Energy Acceptable l Specimen Bath Temperature Absorbed Range l Identification Medium (*F) (ft-lb) (ft-lb) l MM-16 307 Methanol -40 71.0- l MM-16 465 Methanol -40 72.0 l MM-16 509 Methanol -40 75.0 l MM-16 552 Methanol -40 71.5 l MM-16 915 Methanol -40 73.0 l Average 72.5 69.4 1 3.5 l LL-19 124 Methanol -40 11.5 l LL-19 354 Methanol -40 11.5 l LL-19 411 Methanol -40 12.0 l LL-19 702 Methanol -40 10.5' l LL-19 925 Methanol -40 11.5 l Average 11.4 11.9 1.0 l l
5-7
Table 5-4 CHARPY V-NOTCH IMPACT TEST RESULTS FOR UNIRRADIATED RPV MATERIALS Test Fracture Lateral Percent Shear Specimen Temperature Energy Expansion (Method 1) Identification (*F) (ft-lb) (mils) (%) - Base: l 75K -40 18.5 14.5 0 763 -20 20.0 17.0 15 764 0 30.0 28.0 20 76C 20 58.0 44.0 36 766 20 68.0 49.0 33 75T 40 77.0 58.0 40 75L 60 100.0 70.5 60 76A 80 112.0 79.0 82 75Y 120 117.0 82.5 89 76B 160 123.5 88.5 100 762 200 123.5 84.5 100 76D 300 134.5 84.5 100 Weld: 7AE -40 17.5 14.0 0 7BB -20 20.0 19.0 6 7BC 0 32.0 27.0 8 s 7BK 20 38.0 28.5 10 7B1 40 35.5 40.0 15 7AU 6C 73.5 57.0 37 7C6 60 63.0 50.0 38 l 7BP 80 68.0 53.5 53 7B2 120 91.0 67.0 77 7BU 160 100.5 80.0 88 7B7 200 106.0 81.5 100 7CE 300 111.0 76.0 100 5-8
l
; Table 5 5 SIGNIFICANT RESULTS OF IRRADIATED AND UNIRRADIATED CHARFY V-NOTCH DATA Upper a Shelf Index Temocrature ('F) Energy (ft-lb)
I Material E-30 fr-lb E-50 ft-lb MLE-35 mil L/T Unirradiated Plate -9 13 7 126/82 l Irradiated Plate -14 20 2 135/88 l Difference -5 7 -S 9 l Unirradiated Veld 4 43 27 110 l l Irradiated Veld 21 56 33 113 l i l l Difference 17 13 6 3 l i l l l l i l a Longitudinal (L) USE is read directly from Figures 5-1 and 5-2. Transverse (T) plate USE is taken as 65% of the longitudinal USE, according to Reference 6. L/T USE values are equal for-weld metal, which has no orientation effect. 5-9 1
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- 6. TENSILE TESTING Eight round bar tensile specimens were recovered from the surveillance capsule. Uniaxial tensile tests were conduct ,1 in air at room temperature, RpV operating temperature, and onset of upper shelf temperature. Tests were conducted in accordance with ASTM E8-81 (Reference 13).
6.1 PROCEDURE All tests were conducted using a screw-driven Instron test frame equipped with a 20-kip load cell and special pull bars and grips. Heating was done with a Satec resistance clamshell furnace centered around the specimen load train. Test temperature was monitored and controlled by a chromel-alumel thermocouple spot-welded to an Inconel clip that was friction-clipped to the surface of the specimen at its midline. Before the elevated temperature tests, a profile of the furnace was conducted at the test temperature of interest using an unirradiated steel specimen of the same geometry. Thermocouples were spot-welded to the top, middle, and bottom of a central 1 inch gage of this specimen. In addition, the clip-on thermocouple was attached to the midline of the specimen. When the target temperatures of the three thermocouples were within 15'F of each other, the temperature of the clip-on thermocouple was noted and subsequently used as the target temperature for the irradiated specimens. All tests were conducted at a calibrated crosshead speed of 0.005 inch / min until well past yield, at which time the speed was increased to 0.05 inch / min until fracture. A one inch span knife edge extensometer was attached directly to each specimen's central gage region and was used to monitor gage extension during test. The test specimens were machined with a minimum diameter of 0.250 inch at the center of the gage length. The three base metal specimens were tested at room temperature (RT - 70'F), onset of upper shelf temperature (estimated at 130*F), and RPV operating temperature (550*F) . 6-1
k
?
t I
; The three weld metal specimens were tested at room temperature, onset j of upper shelf temperature (estimated at 18C'F) and 550'F. Tho two llAZ ,
specimens 1:ere tested at room temperature and 550'F. The yield strength I (Ys) and ultimate tensile strength (UTS) were calculated by dividing the nominal area (0.0491 in.2) into the 0.2% cffset load and into the maximum test load, respectively. The values listed for the uniform and total elongations were obtained from plots that recorded load versus specimen i extension and are based on a one inch gage length. Reduction of area (RA) e valueu were determined frow post test measurements of the necked specimen ; diameters using a calibrated blade mic:ometer and employing the formula: RA - 1001 * (Ao - Af)/Ao After testing, each broken specimen was photographed end on showing the fracture surface and lengthwise showing fracture location and local neckin6 behavior. 6.2 RESULTS Tensile test properties of YS, UTS, RA, uniform elongation (UE) and total elongation (TE) are preset,ted in Table 6 1. Shown in Figure 6-1 is.a stress strain curve for a 550*F base metal specimen typical of the stress strain characteristics of all the specimens tested. Shown graphically in Figures 6 2 and 6 3 are the data in Table 6 1. Photographs of fracture surfaces and necking behavior are given in Figures 6 4, 6 5 and C 6 for base, weld and HAZ specimens, respective ~y. The base, weld and ilAZ materials Senerally follow the trend of doctansine. properties with increasing temperature. The nick in specimen /M3 ep a rs to have made little difference in the results, l 62
._ ,_ _ _ . _ . _ _ _ , _ , _ . _ _ _ . , _ . . _ . , _ _ _ _ ~ _ _ -_ ._ , _ _ _ _
I I 1 ! 1 i a 4 6.3 IRRADIATED VERSUS UP' 7 J ADIATED TENSILE PROPERTIES i 1 j Unirradiated tensile test data were recovered from QA. records for the surveillance specimen plate (lleat C2761 2), as shown in Table 3 2. The i l ! unitradiated data provide average values of YS, UTS, RA and TE at room f temperature. These are compared in Table 6 2 to the irradiated base metal specimen RT data to determine the degree of irradiation effect. The trends of increasing YS and UTS and of decreasing.TE and RA, characteristic of l irradiation embrittlement, are seen in the base metal data. i 1-4 r i - i i e i h. i 4 1 i i i f i l 1-i 63 p.
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- j 1
Table 6 1 J ) TENSILE TEST RESULTS FOR IRRADIATED RPV tMTERIALS 1 4 Test Yield Ultimate Uniforn Total Reduction Specimen Tetrp Strength Strength Elongation Elongation of Area j .,Egaber_ Material I*Il (ksi) (ksi)_ ft) (t) (t) i il 7K1 Baso 69 67.9 92.9 9.0 19.0. 69.3 , 1 7K3 a Base 130 67.5 89.. 8.9 17.5 64.8 f 7JE Base 550 62.9 89.4 9.2 17.9 58.5 , 7KK Veld 71 64.2 87.6 10.1 20.0 67.8 1 7KD Veld 180 70.0 81.7 8.6 18.4 62.1 4 7KL Weld 550 58.1 80.8 8.8 17.2 69.5 7LA HAZ 72 62.5 85.7 7.5 16.1 65.8 .
?!I IMZ 550 59.4 80.6 6.4 14.4 64.4 T
[ a Specimen was nicked, but fracture results appear reasonably unaffected.- l ; 64 _..___.......___a__..__.--__;,,_. ._.,_;___.,_.,,_,, . _ , _ ,
Table 6 2 COMPARISON OF UNIRRADIATED AND' IRRADIATED TENSILE PROPERTIES AT ROOM TEMPERATURE , Yield Ultimate Total Reduction l Strength Stren6th Elongation of Area l' (ksi) (ksi) (%) (%) i j Base (Heat C2761 2): t l I Unirradiated 67.0 91.8 29.3 70.5 Irradiated 67.9 92.9 19.0 69.3 i Difference a 1.3% 1.2% 54.2% 1.7% l i i 1 I 1 i l i i l i i a Difference - [(Irradiated Unirradiated)/ Irradiated)
- 100%
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4 i 4
- 7. DEVELOPMENT OF OPERATING LIMITS CURVES
- i j 7.1 BACKCROUND 3
l Operating limits for pressure and temperature are required for three categories of operation: (a) hydsostatic pressure tests and leak tests, 1 referred to as Curve A; (b) non nuclear heatup/cooldown and low level physics tests, referred to as Curve B; and (c) core critical operation, } referred to as Curve C. There are three vessel regions that affect the 1 i operating limits:_ the closure flange region, the core beltlir region, and 3 4 the remainder of the vessel, or non beltline regions. The closure flange 1 region limita are controlling at-loaer pressures primarily because of t Reference 1 requirements- The non-beltline and beltline region operating ; limits are evaluated according to procedures in_ References 1 and 2, wich the beltline region minimum temperature limits increasing as the vessel is i ! irradiated. 6 j 7.2 NON. BELTLINE REGIONS 3 Non beltline regions are those locations that receive too little fluence to cause any RTNDT increase, Non beltline components include the i nozzles, the closure flanges, some shi ) plates, top and bottom head plates i and the control rod drive (CRD) penetrations. Detailed stress analyses of the non beltline components were performed for the BWR/6, The analyses took into account all mechanical loadings and thermal transients anticipated. Detailed stresses were used according to Reference 2 to i develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T RTNDT). These results are applicable to_the Peach Bottom 2 vessel _ components, since the Peach Bottom 2 geometries are + not significantly different from BWR/6 configurations and the mechanical and thermal loadings are comparable. 71
The non beltline region results were established by adding the highest RTNDT for the non beltline discontinuities to the P versus (T - RTHDT) curves for the most limiting BWR/6 components, which are the CRD penetration and feedwater nozzle. Shown in Figures 7-1 through 7-3 are the Peach Bottom 2 unique calculated non beltline operating litnits for Curves A, B, and C, respectively. The BWR/6 component curves are adjusted to reflect the limiting non beltline RTNDT of the Peach Bott.om 2 vessel, which is $2'r for the bottom head torus plates. 7.3 CORE BELTLINE REGION The pressure temperature (P T) limits for the unitradiated beltline region are shown in Figures 7-1 through 7-3. As the beltline fluence increases during operation, these curves shift to the right by an amount discussed in Subsection 7.6. Typically, the beltline curves shift to become more limiting than the non beltline curves. The stress intensity factors calculated for the beltline region according to Reference 2 procedures are based on a combination of pressure and thermal stresses. The pressure stresses were calculated using thin walled cylinder equations. Thermal stresses were calculated assuming the through-wall temperature distribution of a flat plate subjected to a 100'F/hr thermal gradint. An initial RTNDT of 6'F for the limiting plate is used to adjust the (T - RTNDT) values from Figure G 2210 1 of Reference 2.- 7.4 CLOSURE F1ANGE REGION Reference 1 sets several minimum requirements for pressure and temperature in the closure flange region in addition to those outlined in Reference 2. In some cases, the results of analysis for other regions exceed these requirements and they do not affect the shape of the P T curves. However, some closure flange requirements from Reference 1 do impact the curves. In addition, General Elcetric recom:nends 60*F margin on the required bolt preload temperature. 72 b -v *"
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i As stated in Paragraph C 2222(c) of Reference 2, for application of fi:11 bolt preload and reactor pressure up to 20% of hydrostatic test pressure, the RPV 4etal temperature must be at RTNDT or greater. The GE I { practice is to require (RTNDT + 60'F) for bolt preload, for two reasons: ) i h i a. The original ASME Code of construction requires (RTNDT + 60*F); ] 3 and 3 j i
- b. The highest stressed region during boltup is the closure flange l region, and the flaw size assumed in that region (0.24 inches) is s
less than 1/4 T. This flaw size is detectJ,ie using ultrasonic j testing (UT) techniques. In fact, References 14 and 15 report j that a flaw in the closure flange reSi on of 0.09 inch can be j reliably detected usind UT. 4 I For Peach Bottom 2, (RTNDT + 60*F), or 70*F, is consistent with the j allowable LST for the bolting material. Furthermore, (RTNDT + 60*F) is a j requirement for Curve C, as described in an traph IV.A.3 of Reference 1. ! Reference 1, para 5raph IV. A.2, sets temperature minimum requirements i
- for pressure above 20% % drotest pressure base on the RTUDT of the closure
- region. Curve A temperature mast be no less than (RTNDT + 90*F) and 1
- Curve B temperature no less than (RTNDT + 120*F). The Curve A requirement causes a 30'F shift at 20% hydrotest pressure (312 psig) as shown in l
Figure 7 1. The Curve B requirement has no impact on Figure 7-2 because
- _ the analytical results for the feedwater nozzle require that temperature be l greater than (RTNDT + 120'F) at 312 psig.
i i i 7,5 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX C
- Curve C, the core operation curve shown in Figure 7 3 is generated
- from Figure 7 2, accounting for the requirements of Reference 1 parcgraph IV.A.3. Essentially pr,ragraph IV.A.3 requires'that core critical P-T-limits be 40'F above any Curve A or B limits. Curve B is more limiting l
l ! 7-3 h
3 i t 5 l 4 i I
- than Curve A. so Curve C is Curve B plus 40*F. The (RTNDT + 60*F) minimum j permissible temperature mentioned in Subsection 7.4 for curve C is an i
j exception for BWRs, allowing critical operation at temperatures below the j hydrostatic cest temperature. ] 7.6 EVALUATION OF RADIATION EFFECTS , l The impact on adjusted reference temperature (ART) due to irradiation 2 in the beltline materials is detsrm'ined according to the methods in Reference 5, as a function of neutron fluence and the element contents of copper (Cu) and nickel (Ni). The specific relationship from Reference 5 is: , ART - Initial RTNDT + SHIFT + Margin
~
(7 1)_ vhere: SHIFT - [CF]*f(0.28 0.10 los f) (7-2) Margin - 2*(aI2 + ,32)1/2 (7 3) CF - chemistry factor from Tables 1 or 2 of Reference 5, i f - fluence (n/cm )2 divided by 1019, ! ay - standard deviation on initial RTNDT. og - standard deviation on RTNDT shift, is 28'F for velds and 17'F for base material, except that og need not exceed 0.50 times the SHIFT value. 7-4
. - -.-. - - ~ . . - . . - - - . . . . = . . - . . . - . . . - . . . - , - . + - - . . - , , , ,. -,.
_ __ _ ..__ _ _ __ ~ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _.. ._. _ - . _ _ _ _ _. _ ._ . _ _ . The limitin5 beltline plate and weld are determined based on the Cu Ni content and initial F.TNDT of the materials. Calculations based s he information in Tables 3 1 and 3 2 show the following: Limiting plate: 0.121Cu, 0.571Ni, Initial RTNDT - 6'F. ay - 0*F Limiting veld: 0.211Cu, 0.211Ni, Initial RTNDT - 45'F, og - 16.4'F This material information is used to evaluate irradiation shift versus ; fluence. 7.6.1 Measured Versus Predicted Surveillance Shift . Table 5 5 presents an estimated measured shift for the base and weld metal of .5'F and 17'F, respectively. The predicted shifts of the surveillance plate and weld are detertnined in Section 5.4 Assuming the best estimate fluence for the surveillance capsule of 1.8x1017 n/cm2 , the predicted values of (SHIFT + Margin) are 21*F for the plate and 27.2'F for the weld. Reference 5 states that surveillance data may be used in place of Equations 7 2 and 7 3 when two sets of credible data are available. This is only the first set of data for Peach Bottom 2. Therefore, determinations of ART for the purposes of developing pressure temperature curves are based on Reference 5 predictions. 7.6.2 ART Versus EFPY Equations 7-1 through 7 3 can be plotted as a function of EFPY for the base metal and weld metal. Subsection 4.3 concludes that the 32 EFPY 1/4 T fluence is 6.9x1017 n/cm 2 . Figure 7-4 shows the ART tor the limiting plate and weld materials, based on the initial RTNDT values and chemistry data in Subsection 7.6. As shown in the figure, the plate is the limiting beltline material through 32 EFPY, because of its higher initial RTNDT. l 75 { l
7.6.3 Fracture Touthness Conditions at 32 EPPY Paragraph IV.B of Reference 1 sets limits on the ART and on the upper { shelf energy (USE) of the beltline materials. The ART must be less than > 200*F, and the USE must be above 50 f t lb. Based on Figure 7 4, the ART j values at 32 EFPY of 43*F for the weld and $1*F for the plate are J J acceptable. ) { calculations of decrease in USE, using Reference 5, are summarized in Table 7-1. The equivalent transverse USE of the plate material is taken as
- 65% of the longitudinal USE, accordin5 to Reference-6. The weld metal USE l
] has no transverse / longitudinal correction because weld metal has no orientation effect. The surveillance plate result is used as the measured f reference data - for USE, because USE data were not obtained for the unirradiated condition. The highest charpy energy value from Appendix A was assumed as the USE for the electroslag veld material. Extrapolating to-the 32 EFPY fluance according to Reference 5, the predicted USE for the plate is 86 f t-lb, and for the veld is 80 ft-lb. There were no USE data available in fabrication records for the other beltline. plates or the circumferential weld. Typical plate material of the type used in the Unit 2 beltline, for which USE data are known, indicate that the Uni.t 2 plates have adequate USE, especially considering the copper contents of the plates and the low fluence predicted for 32 EFPY. Similarly, given the low copper, low fluence and typical USE of submerged are welds, the circumferential weld is expected to have adequate USE. Therefore, irradiation effects are not severe enough to necessitate RPV annealing before 32 EFPY. 7.7 OPERATING LIMITS CURVES VALID TO 32- EFPY The ART selected for the core beltline curves depends on the amount of-operation for which the curves will be valid. Operation to 32 EFPY was assumed because, in reviewing the shift in beltline curves, it was determined that the non beltline curves are still limiting at 32 EFPY, The 76
-_ ..- , _- ._.__.._u,a.._ , _ _ - . - ~ . . -- ~_ . _ _ - _ _ . _ _ _
beltline ART for 32 EFPY estimated with Figure 7 4 is S1'F based on the plate material. Adjusting the unitradiated beltline curves in Figures 7 1 through 7 3, with their initial RTHDT Of *6'F, the beltline limits are at . least 20*F lower than the non beltline limits. Therefore, the 32 EFPY limits are the non beltline region curves, as shown in Figures 7 5, 7-6 and 7 7, respectiveAy, The beltline shift is based strictly on predictions of Regulatory Guide 1. 9'), Revision 2. The surveillance data are r.ot factored into the shift calculations because the Regulatory Guide requires two credible data points for each material, and Peach Bottom 2 now has only one data point for plate material, When the second surveillance capsule is tested, around 15 ETPY, the pressure temperature curves may require revision to account-for the beltline shift as predicted by the surveillance test results at that time. 7.8 REACTOR OPERATION VERSUS OPERATING LIMITS For most reactor operating conditions, pressure and temperature are at saturation conditions, which are.in the operating zone of the limits curves. The ost severe unplanned transient is an upset condition consisting of several transients which result in a SCRAM. The worst combination of pressure and temperature during this postulated' event is 1180 psi 5 Vith temperatures in the lower head at 250'F. At the same time, the steam space coolant temperature is still nearly 550*F. Steam space coolant temperature is-_used to identify the appropriate curve'to'ba , cpplied. In this case, the core is not critical and, according to the steam space coolant temperature, there is no significant cooldown occurring, so the hydrostatic pressure curve applies (Curve A), As seen for Curve A in Figure 7 5, at 1180 psi the minimum transient temperature in the vessel of 250'F lies in the safe operating zone. Therefore, violation of the operating limits curves is only a concern in cases where operator interaction occurs, such as hydrostatic pressure testing and initiation of-criticality. 7-7
l i l Tabin 7 1 ESTIMATE OF UPPER shelf ENERGY FOR BELTLINE MATERIALS , Upper Shelf (ft.lb) Longitudinal / Transverse , 17 Identification L.fu Unirradiated f=1. B r[ f-6.9x10 Plate: Heat C2761 2 0.11 . 135'/88 132/86 l Weld: b Heat 37C065 0.21 99 . 80 h l " USE value taken from surveillance test data. Other USE. values are calculated using Reference 5. b USE value taken from fabrication test data in Appendix A. 78
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__J-. !..__ 4.. i 4 . ... ,! i - l
^
- .4.-- .-- ., .. - ;. , . . _ .
O 0 100 200 300 400 500 500 RPV Metal Temperature ( F) Figure 7-1. Components of Operating Limits Curve for Pressure Tests (Curve A) 7-9
, . _. ._m _ _ __ _
, i 7
'- 1
+3 I
, e I i ++
i_ _ 3 g i! i _1 _
~
UNIRRADIATED i i ! i
, i DELTLINE REGION i ,
l-
' d 0 ' ' ! ' I 4 (RTNDT = .6 F)
,i ,
1400 l l r, , , i # i _ 1 I r I i l -F- l ! I H t i ! n ! _i > 1200 , i , , ,
, i i
, . . t !
%r 3 J i i . i .
; . ,- i !
, e i , .
a l i I 6 1 1 1 , l I 1 i i i 1000 ; , , i . 1 i m < , ; . i n ; ~ , y 1 1
- i -
j NON. BELTLINE REGION c- - ' ' ' 0 i o - i/ (RTNDT = 52 F) H F \ E 800 ., -l . cc , 1 f , .ce ,a l v 6 2 I l - m__ ~l e '
~ ..
g,, .g I I 600 f / , 8
- - /
/ H ,
l- - l - h ! / f ! O , / 1 $ / ,
//. 10CFR50 APPENDIX G, IV.A.2 - -
> /I 0 r REQUIREMENT (RTNDT = 10 F) 400 ,, ,i i
i
/ ,
I i i < ! r ~ 8 i i
, , ___,I . ,i , ,
, ~ , ,
~
I 200 ,
.f.' _ . - ,- , ,
7 ;
. /
- L. !
- ._' ; j~ ~ '
. i
~
BOLT PRE'.0A0
' i
~~
TEMPERATURE = 70 F 0 I i -.~~ . '; 3 y i 4 4- ---- -- -- ! t 0 O 100 200 300 400 500 600 RPV Metal Temperature ( F) Figure 7-2, Components of Operating Limits curve for l Non-Nuclear Heatup/Cooldown (Curve B) 7-10
*I i ' i* . ? i I
!! i , ! ! :4'
~ , UNIRRADIATED H
~~
i :
! ' i s BELTLINE REGION , , ! ,. i .
/i '
(RTNDY " *0 II 11
! ./ I
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f i
, N l \
i \ l ! 3 l I i k O i J h i N * . 3
~
k \ \ M I 1 m C O U N e x 4C E e4 4 cm
- 5 a h (
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/ 5 U Y z / > u N
E " E s g O " E
\ > $
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\ w
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, a T
+ <
\
N\ j x _ C m O O O O O O O O O c n t M N w e N I i (.:l} 3Bn1YW3dW313DN3W3.43W 031Snr0Y 7-12
a 4 j . , , . . i e t . # * , * . ., ; l
;. ; . ,, ; !, , , ,, VALID 10 32 EfrY
, i . ., , i ,, ,,
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. ,i ,.
1400 , I
- . , . . ,7 , , ,
i it i > EELTLINE 1/4 T TLAW
' ' 1 WITH 32 EFPY SHIFT PER 1.99 REY. 2
' ' ,' 15LESSl.!MITING THAN NON. BELTLINE l ! l' ll i
l , ' i -t
~
1200 l !
, , 1 4 , e
' i i I . I 1
I _ t r___ ! l t I
- * ! t a 1 I _. ,,
i
- n
.x o
,
- f I t 1
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.I I ' 7 $AFE I 1000 / ' OPERATING REGION l ~ 1 _ T *
.- =
n i ,i _ , 2
.i
( .-- C f - E i t 800 /,_ _.._. f x, NON-BELTLINE c / FW N0ZZLE LIMITS, ~
- 0
/ 1/4 i FLAW, RTNOT s $2 F e /
e / . _ ---. _ _ _ _ 7 -
.. -,/
600 ,
/ - _ - ,
e ._, _ ___ ._ t v. 0 ' g _ , 400 ' 4 _
, A.
- 312 PSIG '
i . e I $ l - 200 i _. .
~4- -
~'
BOLT PRELOAL TEMPERAftlRE = 70 F 0 0 i l i
~'"
[. FLANGE RFGION, RTHDT = 10 F ~ !
}
9 '- i , , i
, - _. 3, .- .
0 600 100 200 300 400 500 0 RPV Metal Temperature (OT) Figure 7-5. Minimum Temperature for Pressure Tests Such as Required by Section XI 7-13
._.m
b
, , , , . . ,, ii .,i 1 .. ... 1 . , i.., ;
a * ' ' '
. . 1 . l4
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. 4 i. I ! I f ! I i
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. , i .,i 1400 , r l l l ! ,
BELTLINE 1/4 T FLAW WITH 32 EFPY SHIFT REV. 2 i i PER1.99(IMITING IS LE$$ l THAN NON.tCLTLINg
, l j , ,
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5 l e 1 { 1200 , , , , , ,
~
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i
- SAFE '
1000 ' OPERATING ' REGION
=
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/ -- -- -- ..--
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- c. . I x g_ -_.
, ~
'- I $ NON. BELTLINE -
' -Z 400 - / --
FW N0ZZLE LIMITS, 0
, 1/4 T FLAW, RTNOT = 52 F ,
i i / . I r
,I ,
I ' . I ' 3
- / 8
/ '
- t 200 j
- .- . - . . . ,=- == :
. , . / '
' 'f-
_.'. BOLT PRELOAD -- 0 ' TEMPERATURE = 70 F 8 f( -/--- - FLANGE REGION, RTNDT = 100 F .~_ l
+ -~1 l .
7.' , 0 200 300 400 500 600 O 100 PJV Metal Temperature'( F) Figure 7-6. Minin.am Temperature for Mechanical Heatup or Cooldown Following Nuclear Shutdown 7-14
.es-+eTTw--v=Pum-'T'-+7' tem-*NMNS-gr r -m sty I'"T-* **"'**M'I 1 *"'ed"*'"IM*" 'PN- TT 7E-4P- t'tc't'"P-P*i+P""----C t'
- _ - _ - _ _ _ . . _ - . . - - - . _ . _ . - . -. - ~_ - .
i l l I
. . i W-. . .' j l' 1 .
l EN i 1 VALID TO 32 EFPY e 9 ! I I i 4 l t f l .
. . . . , , . . i
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. s
' t
, ' BELTLINE 1/4 T FLLW
' WITH 32 EFPY $HIFT
. , PER 1 REV. 2
- l ,
15 LE 9[IMITING i I I THAN NON. BELTLINE i i i 1200 i ' ' ' i - i I ., , i e l 1
;_ I i
. i t
i *
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- -4
~
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1000 ' OPERATING
- REGION --
g - y c. O i . b _
~
t EM ..__ . _ _ . . _ u _ . _ .
' j
_e p --- _ .
,y ._. q ..__ _ ._. _ _ _ . _ _ _ _ _ _ _ _ . _ .
a .. _ . _ . . ._ .. -._ e. 600 _-_ 4 _.. -
, ,i _ ._
~ > _N .
' NON. BELTLINE FW N0ZZLE LIMITS, 0
. / 1/4 T FLAW, RTNDT s 52 F e -
I I
- 4 i ~~
I f ~
/ ,
j ,
~ ~
- i / t
~
f -l _ _ 200 ._...- j _. m_ _ ,
, f, ; . . _ . ,
~7
~~ ~
BOLT PRELOAD 0
~~~~~
j i (- - TEMPERATURE = 70 F ' i 0 ---- s/~d s c, .
. FLANGE REGION, RTNOT = 10 f -
-- i 1
0 O 100 1200 300 400 500 600 RPV Metal Temperature (OF) l Figure 7-7. Minimum Temperature for Core Operation (Criticality)
'7-15
? 4 1 2 ) l ! 8. REFERENCES
- 1. " Fracture Toughness Requirements," Appendix C to Part 50 of Title 10 of the Code of Federal Regulations, July 1983.
1 { 2. " Protection Against Non Ductile Failure," Appendix C to Section III of l the ASME Boiler & Pressure Vessel Code, Addenda to and including , j Winter 1987. 1
- 3. " Reactor Vessel Material Surveillance Program Requirements,"
I Appendix H to Part 50 of Title 10 of the Code of Federal Re;;ulations, l July 1983. l 4. "Condutting Surveillance Tests for Light Water Cooled Nuclear Power Reacte Vessels," Annual Book of ASTM Standards, E185 82, July 1982. i
- 5. " Radiation Embrittlement of Reactor Vessel Materials," USNRC l ReSulatory Guide 1.99, Draf t Revision, August 1987.
- 6. " Fracture Toughness Requirements," USNRC Branch Technical Position MTEL $+2, Revision 1, July 1981, i
- 7. " Capsule Basket," General Electric Drawing 117C4048, Revision 1, October 1969. '
- 8. Schmidt, C. L., " Test Results from Electroslag Weld Prolongations -
610 0139 Units 1 and 2," Babcock & Wilcox Manufacturing Engineering Report No. 1, October 1968,
- 9. Hodge, J. M. , " Properties of Heavy Section Nuclear Reactor Steela,"
Welding Research Council Bulletin 217, July 1976. 81 _ -._.__._,._...____.u..,_.__._ ._._.__...._,__.,_,__.___..___.._;_._,_..
- - - . - _ - _ . . - _ . _ = _ - _ _ - _ - _ _
- 10. " Reactor Pressure Vessel," CE Purchase Specification 21A1111, i Revision 9, August 1970. !
- 11. " Standard Methods for Notched Bar impact Testing of Metallic Materia".s," Annual Book of ASTM Standards, E23 82, March 1982.
- 12. "Nucioar Plant Irradiated Steel Handbook," EPRI Report NP 4797, September 1986. ;
- 13. " Standard Methods of Tension Testing of Metallic Materials," Annual Book of ASTM Standards, E8 81.
- 14. " Ultrasonic Examination for Cracks in the Top Hoad Plange," CBI Huclear, Development Report 74-9047, December 1975.
- 15. " Ultrasonic Examination for Cracks in the Shell Flange," CL' Nuclear, Development Report 74 9056, November 1975.
t T M 8-2
, , . . . . . . . . . . . . - - - - , . . . - , - . , - - - - . . - , . - , - , - . - - - - . . - - - . , , - . , - ~ , ~ , , . -,~...- ...
APPENDIX A DETERMINATION (d ELECTROSLAG VEL.D PROPERTIES Babcock & Wilcox performed testing on nine electroslag wold prolongations from the Peach Bottom 2 ves.nel, all made from the same weld heat. The data are evaluated in nis appendix to determine conservative values of weld chenistry and initial RTNDT representative of the welds. A.1 ELECTROSLAG VELD CHEMISTRY The weld chemistry analyses are summarized in Table A-1. The average f element values were used as representative values, except for copper and nickel, where the maximum recorded values were assumed. A.2 ELECTROSIAG WELD RTHDT Nine sets of Charpy data from the prolongation tests for Peach Bottom 2 welds are plotted in Figures A 1 through A 9, The testing was not done with the objective of definin6 a full Charpy curve, so upper shelf values are not available. The data sets vece fit with the function Energy - A + B
- TANH [ (T - To) / C ), where l T - test temperature, 'F A, B, C and To are constants.
The highest and lowest Charpy ener8 1 er recorded were 99 f t lb and 2 f t lb, i respectively. For purposes of curve fitting, these were assumed to be the upper and lower shelf-values, thus setting the values of A and B. The values of C and To were determined by nonlinear regression to provide a best fit to the transition regions of the curves. l l A-1 l
. - - . -. - -. - - _ .._... -. . m-,- . . - _ . - - . - _ - -
i i The TANH curve fit values of 50 ft-lb are tabulated 'in Table A 2. These values were averaged to_ arrive'at the mean RTy ' valt i of 45'F. Determination of the standard deviation is shown in Tabte A 2 ae vell. t i 1 4 4 A
}
i, 4 f f A-2 t %; , _ , . , . . . , , - , - , . . . , ~ , , . , . - , - - - ,...... ,, ~ ,... - -,- .. -,,. - ..-- --- --- - - - ~ ~ - - - , - - - - - - - ~ - - ~ - -
. - . . - . . .. - - - , . .- - - ~ = . - - - - . - . .. . ..
.i
'l J
Table A-1 1 1 ELECTROSLAG WELD CHEMISTRY TEST RESULTS j i i Test i Veld No. C A P S 1 ,_C,I ,J1 32- A V l i
- 1 .18 1.40 .014 .012 .09 .05 .21 .53 .16 .01 2 .17 1.43 .016 .012 .10 06 .21 .54 .19 .01 f
3 .17 1.41 .014 .012 .08 .05 .19 .53 .20 .01 i
; 4 .16 1.37 .015 .011 .08 .04 .14 ,51 .21 .01
! 5 .18 1.41 .015 .012 .09 .04 .16 .52 .16 .01 ! 6 .17 1,38 .014 . 01?. .08 .04 .16 .53 .19 .01 7 .16 1.43 .015 .016 .09 .06 .19 .53- .18 - i 8 .16 1.44 .016 .015 .09 .05 .19 .54 .17 - 9 .17 1.45 .014 .014 .09 .05 .19 .54 .19 . i i Average .17 1.41 .015 .013 .09 .05 .18 .53 .18 .01 i Representativa < Ohemistry: 1
- c .En._ P S J.i __CI _Ei A A V
.17 1.41 .015 _013 .09 .05 .21 .53 .21 .01-4
[ 1 I A-3
- l. -. . . , . - - .... - . . . . - - - . . . -. - , - - . . . . . - - , . , - .. - ,--, . - . . .
.. . . _ _ . . _ . . _ .~ ..._._ _ . - . _ ._.__
Table A-2 DETERMINATION OF ELECTROSLAG VELD RTNDT Test Veld No. Index Temperature T$o Variation (730 - T $o)2 1 18.8 11.51 2 3.0 155.45 3 31.7 265.76 4 32.1 277.63 5 26.7 126.39 6 -20.7 1303.77 7 15.9 0.20 8 12.5 8.40 9 18.9 11,78 Sum 138.9 2160.89 Mean: T - 138.9 / 9 - 15.4'F, so SO RT ( 50 - 60*F) - 45'F
~
NDT Standard Deviation: 9 1/2 a7 - [ (1/8)
- S1(T50 ' 50)2 I ay - 16.44"F A-4 i
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... . . ~ . - - . - . . . . _ -
N i i APPENDIX B IRRADIATED CHARPY V NOTCH FRACTLCE SURFACE PHOTOCRAPHS' l l-
- Photographs of each Charpy specimen fracture surface were taken to facilitate the determination of percent shear, and to comply with the , requirements of ASTM E185-82. The pages following show the fracture 3 sutface photographs along with a summary of the Charpy test results for each specimen. The pictures are arranged by increasing test temperature for each material, with the :naterials in the order of base, weld and HAZ.
l ? l i I i i l 4 i h 5 I E 1 i a i i B-1 i t c,m. ..,, . , , ,. ,,----,-w--- ,.,e--w..,..+-,,- m- .-y-.w . ,,.e .*.,...-n --- .y- ,a ..-- .-_ .- . , , . -.-,,.s.+.,--w.sr-., w _.-- ~,- .
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BASE: 772 MLE: 24 TEMP: - 20*F % SHEAR: 6 ENERGY: 21.0 fr-lb l l 1 l
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I BASE: 765 MLE: 38 TEMP: O'F % SHEAR: 18 ENERGY: 47.0 ft-lb l l l B-2 l
-- .- ..m . _ . . _ - . . _ . _ _ _ , .__.,.._.m. -,._,-._...__-___....-.......__.~..-._.-_m. - - . .
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APPENDIX C l UNIRRADIATED CHARPY V-NOTCH FRACTUPI SURFACE PHOTOCRAPHS l Photographs of each Charpy specimen fracture surface were taken per l I the requirements of ASTM E185-82.' The pages following show the fracture j surface photographs along with a summary of the Charpy test results for l each specimen. The pictures are arranged by increasing test temperature l
- for each material, with the r.aterials in the order of base and weld. l I
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l Enclosure 2 to Attachment A
~
\ -
L FILE i V June 27, 1990 Docket No. 50-278 License No. DPR-56 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
Reactor Vessel Specimen Surveillance Test Results for Peach Bottom Atomic Power Station, Unit 3 Dear Sirt Attached is the report entitled " Peach Bottom Atomic Power __ Rtat_i,oAUrd t _ a_,,Ves sel _Su rveillance Ma t erd als- Tes ti ng..and --Fractu r e- - - - " - ' Toughness Analysis", which was prepared for the Philadelphia Electric Company by the General Electric Company. This report is submitted to satisfy 10 CFR 50 Appendix H, " Reactor Vessel Paterial Surveillance Program Requirements," which requires the submittal of the test results within one year of the withdrawal of the specimen capsule. The specimen capsule was removed from the Unit 3 reactor vessel on June 28, 1989. An Application of amendment to revise the Unit 3 Thermal and Pressurl=ation March 29, 1991 Limitations, Sections 3/4.6, will be submitted by to reflect the test results. If you have any questions, please do not hesitate to contact us. Very truly yours,
. h. , 1, G. A. Bunge , r.
Manager-Licensing Nuclear Engineering & Services ' TRL:lc Attachment cc: T. T. . Martin, Administrator, Region I, USNRC J. J, Lyash, USNRC Senior Resident Inspector
^^
y
a g SASR 90 $0 DRF B11 00494 June-1990 PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 VESSEL SURVEILIANCE MATERIALS TESTING AND FRACTURE TOUGHNESS ANALYSIS ( Prepared by: a a e T. A. Caine, Senior Engineer Materials Monitoring-& ~ Structural Analysis Services A
.\
Verified by: 9 h 1 %h, I ' B. J. Branlund, Engineer-Materials Monitoring & ( Structural Analysis _ Services t
..e.
7
-Approved-by: /\
a -# J
// -
-l S. Ranganath, Manager }
Materials Monitoring & Structural Analysis Services' - 9 - GENuclearEnergy .g o,n. can=;n m .a w !M_pp -
' _-____ - __-_-__- - - - - - _n. _
e IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the use of Philadelphia Electric Company. The information contained in this report is believed by Cerietal Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was preparet. The only undertakings of the General Electric Company respecting information in this document are contained in the contract between the customer,and General Electric Company, as identified in the purchase order for this report and nothing containe'd in this document shall be construed as changing the contract. The use of this information by anyone ether than the customer or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized ute, General Electric Cor pany makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
, 11 1
_ _ _ _ _ _ _ _ _ - _ - _ - _ - _ _ _ . - - _ _ _ _ -- _ 1
CONTENTS EALE ABSTRACT Vili ACKN0'JLEDGMENTS IX
- 1. INTRODUCTION 1-1
- 2. SUMHARY AND CONCLUSIONS 21 2.1 Summary of Results 21 2.2 Conclusions 25
- 3. SURVEILLANCE PROGRAM BACKCROUND 31 3.1 Capsule Recover;* 31 3.2 RPV Materials and Fabrication Background 3-1 3.2.1 Fabrication History 3-1
. 3.2.2 Material Properties of RPV at Fabrication 32 3.2.3 Specimen Chemical Composition 32 3.2.4 Initial Reference Temperature 3-3 3.3 Specimen Description 3-5 3.3.1 Charpy Specimens 3-5 3.3.2 Tensile Specimens 36
- 4. PEAK RPV FLUENCE EVALUATION 41 4.1- Flux Vire Analysis 41 4.1.1 Procedure 41 4.1.2 Results 42 4.2 Determination of Lead Factors 43 4.2.1 Procedure 4-3 4.2.2 Results .
44 4.3 Estimate of 32 EFPY Fluence 45
- 5. Cl!ARPY V-NOTCH IMPACT TESTING 51 5.1 Impact Test Procedure 5-1 5.2 Impact Test Resulta 52 5.3 Irradiated Versus Unirradiated Cherpy V Notch Properties 53 5.4 Comparison to Predicted Shift 54
- 6. TENSILE TESTING 61 6.1 Procedure 6-1 6.2 Results 6-2 6.3 Irradiated Versus Unirradiated Tensile Properties 62
- 7. DEVELOPMENT OF OPERATING LIMITS C' JVES 7-1 7.1 Background " "' 7-1 7.2 Non Beltline Regions 71 7.3 Core Beltline Region 72 7.4 Closure Flange Region 7-2 7.5 Core Critical Operation Requirements of 10CFR50, Appendix G 7-3 iii
CONTENTS EALt 7.6 Evaluation of Radiation Effects -74 7.6.1 Heasured Versus Predicted Surveillance Shif t 75 7.6.2 ART Versus ETPY 75 7.6.3 Fracture Toughness Conditions at 32 EFPY 75 7.7 Operating Litnits Curves Valid to 32 EFPY , , 76 7.8 Reactor Operation Versus Operating Lirnits 76
- 8. RETERDiCES 81 APPENDICES A. IRRADIATED CHARPY SPECIMEN FRACTURE SURFACE PHOTOCRAPHS A1 B. UNIRRADIATED CHARPY SPECIMD1 TRACTURE SURFACE PHOTOCPAPHS B1 C. BASIS FOR CONSERVATIVE RT C1 NDT 6
..<s.
4
- iv i
TABLES Table Title Eagt 3-1 Chemical Composition of RPV Beltline Materials 38 32 Mechanical Properties of Beltline and Other 39 Selected RPV Materials 3-3 Che~mical Composition of Irradiated Surveillance - 3 10 Specimens 34 Identification of Unit 3 Surveillance Specimens 3 11 41 Summary of Daily Power History 46 42 Surveillance Capsule Plux and Fluence for 47 Irradiation from 9/1/74 to 3/31/87 51 Vallecitos Riehle Charpy Machine Qualification Test Results 5-5 Using NIST Standard Reference Specimens (Tested 3/5/90) 52 Charpy V Notch Impact Test Results for Irradiated RPV 56 Materials Unit 3 .. 53 San Jose Tinius Olsen Charpy Machine Qualification Test 57 Results Using NIST Standard Reference Specimens (Tested 5/14/90) 54 Charpy V Notch Ig t Test Results for Unirradiated RPV 5-8 Materials - Unit 3 54 Significant Results of Irradiated and Unitradiated Charpy 5-9
- v. Notch Data for Unit 3 61 Tensile Test Results for Irradiated RPV Materials for Unit 3 6 3 62 Comparison of Unitradiated and Irradiated 64 Tensile Properties at Room Temperature for Unit 3 71 Beltline Evaluation for Peach Bottom 3 7-7 72 Estimate of Upper Shelf Energy for Beltline Materials 78
.~'.
V 1 ___________J
ILUUSTRATIONS Firure Titic Engg 21 Pressure Temperature curves Valid for 32 EFPY of Operation 26 31 Schematic of Identification on Basket Recovered from Unit 3 3 12 32 Schematic of the RPV Showing Identification 3 13
* ~
of Vessel Beltline Plates and Welds 33 Fabrication Method for Base Metal Charpy Specimens 3 14 34 Fabrication Method for Veld Metal Charpy Specimens 3 15 35 Fabrication Method for HAZ Charpy Specimens 3 16 36 Fabrication Method for Base Metal Tensile Specimens 3 17
^
3-7 Fabrication Method for Veld Metal Tensile Specimens 3 18 38 Fabrication Method for HAZ Tensile Spccimens 3 19 41 Schematic of Model for Two Dimen;ional Flux Distribution- 48 Analysis 51 irradiated Base Metal Impact Energy 5 10 52 Irradiated Veld Metal Impact Energy 5 11 53 Irradiated HAZ Impact Energy 5 12 54 Irradiated Base Metal Lateral Expansion 5 13 55 2rradiated Weld Metal Lateral Expansion 5 14 56 Irradiated HAZ Lateral Expansion 5 15 57 Comparison of Unitradiated and Irradiated 5 16 Ease Metal Impact Energy 5-8 Comparison of Unirradiated and Irradiated 5 17 Veld Metal Impact Energy 5-9 Comparison of Unirradiated and Irradiated 5 18 Base Metal Lateral Expansion ,,,,, 5-10 Comparison of Unirradiated and Irradiated 5 19 Weld Metal Lateral Expansion vi
ILLUSTRATIONS Figure Title Eart 61 Typical Engineering Stress Straia for Irradiated 6-$ RPV Materials - 62 Strength vs. Temperature for Irradiated Materials 66 63 Due'tility vs. Temperature for Irradiated Materials 67 64 Fracture location, Necking Behavior and Fracture 68 Appearance for Irradiated Base Metal Tensile dpecimens 65 Fracture tocation, Necking Behavior and Fracture 69 Appearance for Irradiated Weld Metal Tensile Specimens 66 Fracture Location, Hecking Behavior and Fracture 6 10
, Appearance for Irrad.inted HAZ Tensile Specimens 71 ART of Limiting Beltline Material 79 72 Pressure Temperature Curves Valid for 32 EFPY of Operation 7 10 1 4
e4 4'e vii
e ABSTRACT A surveillance capsule was removed from the peach Botton Unit 3 reactor during the outage following fuel cycle 7. The capsule contained flux vires for neutron fluence measurement and Cherpy and tensile test specimens for material property evaluation. Flux vire testing and the lead factor from pr'vious e computer analysis were used to establisti the vessel peak flux location and magnitude. Charpy V Notch impact testing and uniaxial tensile = sting veru performed to establish the properties of the irradiated survellaance materials. The irradiation effects were projected, based on Regulatory Guide 1.99, Revision 2, to conditions for 32 effective full power years (D tY) of operation. The 32 EFPY conditions are predicted to be less severe than the limits in 10CFR50 Appendix C requiring provisions for vessel thermal annealing. Pressure temperature sperating limits curves valid to 32 EFpY vere developed to 10CFR$0 Appendix C requirements, accounting for irradiation shift per Regulatory Guide 1.99, Revision 2. In conjunction with the surveillance capsule testing, unirradiated plate and weld Charpy specimens were tested to establish baseline curves. The irradiated Charpy data for the Unit 3 plate and weld specimens were compared to these unirradiated data to determine the shift in Charpy curves due to irradiation. The results are just above mean shift predictions, but below the upper bound predictions-of the Regulatory Guide. g& 4 viii
I-ACKNOWLEDCMENTS The author gratefully acknowledges the efforts of the people mentioned below. . Flux wire testing was performed by G. C. Hartin. Charpy testing was-completed by G. P. Vozadio and G. E. Dunning. Tensile specimen testing vat done by S. B. Visner and G. H. Henderson, and chemical cotcposition analysis was performed by C. R. Judd.
..u.
D
. ix
4
- 1. INTRODUCTION Part of the effort to assure reactor vessel integrity involves evaluation of the fracture toughness of the vessel ferritic materials. The '
key values which characterize a material's fracture toughness are the reference temperature of nil ductility transition (RTNDT) and the upper shelf energy (USE). These are defined in 10CTR50 Appendix C (Reference 1) and in Appendix 0 of the ASME Boiler and Pressure Vessei Code, Section XI (Reference 2). These documents contain requirements used to establish the pressure temperature operating limits which must be met to avoid brittle fracture. Appendix H of 10CFR50 (Reference 3) and ASTM E185 (Reference 4) establish the methods to be used' for sut .eillance of the reactor vessel materials. The first vessel surveillance specimen capsule required by Reference 3 was removed from Unit 3 in June 1989. The capsule was sent to the CE Vallecitos Nuclear Center (VNC) for testing after exposure to seven fuel cycles of irradiation. The surveillance capsule contained flux wires for neutron flux monitoring and Charpy V Notch impact test specimens and uniaxial tensile test specimens f abricated from materials from, or representative of, the vessel materials nearest the core (beltline). The impact and tensile specimens were tested to establish properties for the irradiated materials. The results of the surveillance specimen testin5 are presented in this report. The irradiated material properties are compared to unirradiated properties from testing of archive specimens, also presented in this report. Predictions of the RTNDT end USE at 32 EFPY are made for compariso'n with allowable values in Reference 1. Predictions of 32 EFPY properties vert made based on Regulatory Guide 1.99, Revision 2 (Reference 5). . . - . I-
, 1-1
4 Operating limits curves for the Unit 3 reactor vessel are presented in this report. The curves account for current requirements of References 1 and 2. Geometric discontinuities and highly stressed regions, such as the feedwater nozzles and the closure flanges, are evaluated separately from the core beltline region. The operating limits developed consider the most limiting conditions of the discontinuity regions and the boltline region, so as to bound' all operating conditions. The operating limits developed for the beltline region include irradiation shift, based on Reference 5 methods. 4 1
<gk b e
- 1-2
i ! i
- 2.
SUMMARY
AND CONCLUSIONS 4 l
- 2.1
SUMMARY
OF RESULTS i , l t j Surveillance capsule 1 was removed from Peach Bottom Unit 3 during the ! outage following Fuel Cycle 7 end shipped to VNC. The flux vires, Charpy V Notch and tensile test specimens removed from the capsule.were tested ]* according to ASTM E185-82 (Reference 4). Revised operating limits curves
- vere developed usinb the flux vire test results and the requirements and l methods of 10CFR50 Appendix 0 (Reference 1) and Appendix C of ASME Code 1
- Section XI (Reference 2). The methods and results of the fracture
; toughness evalaation are presented in this report as follows:
r
- a. Section 3: Surveillahce Program Background
-l i
- s. Section 4: Peak RPV Fluence Evaluation
) c. Section 5: Charpy V Notch Impact Testing [ d. Section 6: Tensile Testing
- e. Section 7: Development of Operating Limits Curves Photographs of fractured Charpy specimens are in Appendix A. The significant results of the evaluation are below:
4 )
- a. Capsule 1 was removed from the 30' azimuth position of the reactor. The capsule contained 9 flux vires: 3 copper (Cu),
3 1ron (Fe), and 3 nickel (Ni). There were 36 Charpy V Notch specimens in the capsule: 12 each of plate material, veld material and heat affected zone (HAZ) material. The 8 tensile .. '. 4 specimens removed consisted of 3 plate, 2 veld and 3 HAZ metal specimens. 2-1 me - - - - , ,.. - _ , , , _ _ _ , . , , _ _ , , . _ _ , _ _~ _
I
- b. The chemical compositions of the beltline materials were determined from data obtained from CE QA records or from correspondence with Babcock 6 Wilcox (B&W) The copper (Cu) snd nickel (Ni) contents were determined for al
- eats of plate and 1 weld material. The values for the limit u 'altline plate ar-0.15% Cu and 0.49% Ni. The limiting beltlans weld values are 1
0.21% Cu and 0.21% Ni. .
- c. Charpy and dropweight test results from the fabrication program materials certification testing were adjusted to be equivalent to test results done to current standards. The initial RTNDT values for locations of interest in l.he vessel were determined. They are 10*F for the limiting beltline plate, -45'F for the limiting beltline weld (with ai - 16.44*F), 10*F for the closure flange region, 50'T for the limiting nozzle and 54'T for the bottom head region.
- d. The flux wires were tested to determine the neutron flux at the surveillance capsule location. The fast flux (>1.0 MeV) measured was 6.8x108 n/cm2 sec. Based on the flux wire data, the surveillance specimens received a best estimate flusnce of l 1.6x1017 n/cm2 at removal.
- e. The vessel inside surface lead factors were established using an analysis performed for Peach Bottom Unit 2 that combined tvo dimensional and one dimensional finite difference transport analysis. The flux peak occurs at an azimuthal location 25.5' from the vessel quadrant references. The lead factor for the surveillance capsules is 0.95 to the peak vessel inside surface location, 1 ....
l 2-2 4 l
l l 4
- f. T'..e maximum accumulated neutron fluence at 32 EFPY vas determined 1
at the peak 1/4 T location, using the flux vire test results, the J ] ' lead I.: *.or and the methods of Reference 5. The maximum 1/4 7 $ vessel 32 ETPY fluence is 5.0x1017 n/cm2 ,
- g. The surveillance Charpy V Notch specimens were impact tested at temperatures selected to define the tranrtition of the fracture 1
toughness curves of the plate, veld, and MAZ materials. Measurements were taken of abs. orbed energy, lateral expansion and
! percentage shear. Fracture surface photographs of each specimen
- are presented in Appendix A. From absorbed energy and lateral i
expansion results for the plate and weld materials the following !] values were calculated: index temperatures for 30 f t lb , 50 ft-lb, and 35 mil lateral expansion (MLE) values and USE. { Similar test results were obtained for archive plate and veld Charpy specimens provided by PECO. The unirradiated specimen
- fracture surface photographs are presented in Appendix B.
i i h. The curves of irradiated plate and veld specimen Charpy impact ] energies were compared to the corresponding unitradiated data to establish the ~ 30 f t-lb, 50 f t lb and 35 MLE index temperature 3 irradiation shifts, and decreases in USE. 'Ihe surveillance plate material showed an estimated 16"F shif t and a 10% decrease in i USE. The veld material estimated shift was also 16'F, with a 1' decrease in USE of 9%. l j i. The irradiated tensile specimens were tested at room temperature (70*F), reactor operating temperature ( 5 5 0 ' F) , and estimated onset to upper shelf temperature (160*F). The.results tabulated ! for each specimen include yield and ultimate tensile strength, 5 uniform and total elongation, and reduction of area.
- j. The irradiated plate and weld tensile test results for Unit 3 4
were compared to unirradiated data from the vessel fabrication- . test. program records. The results generally showed increasing ^ strength and decreasing ductility, consistent with expectations for irradiation embrittlement. 2-3
.,,,,....,-.-,.----m.,. ,- -%
/ a . i 5
} k. An a part of the development of the pressure temperature (P.7) operating limits curves, the irradiation shifts in RTHDT were l
i predicted, based on the methods of Regulatory Guide 1.99, f Revision 2 (Reference 5). For information purposes, the measured
+
l shifts for the surveillance plate and weld materials were
! compared to the shif ts predicted by Reference 5. The measured i
j shift of 16*F in the plate material 30 ft-lb index, for a fluence 3 of 1.6x1017 n/em2, is just above the predicted mean shift of 4 j 14'F, but is well below the upper limit shift prediction of 28'F. For the veld material, the measured shift of 16'r is slightly i Ligher than the predicted mean shif t of 15'T and well below the I upper bound of 30'F. i i 1. Data for prediction of most beltline naterial USE values were not i available. Data for the surveillance materials and the ] longitudinal beltline welds were used to pre dict USE at 32 EFPY using the methods in Reference 5. The beltline transverse plate l and weld USE values were predicted to be 78 ft lb and 82 ft lb, ! respectively, at 32 ETPY. * ? i 1 l m. P-T curves were developed for three reactor conditions: hydrostatic pressure test (Curve A), non nuclear heatup and 4 l cooldovn (Curve B), and core critical operation (Curve C), The curves are valid up to 32 EFPY of operation. The limiting l regions of the vessel affecting the curves' shapes are the nozzle, bottom head and closure flange regions. The bolt preload and minimum permissible operating temperatures were determined to 4 ~ be 7 0
- F. The predicted irradiation shifts for-the Unit 3 i
- beltline materials are low enough that the beltline is not -
predicted to be limiting through 32 EFPY of operation. The P T-j curves for Unit 3 are shown in Figure 2 1.
..~'.
I 2 ts
-w,wv---wwr,-- -e- ,ew.,r -e- -, * + , - - -w,v.-e. ,,-y--%,s--rwi,c
+--w-v.r- .-+ y-c ewey- . . += - . , - ---r-- + - - - - - '
2.2 CONCLUSION
S The requirements of Reference 1 deal basically with vessel design life conditions and with .imits of operation designed to prevent brittle fracture. Based on the evalu.ition of surveillance testing results, and the
- : soc! ated analyses, the following conc 2usions are made:
- a. The adjusted reference temperatures at 32 EFFY for the limiting beltline material of 71'F is below the Reference 1 allouable e linit of 200*F, above which specic.1 analyses or provisions for arnealing are required.
- b. ,
The 32 EFPY values of USE could not be calculated for all beltline plates, but 'the 32 EFPY USE for the surveillance beltline plate, 78 ft-lb, and the lowest USE predicted for beltline velds, 82 ft-lb, are well above the Reference 1 allowable minimum of 50 ft-lb. The surveillance plate USE should be representative of the other beltline plates' USE values, so there is no need, based on USE values, for special analyses or provisions for annealing the Unit 3 vessel beltline,
- c. Examination of the normal and upset operating conditions expected for the reactor shows that the worst pressure-temperature conditions expected from unplanned temperature transients are acceptable relative to the limits in Figure 2 1. Therefore, the only operating conditions for which the operating limir T aa concern are those involving operator interaction, su a as pressure testing and initiation of core criticality.
, 2-5
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j WITH FUEL IN = VESSEL f - B,B' - NON-NUCLEAR HEATUP/ b j ' COOLDOWN LIMIT y _. J C,C* - NUCLEAR (CORE CRITICAL) LIMIT w e 400- VESSEL DISCONTINUITY
$ LIMITS
- l. g 312 s - - - SHIFTED BELTLINE LIMITS x / f CURVES A,0.C ARE VAUD
- c. ,
[ FOR 32 EFPY OF OPERATION
-200 i BOLTUP l 70'r CURVES A',B',C' FOR BELTLINE -
ARE LESS UMITING THAN .
/['
A,B,C AND ARE FOR INFO ONLY ' s 0 I i 1 0 100 .200- .300 400 . 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE ( F) Figure 2-1. Pressure-Temperature Curves Volid for 32 EFPY of Operation 2-6 Y y- ) 7 - y- 'n-19.'F"'FFTf'W'-T-t 4=
-- t v i rw t *Wre-'W '+w e9 ' r*4 + 9'y'9=w+ 54P+ m -Maa re-ve.> -M- W u eT-e*'*t- v ieer+W- =>++=evee d>M,c awmer---**-t et9 d t-4 8e*Wm +cr .9+we At+8e = e- m e e -s w+st- em e wm-,&+'
4
- 3. SURVEILLANCE PROGRAM BACKGROUND
' 3.1 CAPSULE RECOVERY + The Peach Bottom Unit 3 reactor was shut down in March, 1987 . The accumulated thermal power output was 9.11x106 mwd or 7.57 EFPY. The-I reactor pressure vessel (RPV) originally contained three surveillance capsules, at 30*, 120' and 300* azimuchs at the core midplane. The specimen capsules are held against the RPV inside surface by a spring loaded specimen holder. Each capsule receives equal irradiation because of core symmetry. In June 1989, Capsule 1 at 30' was removed. The capsule i was cut from its holder assembly and shipped by cask to the GE Vallecitos Nuclear Qenter (VNC), whero testing was performed. 1 Upon arrival at VNC, the capsule was examined for identification. The drilled hole binary code e che basket showed the reactor number of 27 and the capsule number of 1, as shown in Figure 3-1. The capsule contained three Charpy specimen packets and four tensile specimen tubes. Each l tensile specimen tube contained two tensile specimens. Each Charpy ! specimen packet contained 12 plate, weld or HAZ Charpy specimens and 3 ilux wires (one iron, one copper and one nici.el) in a sealed helium environment. 3.2 RPV MATERIALS AND FABRICATION BACKGROUND j 3.2.1 Fabrication History The Peach Bottom 3 RPV is a 251 inch diameter BWR/4 design. Construction was begun by B&W to the Winter 1965 Addenda of the 1965 edition of the ASME Code. Chicago Bridge & Iron (CB&l) completed 'the vessel, generally to the 1968 edition of the ASME Code. The shell and head plate materials are ASME SA302, Grade B, low alloy steel (LAS), modified ,. per Code Case 1339. The nozzles and closure flanges are ASTM A508 Class 2 LAS, modified per Code Case 1332-3, and the closure flange bolting materials are ASTM A540 Grade B23. The fabrication process employed double quench and temper heat treatment immediately after hot forming, then electroslag or submerged are welding and post-weld heat treatment. The 3-1
. _ . . _ _ __ .A _ . . _ _
4 i 4 i
, post-weld heat treatment was typically for 30 hours at temperatures of .
1125.1 25'F. The identification of pintes and welds in the beltline region is shown in Figure 3-2. l 7 ! 3.2.2 Material Properties of RPV at Fabrication i j Material ~certifir tion records were re trieved f rom -GE Quality l Assurance (QA) records to determine chemical and mechanical properties of the vessel materials. m addition, data on the electroslag welds was ! obtained from the Unit 2 surveillance test report (Reference 6). Table 3 1 1 l shows the chemistry data for the Unit 3 beltline materials. i Results of certification mechanical property tests performed during i' RPV f abrication were examined, specifically Charpy V-Notch and dropweight impact test results. Properties of the beltline materials and other l locations of interest are presented in Table 3-2. The Charpy data collected were used to establish the RTNDT values for each vessel component, as described in Subsection 3.2.4. 3.2.3 Soecimen chemical comnosition l Sampics were taken from the surveillance plate and weld tensile , specimens after they were tested. Chemical analyses were performed using a Spectraspan III plasma emission spectrometer. Each sample was dissolved in an acid solution to a concentration of 40 mg steel per 40 mi solution. The spectrometer was calibrated for determination of Mn, Ni, Mo and Cu-by i diluting National Institute of Standards and Technology (NIST) ! Spectrometric S tandard Solutions . The phosphorus calibration involved analysis of four reference materials from'NIST with known phosphorus l 1evels. Analysis accuracies are 10.003% (absolute) for phosphorus and 15% (relative) for other elements. The chemical composition results are ,given , in Table 3-3 for the Unit 3 surveillance plate and weld materials. The results for the plate show reasonable agreement with corresponding data i from fabrication records in Table 3-1. However, the surveillance weld chemistry results do not correspond to the results for the beltline electroslag weld heat, as would be expected from normal B&W practices. The surveillance weld is, however, representative of the beltline welds. 3 -2 l
, _ - _ . . _ . _ , . . . . , . _ . . . - , _ _ ,. - _ . _ . . _ .- , , , , . - , -- ., . _ , , 7___
- . . . . .- . ., . . - . --. . . . . . ~ ~ - _ - . - .. - ..- - .-
) j + , i 3.2.4 Initial Reference Temneratures The requit ements for establishing the vessel component RTNDT values l per the ASME Code prior to 1972 are summarized as follows: , f a. Test specimens shall be longitudinally oriented Charpy V Notch j specimens. - -
- b. At the RTNDT, no impact test result shall be less than 25 f t lb, i
- and the average of three test results shall be at least 30 ft lb.
I i
- c. Pressuze tests shall be conducted at a temperature at least 60*F l, , above the acceptable RTNDT for the vessel.
The current requirements for establishing RTNDT are significantly j different. For plants constructed to the ASME Code after Summer 1972, the , requirements are as follows: i
- a. Charpy V-Notch specimens shall be oriented nor g1 to the rolling
- direction (transverse).
a
- b. RTNDT is defined as the higher of the dropweight NDT or 60*F 2
l below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion are met. e
- c. Bolt-up in preparation for a pressure test or normal operation j shall be performed at or above the RTNDT or lowest service 2 temperature (LST), whichever is greater.
~
Reference 1 states that for vessels constructed to a version of the I ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses must be supplemented in an approved manner. CE has developed methods for analytically converting fracture toughness data for vessels constructed before 1972 to comply with current-requirements. CE developed these methods from data in WRC Bulletin 217 (Reference 7) and from data collected to respond to NRC questions on FSAR submittals in the late 1970s, i 1 r 3-3 1 y ,,-m, ..w- -- r w - ,.:- - v.e-, m, - , - .-.,-,--y
. -.,-# .,.-,._~.--,--wre.-... w - v us ..,.*#e- -v- . . . - v -
S .s These methods and example RTNDT calculations for vessel plate, veld, veld i HAZ, forging, and bolting material are summarized in the remainder of this ! sub se cti on. Calculated RTNDT values for selected RFV locations are given ! in Table 3-2. I For vessel plate material, the first step in calculating RTNDT is to establish the 50 f t lb transverse test temperature from longit.udinal test l. specimen datt.. There are typically . three energy values at a given test temperature. The lowesu energy Charpy value is adjusted by adding 2*F per I f t-lb energy to 50 f t lb. For example", for the limiting Unit 3 beltline
- plate the test temperature and lowest Charpy energy from Table 3-2 is I 44 ft-lb at +10'F fot- Heat C2773-2. The equivalent 50 f t lb longitudinal test temperature is
)J
; T50L - 10*F + [(50 - 44) ft-lb
- 2*F/ft-lb) - 22*F
+ - The transition from longitudinal data to transverse data is made by adding 30*F to the test temperature. In this case, the 50 ft lb transverse Charpy test temperature is T50T - 52*F. The RTypy_ is the greater of NDT or
- (750T - 60*F). From Table 3-2, the NDT for Heat C2773-2 is 10*F, and (T50T - 60*F) is -8'F. Thus, the RTNDT for that beltline plate is 10'F.
For vessel veld material, the Charpy V-Notch results are usually limiting in establishing RTNDT. The 50 ft-lb test temperature is established as for the plate material, but the 30*F adjustment to convert longitudinal data to transverse data is not applicable to veld metal. In Reference 6, nine sets- of Charpy curves for the electroslag veld heat used in Units 2 and 3 were evaluated to determine (T50T - 60*F) of -45'F. There are no NDT data available for the longitudinal electroslag welds. The GE 1 procedure requires that, when no NDT is available, the resulting RTNDT be
- -50*F or higher. In thi.s example, (T50T - 60'F) is -45'F, so the RTNDT 18
-45'F. Since the RTNDT for the electroslag welds was based on stati's' tic'al analysis of 9 curves, the value of aI discussed in Reference 5 is the standard deviation from the statistical analysis of the 9 curves, or 16.44*F.
3-4
-- -- - v -, e .
For the vessel weld HAZ material, the RTNDT is assumed to be the same as for the base material since ASME Code veld procedure qualification test requirements and post veld heat treatment data indicate this assumption is valid. For vessel forging material, such as nozzles and closurc~ flanges the method for establishing RTNDT is the same as for vessel plate material. For the recirculation inlet nozzles N2, the lowest Charpy data at 40*F is 30 ft-lb, and the NDT is 10*F. In this case, (T50T - 60'F) is greater than NDT, so the RTNDT is (40 + (50 30)*2 + 30 - 60), or 50*F. For bolting material, the current ASME Code requirements define the LST as the temperature at which transverse Charpy V-Notch energy of 45 f t-lb and 25 mils lateral expansi'on (MLE) are achieved. If the required Charpy results are not met, or are not reported, but the Charpy V-Notch energy reported is above 30 f t-lb, the requirements of the ASME Code at construction are applied, namely that the 30 ft-lb test temperature plus 60*F is the lowest service temperature (LST) for the bolting materials. Charpy data for the studs did not meet the 45 f t-lb, 25 MLE requirement, but 30 ft-lb energies were met at 10*F. Therefore, the bolting material LST is 70*F. 3.3 SPECIMDI DESCRIPTION The surveillance capsule contained 36 Charpy specimens: base metal (12), weld metal (12), and HAZ (12). There were 8 tensile specimens: base me ta l-- (3 ) , weld metal (2), and HAZ (3). The capsule contained 9 flur.
' vires: 3 iron, 3 nickel and 3 copper. The chemistry and fabrication history for the Charpy and tensile specimens are described in this section.
3.3.1 charov soecimens ~ ' Charpy specimen fabrication is described in the vessel purchast specification (Reference 8). All specimen materials were specified to be beltline materials, but the chemistry test results indicate that the veld wire heat used is not the same as that in the Unit 3 beltline. It is, however, representative of the beltline electroslag welds. 3-5 l
-- . . -~ - - - . . -. - - . .. - .- . --. ~
The base metal specimens were cut from Heat C3103 1. The surveillance test plates received the same heat treatment as the f abrication specimens j for Heat C3103-1, including the post weld heat treatme.nt for 30 hours at l 1125'F i 25'F. The method used to machine the specimens from the test 4 j . plate is shown in Figure 3*3. Specimens were machined from the 1/4 T and ! 3/4 T positions in the plate, in the longitudinal orientation (long axis - parallel to the rolling direction). The identifications of the base metal i Charpy specimens recovered from the surveillance capsu?e are shown in Table 3 4. l The veld metal and HAZ Charpy specimens were fabricated by welding together two pieces of the surveillance test plate with a weld which was l specified to be identical to the beltline longitudinal seam welds. The ] same veld procedure was used, but'the actual weld wire heat and flux lot l probably were not used in the vessel beltline welds, based on the results i of the specimen chemistry tests. The welded test plate for the weld and HAZ Charry specimens each received stress relief heat treatment at 1125*F i 25'F for 30 hours to simulate the fabrication specimen conditions. The weld specimens and HAZ specimens were fabricated as shown in Figures 3 4 and 3-5, respectively. The base metal orientation in the weld and HAZ specimens was longitudinal. The specimens were stamped on one end with the ! identifications shown in Table 3-4. i 3.3.2 Tensile Snecimens l Fabrication of the surveillance tensile specimens is described in Reference 8. The materials, and thus the chemical compositions and heat i treatments for the base, weld and HAZ tensiles are the same as those for rh'e corresponding Charpy specimens. The identifications of the base, weld, and HAZ tensile specimens are given in Table 3-4. A summary o f the fabrication methods follows. The base metal specimens were machined from material at the 1/4 T and 3/4 T depth. The specimens, oriented along the plate rolling direction, , were machined to the dimensions shown in Figure 3-6. The gage section was l tapered to a minimum diameter of 0.250 inch at the center. The weld metal I-l 3-6
_ . . . - - . . . -. ... --. .~ .- . . - __ 4 4 I tensile specimen materials were cut from the velded test plates, as shown i in Figure 3-7. The specimens were machined entirely from veld metal, 3 scrapping material that might include base metal. The fabrication method 1 for the HAZ tensile specimens is illustrated in Figure 3 8. The specimen blanks were cut f rom the velded test plates such that the Sage sec'tion minimum diameters were machined at the veld fusion line. The finished HAZ specimens are approximately half weld metal and half base met;al oriented l along the plate rolling direction. i 4 j , t i i ( i 4 4 d I k i . s. 4 3-7 l ,
4
.i i
Table 3-1 CHEMICAL COMPOSITION OF RPV BELTLINE MATERIALS Composition by Veicht Percent , i Identification Heat / Lot No. C Mn P ' S Si Ni Mo _Cg_ Lower Shell Plates: 0.60 0.12 6-146-1 C4689-2 0.21 1.27 0.011 0.016 0.25 10.56 j 6-146-3 C4684-2 0.22 1.41 0.013 0.016 0.24 0.58 0.55 0.13 ' C4627-1 0.22 1.32 0.015 0.016 0.25 0.57 0.56 0.12 6-146-7 Lower-Intermediate Shell Plates: 0.48 0.15 l i 6-139-10 C2773-2 0.22 1.30 0.012 0.018 0.24 0.49 ' C2775-1 0.21 1.34 0.010 0.018 0.19 0.46 0.46 0.13 6-139-11 6-139 C3103-1 0.21 1.35 0.011 0.016 0.24 0.60 0.47 0.14 } Intermediate Shell Plates! 0.55 0.12 6-146-5 C4608-1 0.23 1.30 0.011 'O.015 0.24 0.55 l 0.016 0.25 0.56 0.60 0.12 y 6-146-4 C4689-1 0.21 1.27 0.011 l m 6-146-2 C4654-1 0.23 1.38 0.010 0.015 0.24 0.55 0.56 0.11 Surveillance Plate: C3103-1 see above , 4 I i Longitudinal Welds: . 0.21 D1,D2,D3, 37C065 0.17 1.41 0.015 0.013 0.09 0.21 0.53 j l E1.E2,E3,F1 F2,F3 Lower to Lower-Intermediate Girth Veld: DE 3P4000, Linde 124 0.066 1.37 0.015 0.014 0.36 .0.96 0.46 0.02 , l L Flux Lot 3932 - Intermediate to Lower-Intermediate G!rth Weld: 0.96 0.47 0.051 1.37 0.017 0.018 0.45 0.11 EF : IP4217, Linde 124 Flux Lot 3929 Surveillance Veld:
- 1.56 0.009 - 0.19 0.41 0.51 0.11 i
i 4 I l
4 l Table 3 2 MECHANICAL PROPERTIES OF BELTLINE AND OTHER SELECTED RPV MATERIALS Test Charpy -60 T RT Ident. Heat Temp. Energy NDT 50T NDT Location Number. Number 1*Il (ft-1b) 1*El (*F) (*F) Beltline: , , Lower Shell Plates 6-146-1 C4689-2 10 101,78,69 -10 20 -10 6 146-3 C4684 2 10 63,60,60 -20 -20 -20 6-146-7 C4627-1 -10 65,82,90 20 -20 20 Lower Intermediate 6-139-10 C2773 2 10 44,56,46 10 -8 10 Shell Plates 6-139 11 C2775-1 10 58,66,68 10 -20 10. 6-139 12 C3103-1 10 57,66,76 10 -20 10 Intermediate 6-146-5 C4608-1 10 82,108,90 10 -20 10 Shell Plates 6 146-4 C4689-1 10 60,96,105 10 -20 10 6-146 2 C4654-1 10 55,77,78 10 -20 10 Lon5i tudinal Velds D1,D2,D3 37C065 see Reference 6 (ay-16.44*F) -45 E1.E2,E3 F1,F2,F3 Lower to Lower-Int. Cirth Veld DE 3P4000 10 97,96,90 N/A 50 Lower-Int, to Int. Cirth Veld EF IP4217 10 56,71,60 N/A -50 -50 Non-Beltline: Upper Shell Plate 15-146-2 C4598-1 10 42,80,80 10 -4 10 Vessel Flange 48-146-1 ACN97 10 96,109,127 10 -20 10 Head Flange 209-139-2 AAL95 10 81,87,94 10 -- -20 10 Top Head Torus 202-139-11 C1982-1 10 45,60,90 10 -10 10 Bottom Head Torus A2-146-4 B7255-2 40 81,28,b4 40 54 54 Recire Inlet N2 A7-139-15 EV9934 40 30,40,31 10 50 . . . 50 NOTE: N/A - not available 3-9 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -. _ ... u
6 Table 3-3 CHEMICAL COMPOSITION OF IRRADIATED SURVEILIMICE SPECIMENS
' Composition by Weicht Percent S $1 . Ni Mo Cu Identification C tfg P Plate:
- 0.007 0.24 0.62 0.50 0.13 A41
- 1.45 -
0.007 0.22 0.64 0.52 0.13 A46
- 1.48 -
Veld: 0.009 / 0.19 0.40 0.50 0.11 1.55 y A4P 5 0.009 0.19 0.41 0.51 0.11 AS4-
- 1.57 -
9 F a e
'a A - mark denotes an element'that was not evaluated.
9 Table 3-4 IDENTIFICATION OF UNIT 3 SURVEILIANCE SPECIMENS Charov specimens a Irradiated Unirradiated Irradiated Unirradiated Irradiated EB.E2 1912 Etld E11.4 E 7P1 7PC 7TJ- 7TE A27 7LU 7LY 7TL 7U1 A1C 7T6 7MT -7UA 7U6 A1E 7TB 7MY 7TK 7U2 All 7MU 7P4 7UU 7U3 AIK 7MS 7M3 7TU 7TP A3A 7PP 7M2 7UP 7TM A22 7M7 7P7 7TD 7U4 A2E 7P2 7MD All 7TT A23 7MJ 7PB A15 7US A1D 4 7MB 7MM 7YD 7TY A1J 7PY 7PD A14 7U7 A2T Tensile Specimens A Irradiated- Irr=41ated Irradiated 191t EtM E
-A41 A4P A62 A4A AS4 A5M-A46- ASE a All specimen identifications include two dots.
. 3-11
..__.____j
BASKET CODE REACTOR CODE eeoes 1 +2+8+ 16 = 27 1 e000 e ese e se 1248 124816
\, ,/
OOOO #e000 e se o ese l u b l l i . J ~ _ 1 O Dimple e Hole i ! Figure 3-1. Schematic of Identification on Basket l Recovered from Unit 3 l
Top llead Enclosure i i
. Top llead Flange . .
Vessel Flange 9 O j Upper Shell Transition Shell C g '@ @g > s Plate lleats C4689-1 C4608-1 Uppe'r Intermediate Shell C4654-1
/
EF
- FI, F2, F3's I Plate Heats C2775-1 eltline Lower Intermediate Sheb hhhhh[h "
Region : l Et,E2,E3M [DE U - L wer Shell Plate Heats C4684-2 C @ i @ g @@ ) Eni?:
- \ D1,D2,D3 e i Bottom llead Enclosure . , ' ,
Figure 3-2. Schematic of the RPV Showing Identification of Vessel Beltline Plates and Welds 3-13
e (Cv0H p.0LL
\
\\
\\
= = g\\ HEAT C3103-1 g\ -
\\\ \
\\\\
\\\\ ,
\ \\\ 1/4T
\g\\
\
\
sks\ s\ /T '
\\* L
\\~-k O
F 90 deg e O deg 10 min 1.02 . MARKED END ( -
\ p UNMARKED END
= 2.165 d o.015 :
211/2 deg 0.010 t 0.001 R M .394
' I 3
iW -90 deg 2 0 deg 10 min 0.315 0.394 ALL CORNERS Figure 3-3. Fabrication Method f or Base Metal Charpy Specimens 3-14
DfRE ON- ELECTROSLAG WELD PER WR 12 4, PO-1351 xx r xx
\\ \\ HEAT C3103 1
\N \\
, \\ \
HEAT C3103-1
'N N ./ 5 \
\S
\N N\
\\ \\
\s NN xx gg NN NA
\ \ l \
% l '
\ \ ' \ \ l
\ i N i i s g I
\ \ l % /
\
SCRAP \ k I K\ ' A
\ \ f
'l MeV fast flux reaction cross sections for the iron and copper wires were estimated to be 0.212 barn and 0.00374 barn, respectively. These values wero"ob'tained from measured cross section functions determined at CE's Vallecitos Nuclear Center from more than 65 spectral determinations for BWRs and for the General Electric Test Reactor using activation monitors and spectral unfolding techniques. These data functions are applied to BWR pressure vessel locations 4-1
._m_ _ _ _ _ _ _
based on water gsp (fuel to vessel wall) distances. The cross sections for
>0.1 MeV flux were determined from the measured 1-to-0.1 MeV cross section ratio of 1.6.
4,1.2 Results
~
The measured activity, reaction rate and determined full-power flux results for the surveillance capsule are S iven in Table 4 2. The >l MeV flux value of 6.8x108 n/cm 2-s from the flux wires was calculated by dividing the reaction rate measurement data by the appropriate cross sections. The corresponding fluence results, 1.6x1017 n/cm2 (>1 MeV) wat obtained by multiplying the full-power flux density valut.s by the product of the total time of irradiation and the full power. fraction, shown in table 4-1. Generally, for long term irradiations, dosimetry results from copper flux wires are considered the most accurate because of Co-60's long half-life (5.27 years). The iron flux monitor reaction yielding the shorter half-life (312.1 day) Mn 54 gave results_ about 20% lower than the copper reaction. Because of the long decay period of this surveillance capsule, the nickel wire Co-58 activity (70.8 day half-life) was too low to be utilized. The major difference between the copper and iron wire results is due to differences between the local core power distribution near the surveillance capsule and. the average core power, especially during the-last two cycles. Tnerefore, the flux and fluence above are based only on the copper flux wire results. The accuracies of the values in Tables 4-2 for a 2a deviation are estimated to be: 5% for dps/g (disintegrations per second per grar) i 12% for dps/ nucleus (saturated) i 30% for flux and fluence >l MeV .. *. 40% for flux and fluence >0.1 MeV 4-2
Flux vires from Unit 3 were evaluated by General Electric in 1978. The
>1 HeV flux value was 9,8x108 n/cm 2 -s. A re-evaluation in February 1980, using data from fu:1 neutron spectral determinations at the Browns Ferry 3 vessel, resulted in a revised >l MeV flux of 7.1x108 n/cm 2 -s. Evaluation of the Unit 3 flux vire results from 1978 using current cross section information results in a flux of 8.2x108 n/cm 2 -s. The results from this study are lower, indicating that higher flux was incident on the vessel wall early in plant operation. Due to the accumulation of several cycles of typical operation and core power shape, the results from the copper flux vires analyzed in this report provide a reasonable, but conservative prediction of 32 E?PY fluence.
4.2 DETERMINATION OF I.EAD FACTORS The flux vires detect flux at a sit.gle location. The wires will therefore reflect the power fluctuations associated with the operation of the plant. However, the flux wires are not necessarily at the location of peak vessel flux. Lead factors are required to relate the flux at the wires' location to the peak flux. These lead factors are a function of the core and vessel geometry and of the distribution of btndles in the core. Lead factors were generated for the Unit 2 geometry in 1988, using a typical fuel cycle to determine power shape. Based on a review of core management and the similarity of the Unit 2 and 3 flux wire results (within 10%), it is appropriate to use the lead factors calculated in Reference 6 to compute the peak location flux for Unit 3. The methods used to calculate the lead factors in Reference 6 are discussed below. 4.2.1 crocedure Determination of the lead factor for the RPV inside wall was made using a combination of one-dimensional and two 'imensional finite difference couputer analysis. The two-dimensional analysis established the"rlative e fluw ae in the azimuthal direction at the vessel surface. A series of one-dimensional analyses were done to determine the core height of the axial flux peak. and its relationship to the surveillance capsule height. The combination of azimuthal and axial distribution results provides the ratio of 4-3 l
flux, or the lead factor, between the surveillance capsule location and the peak flux location. The two d!aensional DOT IV computer program was used to solve the Boltzman transpor, equation using the discrete ordinate method on an (R,#) geometry, assuming a fixed source. Eighth core symuetry was used with reflective bouhdary conditions at 0* and 45*. Neutron cross' sections were determined for 26 energy groups, with angular scattering approximated by a l third o. der Legendre expansion. A schematic of the two-dimensional vessel j model is shown in Figure 4-1. A total of 113 radial intervals and 45 i azimuthal intervals was used. The model consists of an inner and outer core , region, the shroud, water regions inside and outsics the shroud, the vessel vall and the drywell. Flux as a function of azimuth was calculated, 3 establishing the azimuth of the peak flux and its magnitude relative to the } flux at the surveillance capsule azimuth of 30*. J The one-dimensional t ".puter code (SN1D) was used to calculate radial flux distribution at several core elevations at the azimuth angle of 25.5*, l where the azimuthal peak was determined to exist. The elevation of the peak flux was determined, as well as its magnitude relative to the flux at the j surveillance capsule elevation. l 4.2.2 Results 4
- The two-dimensional calculation in Reference 6 indicated the flux to be I a maximum 25.5* past the RPV quadrant references (0*, 90*, etc.). The peak l closest to the 30* location of the surveillance capsules removed is at 25.5*.
The distribution calculations establish-the lead factor between the surveillance capsule location and the peak location at the inner vessel wall. This lead factor is 0.95. The fracture toughness analysis is based on a 1/4 T depth flaw in the beltline region, so the attenuation of the flux, to, that j depth is considered. This attenuation is calculated according to the requirements in Regulatory Guide 1.99, Revision 2 (Reference 5), as shown in the next subsection. ,t
. 4-4
4.3 ESTIMATE OF 32 EFPY FUUENCE The inside surface fluence (f surf) at 32 EFPY is determined from the best estimate of the measured flux from Table 4 2, using the lead factor in Section 4.2. The time period 32 EFPY, typically assumed for 40-year operation (80% capacity factor) is 1.01x109 secords. The resulting 32 EFPY fluence
* ~
value at the ve'ssel inside surface is: f surf - (6.8x10 8 n/cm2 .3)(1,oixio9 s)/0.95 f surf - 7.2x10 17 n/cm2 , The 1/4 T fluence (f) is calculated according to the following equations from Reference 5: f - f urf(*-0.24x) s (4,1)
- where x - distance, in inches, from the inside surface to the l 1/4 T depth.
The vessel beltline consists of the lower-intermediate and lower shell, both with a thickness of 6.125 inches. For the thickness of 6,125 inches , the l corresponding depths x is 1.53 inches. Equation 4-1 evaluated for Unit 3 is: f - 0.6925 fs urf l The 1/4 T value of 32 EFPY fluence is as follows: l f - 5.0x1017 n/cm2
.h O'e i
k 4-5
Table 4 1
SUMMARY
OF DAILY POWER HISTORY Operating Fraction of Days Between Cycle Cvele Dates Days . Full Peggr Cveles } - . . 846 0,603 k 1 9/1/74 - 12/24/76 ! 108 l: i 2 4/12/77 - 4/1/78 355 0.779 i- 48 3 5/20/78 - 9/15/79 484 0.844 51 ,f 4 11/6/79 - 3/7/81. 488 0.793 j 229 5 10/23/81 - 2/13/83 479 0.903 243 l 6 10/15/83 - 7/15/85 640 0.772 t
- . 228 l 7 3/1/86 3/31/87 JH MM 1
I j 3688 0.750 (average) f i i i 4-6
Table 4-2 SURVEILIR1CE CAPSULE FLUX AND FLUENCE FOR IRRADIATION FROM 9/1/74 TO 3/31/87 Wire dps/g Element Reaction Rate Full Powgr Flux
- Fluenge Wire Weight (at end of [dps/ nucleus (n/cm -s) (n/cm )
(saturated)1 >1 MeV >0.1 MeV >1 MeV >0.1 MeV (Element) __f_gl_ Irradiation) l8 Copper 65156 'O.3630 7.63x10 2.46x10[y8 Copper 65157 0.3697 8.15x10 3 2.63x10 -18 - Copper 65158 0.3691 7.86x10 2.53x10 1 1 Average - 2.54x10" 6.6x10 1.1x10' 1.6x10 2.6x10 Iron 65156 0.1286 3.97x10 1.15x10}I 1 Iron 65157 0.1314 4.18x10 4 * *
-16 e- Iron 65158 0.1331 3.93x10 1.14x10 e
-16 l Average - 1.16x10 5.5x10 8.8x10 1.3x10 2.1x10 J
8 9 1 17 ' Values from this report (copper only) - 6.8x10 1.1x10 1.6x10 2.6x10 i i !
- Full power of 3293 MV .
4 e f I
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, CORE tt:TERIOR:
14lNTERVALS 2 5 2 2 1 1 1 1 1 1 2 2 2 2 2 2 2 2 CORE EXTERIOR. 10 INTERVALS { 2 2 2 2 2 2 l2 < i l " WATER REGION. 23 INTERVALS
, CAPSULE }$ OUD: 7 INTERMS WATER REGION:
i " 35 INTERVALS VESSEL WALL 45 INTERVALS 17 INTERVALS IN AZIMUTHAL DIRECTION ORYWELL AIR RECl0N: 7 INTERVALS t
- 1. CORE INTERIOR FUEL 2 = CORE EXTERIOR FUEL j rigure 4-1. Schematic of Model f or Two-Dimensional Flux .
Distribution Analysis 4-8
s
- 5. CHARPY V NOTCH IMPACT TESTING The 36 Charpy specimens recovered from the Unit 3 surveillance capsule were impac t tested at temperatures selected to establish the toughness transition and upper shelf of the irradiated RPV materials. Testing was conducted in accordance with ASTM E23 82 (Reference 9).
5.1 IMPACT TEST PROCEDURE The testing nachine used was a Riehle Model Pl.-2 impact aachine serial number R 89916 The pendulum has a maximum velocity of 15.44 ft/see and a maximum available hammer energy of 240 f t-lb. The test apparatus and operator were qualified using NIST standard reference material specimens. The standards are designed to fail both at 70.5 f t-lb and 11.5 f t lb at a test temperatur*1 of 40'F. According to Reference 9, the test apparatus averaged results must reproduce the NIST standard values within ar. accuracy of 15% or 11.0 f t-lb, whichever is greater. The qualification of the Riehle machine and operator is summarized in Table 5 1. The calibration results for the low energy specimens are 0.4 ft lb higher thm the allowable variation in Reference 9. This is due to the occasional " kickback" of a broken specimen half contacting the hammer, which a only significant to-t'..e Charpy results at very low energies. Since the kickbacks have little effect on the results, especially compared to the typical scatter in the data, no correction to the test results was made. Charpy V-Notch tests were conducted at temperatures between 40'r and 300*F. For tests below 70'F methanol was used as the cooling fluid. Between 70'F and 212'F, water was used as the temperature conditioning fluid. The specimens were heated in oil above 212*F. Cooling of"the conditioning fluids was done with liquid nitrogen, and heating by an-immersion heater. The fluids were mechanically stirred to maintain uniform ! temperatures. The fluid temperature was measured-with a calibrated thermocouple. Once at test temperature, the specimens were manually transferred with centering tongs to the Riehle machine and impacted within 5 seconds. 51
1 For each Charpy V. Notch specimen the test temperature, energy absorbed, lateral expansion, and percent shear were evaluated. In addition, photographs were taken of each fracture surface pair. Lateral ; expansion and percent shear vere measured according to Reference 9 methods. , Percent shear was determined with method one of Subsection 11.2.4.3 of Reference 9, which involves measuring the length and width of the cleavage surface and locating the percent shear value fion Tables' 1 or 2 of . Reference 9. 5.2 IMPACT TEST RESULTS Twelve Charpy V Notch specimens each of base, veld, and HAZ material vere tested at temperatures ( 40*F to 300*T) selected to define the toughness transition and upper shelf portions of the fracture toughness curves. The absorbed energy, lateral expansion, and percent shear data are listed for each material in Table 5 2. Plots of absorbed energy data for base, veld and MAZ materials are presented in Figures 51 through 5 3, respectively. Lattral expansion plots for base, veld and HAZ materials are presented in Figures 5 4 through 5 6, respectively. The fracture surface photographs and a summary of th test results for each specimen are contained in Appendix A. The plate and weld data sets are fit with the hyperbolic tangent function developed by Oldfield for the EPRI Irradiated Steel Handbook (Reference 10): i I Y = A + B
- TANH [( T - To )/C),
where Y - impact energy or lateral expansion l T - test temperature, and ! A, B, To ard C are determined by non-linear ,,.., regression. The TANH function is one of the few continuous functions with a shape characteristic of lov alloy steel fracture toughness transition curves. 52 l
4 5.3 1RRADIATED VERSUS UNIRRADIATED CllARI"I V NOTCil PROPERTIES As a part of the RPV surveillance test program, some Charpy V.Hotch testing was performed. Data for the plate material specimens (Heat C3103 1) were recovered from QA records, but consisted only of three specimen tests at 10'F. For the surveillance veld, positive identitication of the veld wire heat was not obtained from B&W, Therefore, unirradiated archive surveillance specimens were obtained from PECO and tested to establish the baseline needed to determine the irradiation shift. To minina ce sonnel exposure, the unirradiated specimens were tested on the Tinius Olsen Charpy test apparatus in San Jose (serial number 119037), with available hammer energy of 264 f t lb and impact velocity of 16.8 ft/sec. The machine was cal'ibrated with HIST specimens in May 1990. The calibration results, shown in Table 5 3, were within specifications. Twelve each of plate and veld material specimens were tested at , temperatures between .40*F and 300'F. Impact energy, lateral expats;on and percent shear were determined vith the same methods used for the irradiated specimens. The results ase presented in Table 54. Photographs of specimen fracture surfaces are in Appendix B, t l The impact energy and MLE data for the unirradiated materials were fit with the TANH function in the same manner as the irradiated data. The-l results for plate and veld impact energy are plotted in Figures 5-7 and j 5-8, respectively, along with the corresponding irradiated data described in Section 5.2. The results for lateral expansion are plotted in Fi 5ures l 5 9 and 5 10 for the plate and veld material, respective 13 The irradiated and unirradiated Charpy V Notch data were used to estimate the values given in Table 5-5: 30 f t lb , 50 ft-lb and 35 M.LE . index temperatures, and the USE for both irradiated materials and unirradiated materials, where available. Transition temperature shift values are determined as the change in the temperature at which 30 ft-lb impact energy is achieved, as required in Reference 4. Values were not determined for HAZ material because of the amount of scatter in the Charpy data, which is typical for that material.
, 5-3
5.4 COMPARISON TO FREDICTED SHIFT The measured transition temperature shifts for the plate and weld materials were compared to the predictions calculated according to Regulatory Guide 1.99, Revision 2 (1.99) . The 1.99 nethods used to calculate shift are described in Section 7. 6. The inputs for the surveillance p1' ate and weld materials are as follows:- Plate: (based on Table 3 3 values) Copper: 0.13% Nickel: 0.63% fluence: 1.6 x 1017 n/cm2 4 Ecl.d: (based on Table 3.'3 valuns) Copper: 0.11%- Nickel: 0.41% fluence: 1.6 x 1017 n/cm2 The chemistry factors (CF) are 91.8 for the plate and 102.5 for the veld. The fluence factor is 0.1495. The predicted shifts are, therefore, as follows: ARTNDT ARTNDT + Hargin Actual Test 1 Plate 13.7'T 27.4*F 16*F ! Weld 15.3*F 30.6*F 16*F
- The measured shifts of 16*F for both the plate and veld materials are slightly more than the mean prediction, and considerably less than the
! upper bound predictions of 1,99.
,, n.
f
- 5-4 1
t Table 5 1 VALLECITOS RIDILE CHARPY MACHINE QUALIFICATION TEST RESULTS USINC , NIST STANDARD REFERENCE SPECIMENS (TESTED 3/5/90) Test Energy Acceptable [ Specimen Bath Temperature Absorbed Range Identification Medium _ ('F) (ft lb) (ft-lb) HM 15 086 Hethanol 40 70.0 MM 15,503
- 40 69.0 MM 15 364 a 40 71.0 MM 15 244
- 40 67.5 MM 15 326
- 41 66.0 Average 68.7 70.5 1 3.5 LL+18 105 Methanol 40 12.5 LL 18 246
- 40 13.5 LL 18 496 *
-41 13.5 ,
LL-18 1012 ," 40 12.5 LL 18 040 " 40 12.5 Average 12.9 11.5 1 1.0 l l
..i.
5 I
d I, Table 5 2 l . CHARPY V. NOTCH IMPACT TEST RESULTS .i POR IRRADIATED RPV KATERIALS IN UNIT 3 ; i j Test Practure Lateral Percent Shear j Specimen . Temperature Energy - Expansion (Method 1) t identification (*F) (ft-lb) (mils) (%) } J Base: 7P1 40 8.0 6 3 7 00 30 8.0 11 7 20 13.5 13 9 4 7T6 10 15.5 14 8 7TB . 7MU 0 43.5 34 16 7MS 20 46.0 42 22 ' 7PP 40 65.0 50 23 7M7 80 '112.0 BB 73 7P2 120 102.5 70 67 l 7HJ 160 129.0 86 100 7MB 200 121.0 89 100 2 7PY 300 128.0 88 100 Weld: 7TJ 40 11.0 9 4 . 7TL 20 18.0 16 3 i 70A 10 21.5 20 6 - l 77K 0 12.0 14 3 j 7UU 10 20.0 20 15
- 7TU 20 38.5 35 11 7UP 40 50.0 42 23' 7TD 80 61.5 56 32 All 120 80.5 73 75 l 84 i A15 160 96.5 74 7YD 200 100.5 85 100
+ A14 300 98.0 81 100 HAZ: A27 40 16.5 16 2 i Alc 30 23.5 20 12 . A1E 20 37.5 34 8 ) Alt 10 41.5 31 16 2 AIK 0 21.5 21 19 . A3A 20 83.0 60 40 ..". A22 40 111.5 83 47 A2E 80 125.0 91 82 A23 120 99.0 70 81 A1D 160 103.0 87 100 A1J 200 107.0 88 100 A2T 300 162.0 86 100
. 56
l l Table 5 3 SAN JOSE TINIUS OLSON CilARPY MACilINE QUALIFICATION TEST RESULTS USING 4 NIST STANDARD REFERENCE SPECIMENS (TESTED 5/14/90) Test Energy Acceptable i Specimen Bath Temperature Absorbed Range Identification Medium- ('F) (ft 1b) (ft lbi i MM.16 307 Methanol 40 71.0 i MM 16,465 40 72.0 HM 16 $09
- 40 75.0 MM 16 552
- 40 71.5 :
MM 16 915
- 40 73.0 Average 72.5 69.4 1 3.5 LL 19 124 Methanol 40 11.5-LL 19 354
- 40 l't s LL 19 411
- 40 12.0 LL 19 702
- 40 10.5 i
LL.19 925
- 40- 11.5 i
Average- 11.4 11.9 i 1.0 1 1 . l l l .s*s. ' l
- l. >
- 57 . . - . . - _ - _ _ _.-.-._-._..-.....___...--_-_----...--aa_.
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~! Table 5 4 $ CHARPY V+ NOTCH IMPACT TEST RESULTS 1
j TOR UNIRRADIATED RPV MATERIALS IN UNIT 3 1 l li Test Tracture Lateral Percent Shear Specimen Temperature Energy Expansion (Hethod 1)
- (ttils ) * (1)
Identification ('ri (ft lb) Base: l 7PC -40 10.5 8.5 0 1 l 7LY 40 24.5 16.5 0 4 7MT +20 18.5 16.0 16 i 7MY 0 34.0 26.0 15 i 7P4 20 60.5 43.5 27 7M3 40 85.0 62.0 39 l 7M2 60 72.0 51.5 38 i 7P7 80
- 98.5 70,0 67
} 7MD 120 117.5 86.5 83 i 7PB 160 130.0 81.0 100
! 7MM 200 130.0 80.0 100 j 7PD 300 143.5 83.0 100 i
! Weld: i 7TE 40 6.5 3.0 0 i 7U1 20 17.5 15.5 6 I 7U6 0 29.0 20.5 11 i 7U2 0 48.0 41.5 15 4 7U3 20 39.5 30.5 14 7TP 40 56.5 46,0 22 7TM 60 44.0 39.0 39 l 704 80 54.0 40.5 41 77T 120- 92.0 70.0 77
. 705 160 101.0 77.0 100 l- 7TY 200 106.5 80.0 100 7U7 300 104.0 79.0 100 i
. a**.
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I . , I Table 5 5 SIGNIFICANT RESULTS OF IRRADIATED AND UNIRRADIATED CllARPY V NOTCll DATA FOR UNIT 3 l Upper Shelia
' ~
Index Temocrature ('F) Energy (ft lb)
- ,,,_lialtr i al F-30 ft 1b F-$0 f t 112 MLE-35 mil L/T i
1 ! Unirradiated Plate -14 17 12 137/ 89 Irradiated Plate __l 21 11 123/ 80 a Difference 16 6 1 14/ 9 (10%) i i . Unirradiated Weld 2 47 34 110
- Irradiated Weld 11 $4 ,,31 Jpg j Difference 16 7 1 10 (9%)
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Figure 5-8. Comparison of Unirradiated and irradiated Weld Metal impact Energy
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- 6. TENSILE TESTING Eight round bar tensile specimens were recovered from the Peach B:ttom Unit 3 surveillance capsule and tested. Uniaxial tensile tests vere ,
j conducted in air at room temperature (70'F), RFV operating temperature (550*F), and onset of upper shelf temperature (16,0 ' F) . The tests were 3 conducted in accordance with ASTM EB-81 (Reference 12). 6.1 FROCEDURE All tests were conducted using a screw. driven Instron test frame
- equipped with a 20 kip load cell and special pull bars and grips. Heating was done with a Satec resistance clamshell furnace centered around the specimen load train. The test temperature was monitored and controlled by a chrome 1.alumel thermocouple spot welded to an inconel clip that was Before the friction-clipped to the surface of the specimen at its midline.
elevated temperature tests, a profile of the furnace. was conducted at the test temperature of interest using an unirradiated steel specimen of the same geome try . Thermocouples were spot welded to the top, middle, and bottom of a central 1 inch gage of this specimen. In addition, the clip on thermocouple was attached to the midline of the specimen. When the target temperatures of the three thermocouples were within 15'F of each other, the temperature of the clip on thermocouple was noted .and subsequently used as i the target temperature for the irradiated specimens. All tests were conducted at a calibrated crosshead speed of 0.005 inch / min until vell past yield, at which time the speed was increased to 0.05 inch / min until fracture. l
.s .
I ! 61
4 The test specimens were machined with a minimum diameter of 0.250 ine J l I at the center of the gage length. The yield strength (YS) and ul' J j tensile strength (UTS) were calculated by dividing the nomin . o l (0.0491 in2 ) into the 0.2% offset load and into the maximum t< , respectively. The values listed for the uniform and total elongat 5;; 1 ! obtained from plots that recorded load versus specimen extension ;e l based on a 1.0 inch gage length. Reduction of trea (RA)' values were
- determined from post test measurements of the necked specimen diameters using a calibrated blade micrometer and employing the following formula:
i
- RA - 1001 * (Ao - Ar)/Ao i
j After testing, each broken specimen was photographed end on, showing the fracture sw M, and lengthwise,' showing the fracture location and necking. I ! 6.2 RESULTS l Irradiated tensile test properties of Yield Strength (YS), Ultimate 3 Tensile Strength (UTS), Reduction of Area (RA), Uniform Elongation (UE), and Total Elongation (TE) are presented in Table 6-1. A stress strain curve ! for a 550*F base metal irradiated specimen is shown in Figure 6-1. This curve is typical of the stress strain characteristics of all the tested specimens. The data in Table 6-1 are shown graphically in Figures 6 2 and 6-3. As can be seen from Figures 6-2 and 6 3, the base, weld and HAZ j materials generally follow the trend of decreasing-properties with increasing temperature. Photographs of the fracture surfaces 'and necking behavior are given in Figures 6 4 through 6 6. 6.3 IRRAD'IATED VERSUS UNIRRADIATED TENS 11.E PROPERTIES Unirradiated tensile test data, shown in Table 6-2, were recovered from
..o.
QA records for surveillance plate Heat C3103 1. The uniciadiated data provide average values of YS, UTS , RA, ar.d TE at room temperature. -These were compared to the irradiated plate specimen RT data to determine the ' irradiation effect. The trends of increasing YS and UTS and of decreasing RA, characteristic of irradiation embrittlement, are seen. The increasing TE seen may be due to different reference gage lengths for the tests.
, 6-2
i Table 6 1 1 1 1 I TENSILE TEST RESULTS FOR IRRADIATED RPV MATERIALS FOR UNIT 3 l Test Yield" Ultimate Uniform Total Reduction ,' Specimen Temp Strength Strength Elongation Elongation of Area
- Number 1111 (ksi) (ksii (%) (%) "(%)
Base: A41 68 74.0 96.0 14.5 29.5 66.6 A4A 160 68.9 90.4 13.0 27.3 69.9 A46 550 67.3 95.6 14.0 26.9 59.6 i l Weld:
- A4P 66 64.2 86.9 15.0 29.7 68.6 i AS4 550 60 0 83.4 12.5 24.5 60.7 4
i HAZ 4 A62 68 67.0 88.5 13.8 27.4 67.6 , ASM 160 63.6 84.2 13.5 28.8 68.3 j ASE 550 61.5 83.8 10.6 23.0 64.2 1
" Yield Strength is determined by 0.2% offset.
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Table 6 2 COMPARISON OF UNIRRADIATED AND 1RRADIATED TENSILE PROPERTIES AT ROOM TEMPERATURE FOR UNIT 3 Yield Ultimate Total Reduction Strength Strength Elongation , of Area (ksi) (ksi) (%) f%) Plate: l Unitradiated 71.3 91.5 28.1 71.6 Irradiated 74.0 . 96.0 29.5 66.6 Difference a 3.8% 4.9% 5.0% -7.0% a Difference = [(Irradiated Unitradiated)/Unirradiated)
- 100%
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ENGINEERING STRAIN, % Figure 6-1. Typical Engineering Stress-Strain for Irradiated RPV Materials i I
" BASE UTS o HAZ
-- YS + WELD 100 .
90 - N N - T 80 - 5 E
. o w a 5 's s 70 -
A s
's-------_____________ l
' l
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=--________._
i 60 - i ' ' 50 i i 200 400 600 O L TEST TEMPERATURE ( F) Figure 6-2. Strength vs. Temperature for Irradiated Materials
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- TEST TEMPERATURE ( F)
Figure 6-3. Ductility vs. Temperature for irradiated Materials
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k 1 (_ . a, .. . -__ 3 - s ASE 550*F Figure '-6. Fracture Location, Necking Behavior, and Fracture Appearance for Irradiated HAZ Tensile Specimens 6-10
- 7. DEVElDPMENT OF OPERATING LIMITS CURVES 7.1 BACKCROUND 2
Operating limits for pressure and temperature are required for th ee categories of operation: (a) hydrostatic pressure tests and leak tests, referred to a s' Curve A; (b) non-nuclear heatup/looldown a'nd low level physics tests, referred to as Curve b; and (c) core critical operation, referred to as Curve C. There are three vessel regions that affect the operating limits: the closure flange region, the core beltline region, and the temainder of the vessel, or non-beltline regions. The closure' flange region limits are controlling at lower pressures primarily because of Referenc.e 1 requirennts . The non beltline and beltline region operating
~
limits are evaluated according to procedures in References 1 and 2, with the beltline region minimum temperature limits increas'ng as tne vessel ic
.sted.
BELTLINE REGIONS
.t- b el tline regions are those locations that receive too little
.c e to cause any RTNDT increase. Non beltline components include the nozzles, the closure flanges, some' shell plates, top and bottom head plates and the control rod drive (CRD) penetrations. Detailed stress analyses of the non beltline components, considering oporating transients with relatively high pressures and low temperatures, were performed for the BWR/6, specifically for use in developing pressuro-temperature (P-T) limits.
The analyses bounded all mechanical loadings and thermal transients anticipated. Detailed stresses were used according to Reference 2 to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). These results are applicable to the Unit 3 vessel components, since the non-beltline geometries gre.not significantly different from BWR/6 configurations and the mechanical and thermal loadings are comparable. 7-1
4 ? The non beltline region results were established by adding the highest RTNDT for the non beltline discontinuities to the P versus (T RTNDT) curves for the most limiting BWR/6 components, which are the CRD penetration !! and feedwater nozzle. Table 3 2 has the limiting RTNDT values applicable to , the feedwater nozzle limits and to the CRD penetration limits. They are 50*F for the feedwater nozzle limits, based on the RTHDT of the recirculation inlet nozzles, and 54'T for the CRD pe'netration ilmits,. based on the RTNDT of the bottom head torus plates. i. 7.3 CORE BELTLINE REGION i i As the beltline fluence increases during operation, the beltline P-T }
- limit curves shift to the right by an amount discussed in Subsection 7.6.
I
, Depending on the amount of shift, the beltline curves may or-may not become more limiting than the non beltline curves at some time during plant life.
The stress intensity factors calculated for the beltline region according to i Reference 2 procedures are based on a combination of pressure and thermal stresses. The pressure stresses were calculated using thin walled - cylinder equations. Thermal stresses were calculated assuming the through-wall temperature distribution of a flat plate subjected to a 100*F/hr thermal gradient. The adjusted RTNDT (ART) values calculated in Subsection 7.6 for ! the limiting beltline materials were used to adjust the P versua (T - RTNDT)
- values from Figure G 2210-1 of Reference 2. ART is the initial RTNDT plus the shift in RTNDT due to irradiation.
7.4 CIDSURE FLANGE REGION References 1 and 2 have several requirements that affect the P-T ! curves, based on the RTNDT values in the closure flange region. As stated in Paragraph G-2222(c) of Aeference 2 for application of full bolt preload ' . and reactor pressure up to 20% of preservice hydrostatic test pressure -(312 psig), the closure flange region metal temperature must-be at RTNDT or greater. The GE practice, however, is to recommend (RTNDT + 60*F) ir bolt preload, fer two reasons: 7-2 i 6
, , . - - , -- , e ~ -- -n., - + - , ~ -.
- . - , -n- . , , - ,,, , , - , , . , , ,
4 4
- a. The original ASME Code of construction required (RTNDT + 60'F);
and
, b. The highest stressed region during boltup is the closure flange ...
region, and the flaw size asstuned in that region. (0.24 inches) is 1 less than 1/4 T. This flaw size is detectable using ultrasonic testing (UT) techniques. In fact, References 13 and 14 report that a flaw in the closure flange region of 0.09 inches can be reliably detected using UT. For Unit 3, (RTNDT + 60*F) of the closure region materials is 70*F, because the RTNDT values for the upper shell plates, top head plates and flanges are
- all 10*F,, and the LST of the closure studs is 70'F. Therefore, the bolt j preload temperature used in developing the P.T curves was 70'F.
., Reference 1, Paragraph IV.A.2, sets requirements on minimum temperature
; when pressure is above 312 psig. The requirements are based on the RTNDT of the closure region. Curve A temperature must be no less than (RTNDT + 90*F) and Curve B temperature no less than (RTNDT + 120*F)_ The Curve A requirement causes a 30*F shift at 312 psig on the P-T curves. The Curve B requirement is not seen on Curve B, because the non-beltline part of the curve is more limiting.
7.5 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPEM IZ G
- Curve C, the core critical operation curve, is develeped from the requirements of Reference 1, paragraph IV.A.3. Essentially paragraph IV.A.3 4
requires that Curve C be 40'F above any Curve A or B limits. Curve B is more limiting t'aan Curve A, so Curve C is Curve B plus '40*F. Curve C initiates at zero pressure at (RTNDT + 60'F), based on an exception for BWRs j in Paragraph IV.A.3, allowing critical operation at temperatures bel,ow,the hydrostatic pressure (Curve A at 1100 psig) test temperature. This exception is valid only when water level is within the normal' range and pressure is below 312 psig. 4
- 7-3 m,-. , _ - , ._ , - ~,, . . . 4- , .,.,,..--,-r-- - ~ , -
4 1 7.6 EVALUATION OF RADIATION EFFELIS The impact on adjusted reference temperature (ART) due to irradiation in the beltline materials is determined according to the methods in
- Reference 5, as a function of neutron fluence and the element contents of i
ccpper (Cu) and nickel (Ni). The specific relationship from Reference 5 is: ART - Initial RTNDT + Shift + Margin (7-1) ] where: 3 Shift - [CF)*f(0.28 - 0.10 log f) (7 2) Margin - 2*(a72 + ,32 )1/2 (7 3) CF - chemistry factor from Tables 1 or 2 of Reference 5 f - 1/4 T fluence (n/cm )2 divided by 1019, 81 - standard deviation on initial RTNDT. og - standard deviation on RTNDT shift, is 28'F for velds
- and 17'F for base material, except that og need not exceed 0.50 times the Shif t value.
4 The limiting beltline plate and weld are determined based on the Cu-Ni J content and initial RTNDT of the materials. Calculations to determine 32 EFPY ART values, and chus the limiting beltline materials, are summarized in - Table 7-1. The results show that a lower-intermediate shell plate, Heat C2773-2, is the most limiting beltline material. i "ae input to the Reference 5 calculations not shown in Table 7-1 is a1 As determined in Reference 6, the longitudinal electroslag velds hiave
'I - 16.44*F. For all the other beitline materials, where RTNDT values were l
determined from measured Charpy data, a1 - 0"F. The basis for,us,ing a1 - 0*F is discussed in more detail in Appendix C. i i i 7-4 _ _ . , _ - _ _ _ _ . . _ _ _ - . ~ __ .. - - _ . . - _ . . _ _ _ - . _ _ _ _ _ . . _ _ _ . _ _ - .
7.6.1 Measured Versus Predicted surveillance shift Section 5 of this report conipares the measured shift for the. plate surveillance specimens with predicted shif ts for the surveillance plate material based on the method shown in Equation 7-2. R.ference 5 states'that surveillance data may be used in place of Equationn 7-2 and 7-3 when Iw_n i sets of credibl~e data are available. This is only the first set of data for l Unit 3. Therefore, determinations of ART for the purposes of developing P-T curves are based on Reference 5 predictions. 3 7.6.2 ART Versus EFM i i Equations 7-1 through 7-3 were evaluated for the beltline plate and weld metals. Table 7-1 shows the 32 EFPY fluence at the vessel 1/4 T locations to be 5.0x1017 n/cm2 . Calculations of the plate and weld ART values at 32 EFPY show that Heat C2773-2 is the most limiting material throughout plant life. Figure 7-1 shows the ART for Heat C2773-2 as a function of full power years of operation, through 32 EFPY. 7.6.3 Fracture Touchness Conditions at 32 EFPY Paragraph IV.B of Reference 1 sets limits on the ART and on the upper . shelf energy (USE) of the beltline materials. The ART must be less than ! 200*F, and the USE must be above 50 f t-lb. Based on Table 7-1, the ART values at 32 EFPY of 71*F or less are acceptable. Calculations of 32 EFPY USE, using Reference 5 methods, wcre done for the surveillance plate and the beltlins welds , and are summarized in
~
Tab'l e 7-2. The fabrication data for the other beltline plates'did not include USE data, so general conclusions for those plates are drawn based on the results for the surveillance plate. The surveillance plate data are i longitudinal so the equivalent transverse USE of the plate material is taken as 65% of the longitudinal USE, according to Reference 11. The weld metal USE has no transverse / longitudinal correction because weld metal has no orientation effect. The longitudinal veld initial USE was computed in Reference 6, based on data for nine weld prolongations. The girth weld initial USE is based on Charpy data taken at only 10*F, so the girth weld
, 7-5 4
results are conservative. Extrapolating to the 32 EFPY fluence according to Reference 5, the lowest predicted transverse USE values for the plate and l veld materials are 78 f t-lb and 82 f t-lb,- respectively. While there may be some variation in the initial USE values of the beltline plates, the T moderate copper levels and the relatively low fluence at 32 EFPY, considered in conjunction with the results for Heat C3103-1, indicate that the 32 EFPY j USE will be above 50 ft-lb for all beltline plates. - . Based on the above results, it is expected that the beltline materials , will have 32 EFPY USE above 50 ft-lb, as required in Reference 1. Since USE i and ART requirements are met, irradiation effects . are not severe enough to necessitate additional analyses or preparations for RPV annealing. i 7.7 OPERATING LIMITS CURVES VALID ~TO 32 EFPY Figure 7-2 shows - P-T curves for Unit 3, valid to 32 EPPY. The P-T curves are developed by considering the requirements applicable to the non-beltline, beltline and closure flange regions. In reviewing the shifted beltline curves, it was dereunined that the non beltline curves are still limiting at 32 EFPY. Therefore, barring any changes due to future L surveillance data or revisions to regulations, the P-T curves shown in Figure 7-2 will apply for operation through 32 EFPY, ! 7.8 REACTOR OPERATION VERSUS OPERA"'ING LIMITS For most reactor operatin5 conditions, pressure and temperature are at saturation conditions, whit.h are well into the acceptable operating area (to i the right of the P-T curves). The most severe unplanned transient relative to the P-T curves is an upset condition consisting of several transients which result in a SCRAM. -lhe worst combination of pressure and temperature
- during this postulated event is 1180 psig with temperatures in the lower head of 250*F. In this case, the core is not critical, so the non-nuclear heatup/cooldown curve applies (Curve B). As seen for Curve B in Figure 7 2, at'1180 psi the minimum transient temperature of 250*F lies in the 2
acceptable operating area. Therefore, violation of the P-T curves is only a concern in cases where operator interaction occurs, such as durf ng pressure testing and initiation of criticality. 7-6 i
_ . _ . _ . . _ _ _ _ . __ __ . _ _ . _ _ . . - _ _ _ . _ . . . -_ _ - _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . . _ . _ .. __.m_.__ .. .. _ m . Table 7-1 BELTLINE EVALUATION FOR PEACH BOTTOM 3 h Shell Fluence: Thickness - 6.125 inches Peak I.D. fluence - 7.2E+17 - Peak 1/4 T fluence.- 5.0E+17
- Initial 32 EFFY 32 EFPY 32 EFPY.
%Cu INi CF RTndt Delta RTndt Margin Shift ART COMPONENT 1.D. HEAT OR HEAT / LOT PLATES:
Lower Shell 6-146 1 C4689-2 0.12 0.56 82.2 -10 24.0 24.0 48.0 38.0 6-146-3 C4684-2 0.13 0.58 90.4 '- 20 26.4 26.4 52.8 32.8 Lower Shell 48.2 28.2 Lower Shell 6 146-7 C4627 1 0.12 0.57 82.4 20 24.1 24.1 Y w C2773-2 0.15 0.49 104 10 30.4 30.4 60.8 70.8 Low-Int Shell 6-139-10 25.4 25.4 50.7 60.7
- Low-Int Shell 6-139-11 C2775-1 0.13 0.46 86.8 10 j Low-Int Shell 6 139-12 C3103-1 0.14 0.6 100 10 29.2 29.2 58.5 68.5 i
6-146-5 C4608 1 0.12 0.55 82 10 24.0 24.0 47.9 57.9 Intermediate 48.0 58.0 Intermediate 6-146-4 C4689-1 0.12 0.56 82.2 10 24.0 24.0 l 21.5' 53.0 Intermediate 6-146-2 C4654-1 0.11 0.55 73.5 10 21.5 43.0 l WELDS: Lower Long. D1,D2,D3 37C065 0.21 0.21 109.'3 -45 31'.9 45.8. 77.8- 32.8 - Low-Int Long. E1.E2,E3 37C065 'O.21 0.21 109.3 -45 31.9 45.8 77.8 32.3 i ~.
; ' Intermediate 'F1;F2,F3 37C065 0.21 0.21 109.3 -45 31.9 45.8 77.8 32.8 i -50 7.9' 15.8' -34.2 i Lower to DE 3P4000,Linde 124. 0.02 0.96 27- 7.9 ; Low-Int CLrth Flux Lot 3932 Law-Int to EF IP4217,Linde 124 0.11 0.96 147.2 50- 43.0 43.0 86.0- 36.0 i Intermed. Cirth Flux Lot 3929 .
I
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4 l Table ?
- ESTIMAT0 OF UPPER SHELF ENERbf FOR BELTLINE MATERIALS 4
j 32 EFPY a Upper Shelf (ft-lb) Decrease Loncitudinal/ Transverse
- Identification ~ % Cu in USE Unirradiated ._32. EEtL i
Surveillance Plate: C3103 1 0.14- 11% 137/89 122/79 I Beltline Plate (highest copper): 0.15 12% 137/89 120/78 Longitudinal Welds: l 37C065 0.21 17.5% 99 82 .. Lower to Lower-Intermediate Cirth Weld: 3P4000 0.02 8% 97 b 89 Lower Intermediate to Intermediate Cirth Weld: IP4217 0.11 12.5% 71 b 62 a USE decrease percentages based on 32 EFPY fluence of 5.0x1017 n/cm 2 b Values are highest Charpy energy from tests at 10*F. 7-8
'I
- J .
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a. 4 t ' 20 l . o ! 19-i . 0 .. 0 4 8 12 16 .20 24- 28 32 EFFECTIVE FULL POWER YEARS i [ F'igure 7-1. ! ART of Limiting. Beltline. Material j
- i
! .[
.j i
j' i- .
_ _ _ _ . . . - _ _ . _ . . . . . _ . . _ . . . - . . . _ _ . . _ . ~ . . . _ _ _ . . . . . . - 1 4 g s-1600 i
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~~ i f B,B' - NON-NUCLEAR HEATUP/
l b j . COOLDOWN UMIT 3 3 C,C' - NUCLEAR (CORE CRITICAL) UMIT w ,- E 400 VESSEL O!SCONTINUITY j_ uuiTs
@ 312 P#
LO = = SHIFTED BELTUNE UMITS
/ f
[ (/) CURVES A,B,C ARE VAUD - i
; FOR 32 EFPY OF OPERATION 200 BOLTUP - -
70'r - CURVES A',B',C' FOR BELTLINE ARE LESS UMITING'THAN .a ' ' - s A,B,C AND ARE FOR INFO ONLY ' 0 , i i 0 100 200 3001 400. 500 600 MINIMUM REACTOR VESSEL METAL ' TEMPERATURE (OF) i.. Figure 7-2. Pressure-Temperature Curves Volid for 32 EFPY of Operation l 7-10 4 4 _
. . . ......J,..,..,.., ,,_,....,.-o,m,:. , , . _ , _ _ , . . . . . , .,,r , . . , . , , . . . _ , . , _ _ _ _ ,
- 8. REFERENCES
- 1. " Fracture Toughness Requirements," Appendix C to Part 50 of Title 10 of the Code of Federal Regulations, July 1983.
- 2. " Protection Against Non-Ductile Failure," Appendix G to Section XI of the 1989 A'SME Boiler & Pressure Vessel Code.
~
- 3. " Reactor Vessel Material Surveillance Progra'n Requirements," Appendix H to Part 50 of Title 10 of the Code of Federal Regulations, July 1983.
- 4. " Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," Annual Book of ASTM Standards, E185-82, July 1982.
o
- 5. " Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.
- 6. Caine, T. A., " Peach Bottom Unit 2 Surveillance Testing and Fracture
- Toughness Analysis " CE Report SASR 88-24, May 1988.
- 7. Hodge, J. M., " Properties of Heavy Section Nuclear Reactor Steels,"
Welding Research Council Bulletin 217, July 1976. ,
- 8. " Reactor Pressure Vessel," CE Purchase Specification 21 Allli ,
Revision 9. August 1970.
..u.
t 8-1
- 9. " Standard Methods for Notched Bar Impact Testing of Metallic Materials," Annual Book of ASTM Standards, E23-82, March 1982.
- 10. " Nuclear Plant Irradiated Steel Handbook," EPRI Report NP-4797,
, September 1986.
- 11. " Fracture Toughness Requirements,"
USNRC Branch Technical Position MTEB 5-2, Revision 1, July 1981. l 12. " Standard Methods of Tension Testing of Metallic Materials," Annual Book of ASTM Standards, E8-81.
- 13. " Ultrasonic Examination for Cracks in the Top He ad Flange," CBI Nuclear, Development Report 7,4-9047, December 1975.
- 14. " Ultrasonic Examination for Cracks in the Shell Flange," CBI Nuclear, Development Report 74-9056, November 1975.
a h I, J 4 4 0 m& '. t 4 8-2 4 ,
- - - . - - . . . . - , , , , ..,_,,_.._m.. . - , - -- - - * . - - - . , . - . - . - - ,
APPENDIX A IRRADIATED CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS ? Photographs of each Charpy specimen fracture surface were taken per the requirements of ASTM E185-82. The pages following show the frac.ture surface photographs along with a summary of the Charpy test results for l each specimen. The pictures are arranged with the materials in the order J of base, weld and HAZ f 3 l l l l ! .a. I l A-1 l l l 1 - . . . - - .
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