ML20235H690
| ML20235H690 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/31/1987 |
| From: | Branlund B, Caine T, Ranganath S GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20235H672 | List: |
| References | |
| MDE-103-0986, MDE-103-986, NUDOCS 8707150230 | |
| Download: ML20235H690 (112) | |
Text
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MDE-103-0986 hy DRF B13-01389 r
L1 Class I May 1987 i
O' COOPER NUCLEAR STATION REACTOR PRGSURE VESSEL SURVEILLANCE MATERIALS TESTING AND FRACTURE TOUGHNESS ANALYSIS
(< :
!O Prepared by:
b. bM-T. A. Caine, Senior Engineer
. Structural Analysis Services O
t.
Verified by: b _ N 1 *n dj -
p B. J. Branlund, Engineer Structural Analysis Services i
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Ct.
i Approved by 9
S. Ranganath, Manager
)
Structural Analysis Services
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GENER AL h ELECTRIC
~~D 8707150230 870706 PDR ADOCK.05000298 P
PDR.
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___.___________-_-________-__-_____a
D IMPORTANT NOTICE REGaRDING A
CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the use of the b
Nebraska Public Power District.
The information esstained 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.
(i The only undertakings of the General Electric Company respecting information in this document are contained in the contract governing Nebraska Public Power District Purchase Order ha. 221552 and nothing b
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
-1 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 owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
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CONTENTS b
ABSTRACT,
ix ACKNOWLEDGEMENTS x
D 1.
INTRODUCTION-1-1 2.
SUMMARY
AND CONCLUSIONS 2-1 2.1 Summary of Results 2-1 2.2. Conclusions 2-5 0
3.
SURVEILLANCE PROGRAM BACKGROUND 3-1 3.1 Capsule Recovery 3-1 3.2 RPV Materials and Fabrication Background 3-1 3.2.1 Fabrication History 3 3.2.2 Material Properties of RPV at Fabrication 3-2 3.2.3 Specimen-Chemical Composition 3-3 O-3.2.4 Initial Reference Temperature 3-3 3.3 Specimen Description 3-6 3.3.1 Charpy Specimens 3-6 3.3.2 Tensile Specimens 3-7 4.
PEAK RPV FLUENCE EVALUATION 4-1 3
4.1 Flux. Wire Analysis 4-1 4.1.1 Procedure 4-1 4.1.2 Results 4-2 4.2 Determination of Lead Factors 4-3 4.2.1 Procedure 4-4 4.2.2 Results 4-5
- I 4.3 Estimate of End-of-Life Fluence 4-5 5.
CHARPY V-NOTCH IMPACT AND HARDNESS TESTING 5-1 5.1 Impact Test Procedure
's - 1 5.2 Impact Test Resulta 5-2 5.3 Irradiated Versus Unirradiated Charpy V-Notch Properties 5-2 5 -3 5.4 Rockwell Hardness Testing 6.
TENSILE TESTING 6-1 6.1 Procedure 6-1 6.2 Results 6-2 6.3 Irradiated "ersus Unirradiated Tensile Properties 6-3 7.
DEVELOPMENT GF OPERATING LIMITS CURVES 7-1
7.1 Background
7-1 7.2 Non-Beltline Regions 7-1 7.3 Core Beltline Region 7-2 7.4 Closure Flange Region 7-2 7.5 Core Critical Operation Requirements of 10CFR50, Appendix G 7-3 111 g
3
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CONTENTS '(Continued) f l
O Page 1
7.6 Evaluation of Radiation Effects 7-4
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7.6.1 Measured Versus Predicted Surveillance Shif t 7-5 7.6.2 Modification of the Shift Relationship 7-5 7.6.3 Radiation Shift Versus EFPY 7-5 p
7.6.4 End-of-Life Conditions 7-6 l
7.7 Operating Limits Curves Valid to 12 EFPY 7-7 7.8 Reactor Operation Versus Operating Limits 7-7 i
8.
REFERENCES 8-1
- O APPENDICES A.
CHARPY V-NOTCH FRACTURE SURFACE PHOTOGRAPHS A-1
- B.
REVISIONS TO TECHNICAL SPECIFICATIONS B-1 j
b
- C.
REVISIONS TO SAFETY ANALYSIS REPORT C-1
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1
- Deleted. Under internal review.
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O, TABLES A
Table Title.
Page 3-1 Chemical Composition of RPV Beltline Materials 3-8
~
3-2 Results of Fabrication Test Program for Selected RPV 3-9 Locations 3-3 Plasma Emission Spectrometry Chemical Analysis of RPV 3-10 Surveillance Plate and Weld Materials 3-4 Identification of Charpy and Tensile Specimens Removed 3-11 from Surveillance Capsule 4-1 Summary of Daily Power History 4-6 4-2 Surveillance Capsule Location Flux and Fluence for 4-7 Irradiation 1/26/75 to 2/15/85 0
5-1 Qualification Test Results Using U.S. Army Watertown 5-4 Specimens (Tested in February 1986) 5-2 Charpy V-Notch Impact Test Results for Irradiated RPV 5-5 Materials 3-5-3 Significant Results of Irradiated and Unirradiated Charpy 5-6 V-Note.h Data 5-4 Rockwell C Hardness Test Results 5-7 6-1 Tensile Test Results for Irradiated RPV Materials 6-4 6-2 Comparison of Unirradiated and Irradiated 6-5 Tensile Properties at Room Temperature 7-1 Estimate of Upper Shelf Energy for Beltline Materials 7-9 j
7-2 Pressure-Temperature Values for Figure 7-5 (Curve A) 7-10 7-3 Pressure-Temperature Values for Figures 7-6 (Curve B) 7-12 and 7-7 (Curve C)
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ILLUSTRATIONS i
tb Figure Title Page j
j 2-1 Pressure Versus Minimum Temperature for Hydrostatic 2-6 Pressure Tests for Cooper 2-2 Pressure Versus Minimum Temperature for Non-Nuclear 2-7 g3 Heatup and Cooldown for Cooper 2-3 Pressure Versus Minimum Temperar.ure for Core Critical 2-8 l
Operation for Cooper 3-1 Surveillance Capsule Recovered from Cooper Reactor 3-12 p
3-2 Schematic of the RPV Showing Arrangement of Vessel 3-13 Plates and Welds 3-3 Fabrication Method for Base Metal Charpy Specimens 3-14 C
3-4 Fabrication Method for Weld Metal Charpy Specimens 3-15 3-5 Fabrication Method for HAZ Charpy Specimens 3-16 3-6 Fabrication Method for Base Metal Tensila Specimens 3-17 3-7 Fabrication Method for Weld Metal Tensile Specimens 3-18 3-8 Fabrication Method for HAZ Tensile Specimens 3-19 4-1 Schematic of Model for Two-Dimensional Flux Distribution 4-8 3
Analysis 4-2 Relative Fast Neutron Flux Variation With Angular 4-9 Position at the Vessel 5-1 Cooper Base Metal Impact Energy 5-8
'3 5-2 Cocper Weld Metal Impact Energy 5-9 5-3 Cooper Irradiated HAZ Metal Impact Energy 5-10 5-4 Cooper Base Metal Lateral Expansion 5-11 5-5 Cooper Weld Metal Lateral Expension 5-12 5-6 Cooper Irradiated HAZ Metal Lateral Expansion 5-13 6-1 Typical Engineering Stress versus Percent Strain for 6-6 Irradiated RPV Materials
)
vii
W ILLUSTRATIONS U
Figure.
Title Pace 6-2 Strength versus Test Temperature for Irradiated Base, 6-7' 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 6-9 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 Fracture 6-11 Appearance for Irradiated HAZ Metal Tensile Specimens l
7-1 Components of Operating Limits Curve for Pressure Test 7-14
.E (Curve A) for Cooper 1
7-2 Components of Operating Limits curve for Non-Nuclear 7-15 Heatup/Cooldown (Curve B) for Cooper 7-3 Components of Operating Limits Curve for Core Critical 7-16
.y Oper.aticn (Curve C) for Cooper 7-4 Adjusted Reference Temperature of Limiting Plate and 7-17 Weld, Based on Surveillance Specimen Test Results for Cooper 7-5 Pressure Versus Minimum Temperature for Hydrostatic 7-18 Pressure Tests for Cooper 7-6 Pressure Versus Minimum Temperature for Non-Nuclear 7-19 Heatup and Cooldown for Cooper 3
7-7 Pressure Versus Minimum Temperature for Core Critical 7-20 Operation for Cooper
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'G ABSTRACT 05 A surveillance capsule was removed from the Cooper Nuclear Power Plant reactor at the end of Fuel Cycle 9.
The capsule contained flux wires for neutron fluence measurement-and Charpy and tensile test specimens for D'
' material property evaluation.
A combination of flux wire testing and.
computer andlysis 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 C$
beltline.
The irradiation effects were projected to the vessel end-of-life.
The end-of-life conditions are less severe than the limits requiring vessel thermal annealing.
Pressure-temperature operating limits curves valid to 12 effective full power years were developed to July 1983
!O requirements of 10CFR50 Appendix G.
The pressure test curves were expanded to include curves for the beltline at 8, 10 and 12 EFPY, as well as a separate curve-for the bottom head region.
This provides greater flexibility in performing pressure tests.
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.. ACKNOWLEDGMENTS
.G:
Flux wire testing was performed by G. C. Martin.
L. S. Burns provided i
i the' evaluation of flux distribution. The Charpy V-Notch impact testing was d one.. by
.J.
L.
Bennett, and Rockwell hardness testing was done by
'b-G. H. Hendersen.
S. B. Wisner and G.
H. Hendersen performed the tensile
]
specimen-testing.
C. R. Judd performed the chemical composition testing.
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i 1.
INTRODUCTION
.O 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
'O 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 ASME Boiler and Pressure Vessel Code,Section III (Reference 2).
These documents contain requirements used to establish the O
pressure-temperature operating limits which must be met to avoid brittle fracture.
Appendix H of 10CFR50 (Reference 3) and ASTM-E185 (Reference 4)
O establish the methods to be used for surveillance of the reactor vessel materials.
In March, '1986 one of the Cooper vessel surveillance specimen capsules required by Reference 3 was sent to General Electric for testing after exposure to nine fuel cycles of irradiation, or 6.80 effective full
}
power years (EFPY) of operation.
The surveillance capsule contained flux vires for neutron flux monitoring and Charpy V-Notch impact test specimens and uniaxial tensile test specimens f abricated from the vessel materials nearest the core (beltline).
The impact and tensile specimens were tested 3
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 3
unirradiated properties from earlier tests.
Predictions of the RT
- "d NDT USE at end of reactor life (EOL) are made for comparison with allowable I
values in Reference 1.
Predictions of EOL properties were made based on l
surveillance test results, using Regulatory Guide 1.99, Revision 1
)
(Reference 5) as a guide.
1-1 tb
D Operating limits curves for the Cooper reactor vessel are O
presented in this report.
The curves account for recent new 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 0
developed consider the most 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 equivalent to 12.EFPY of operation, based O
on the results of the surveillance tests and the Reference 5 methods.
The pressure test curve was expanded to provide greater operating flexibility during pressure tests.
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1 1-2
o C;
2.
SUMMARY
AND CONCLUSIONS 2.1
SUMMARY
OF RESULTS Surveillance capsale 1 was removed from the Cooper reactor at the
. O~
end of Fuel Cycle 9 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 O
with 10CFR50 Appendix G (Reference 1) and Appendix G of the ASME Code (Reference 2).
The methods and results of the fracture toughness evaluation are preacnted in this report as follows:
D a.
Section 3:
Surveillance Program Background b.
Section 4:
Peak Vessel Fluence Evaluation O
c.
Section 5:
Charpy V-Notch Impact and Hardness Testing d.
Section 6:
Tensile Testing O
e.
Section 7:
Operating Limits Curve Development Photographs of fractured Charpy specimens are in Appendix A.
Suggested j
revisions to the Technical Specifications and Updated Safety Analysis 1
Report (USAR) are contained in Appendices B and C, respectively.
- The significant results of the evaluation are summarized as follows:
?
a.
Capsule 1 was removed from the 30* azimuth position of the reactor.
The capsule contained 9 flux vires:
3 each of pure copper (Cu), iron (Fe), and nickel (N1).
There were 36 Charpy V-Notch specimens:
12 each of plate material, weld material and heat affected zone (HAZ) material.
The 8 tensile specimens removed consisted of 3 plate, 2 veld and 3 HAZ metal specimens.
All specimen materials were positively identified as being from the vessel beltline.
f
.J, 2-1
- Appendices B and C Deleted.
Under internal review.
1
1 O
b.
The chemical compositions of the beltline materials were identified through a combination of literature research and l -
testing.-
The copper (Cu),
phosphorus (P) and nickel (Ni)'
contents were determined for all heats of plate material.
The values for the limiting beltline plate are 0.20%'Cu, 0.010% P and l
'O.68% Ni.
c.
Results from the fabrication program materials certification testing were located and adjusted to be equivalent to test
'O results done to current standards.
The initial RTET **I"* 8 I#
locations of interest in the vessel' were determined.
They are 14*F for the limiting beltline plate. -50*F for the limiting beltline veld, 20'F for the closure flange region and 28'F for D
the bottom head torus,.which is the component with the highest RT in the non-beltline regions.
ET d.
The flux wires were tested to determine the neutron flux at the surveillance capsule location. The fast flux (>1.0 MeV) measured 1.05x10' n/cm -sec.
Based on the flux wire data, the 2
was surveillance specimens had received a best estimate fluence of 1
2.3x10 n/cm2 at removal.
O e.
The vessel peak inner 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 45' to either side of the vessel quadrant references.
The lead factors for the surveillance capsule are 0.64 to the peak vessel surface and 0.87 to the peak 1/4 T depth location.
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3 1
D
O f.
,The maximum accumulated neutron fluence at the assumed vessel O
end-of-life (EOL) of 32 EFPY was determined at the peak 1/4 T location.
The maximum 1/4 T
vessel EOL fluence is 8
2 2
1.2x10 n/cm (best estimate) and 1.5x10 n/cm (upper bound).
b g.
The surveillance Charpy V-Notch specimens were impact tested at temperatures selected to define the transition of the fracture toughness curves of the plate,
- veld, and HAZ materials.
Measurements were taken of absorbed energy, lateral expansion and O
percentage shear.
Fracture surface photographs of each specimen are presented in Appendix A.
From absorbed energy and lateral expansion results for the plate and weld materials the following values are extracted:
index temperatures for 30 ft-lb, 50 ft-lb, O
and 35-mil lateral expansion (MLE) values and USE.
h.
The irradiated plate and weld impact energy curves are compared to unirradiated data from earlier studies to establish the RTg U
irradiation shifts for the surveillance program.
The plate material shows a 74*F shift.
The veld material shif t is 55'F.
Decrease in USE was est.imated for the surveillance materials also, showing a 13% decrease for the plate and 24% decrease for b
the weld, i.
Rockwell C hardness tests were done on one broken half of each l
Charpy specimen.
The average hardness for the base metal was O
15.3 HRC.
The weld metal specimens had an average hardness of 1
16.1 HRC. The average for the HAZ material was 15.5 HRC.
1 j.
The irradiated tensile specimens were tested at room temperature
,2 (76'P), reactor operating temperature (550*F), and estimated i'
onset to upper shelf temperature (185'F).
The results tabulated for each specimen include yield and ultimate tensile strength uniform and total elongation, and reduction of area.
The plate, I'
weld, and HAZ specimens behave similarly for all properties.
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The irradiated plate tensile test results are compared to i
G.
unirradiated data from the vessel fabrication test program.
The materials.show increased strength and generally-decreased ductility, as expected for irradiation embrittlement.
There was j
no comparison done for the veld material because unirradicted j
l0-data were not available.
I 1.
As' a part of the construction of the updated operating limits curves, the surveillance plate metal irradiation shif t in RT NDT O
was compared to predictions calculated with Regulatory Guide 1.99, Revision 1 (Reference 5).
The surveillance test shift of 17 2
74*F in plate material RT fr a fluence of 2.9x10 n/cm ET (upper bound) is greater than the predicted shif t of 31*F by a LO factor, of 2.39.
Accounting for the underprediction of surveillance plate ' material shift, the EOL adjusted reference P us irradiation shif t) of the temperature (ART = initial RT l
NDT limiting beltline plate is 171*F.
The predicted ART at EOL for O
the limiting beltline veld metal, assuming worst-case chemistry, j
1 is 157'F, so the plate is the limiting beltline material.
j i
m.
The USE at EOL is predicted using the methods in Reference 5.
1 C'
The veld metal USE is predicted to be 73 ft-lb at EOL.
The minimum plate USE is 90 ft-lb longitudinal at EOL, Branch Technical Position MTEB 5-2 (Reference 6) recommends 65% of the longitudinal USE as an estimate of transverse USE, so at EOL the 1) plate USE would be 58 ft-lb transverse.
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Operating ' limits curves were constructed for three reactor O.
conditions:
hydrostatic. pressure tests, non-nuclear heatup and cooldown, and core critical operation.
The curves are valid up to < 12 EFPY of operation.
The limiting regions of the vessel affecting'the curves' shapes are the core beltline (shifted to O'
account for irradiation),
the feedwater-nozzle and CRD penetration discontinuities, and the closure flange region.
The I
bolt preload and minimum permissible operating. temperatures on the curves of 80*F provide some. additional margin in the closure O.
flange region where a detectable flaw size of 0.24 inch is used instead of 1/4 T.
The operating limits curves for Cooper are shown in Figures 2-1 through 2-3.
1 0
2.2 CONCLUSION
S The requirements of Reference i deal basically with EOL vessel j
conditions and with limits of operation designed to prevent brittle 3'
fracture.
Based on the evaluation of surveillance testing, the following conclusions are dravn:
a.
The adjusted reference temperature for the plate material of 1
171*F is the limiting beltline E0L value.
This is below the Reference 1 allowable limit of 200*F, above which annealing is required. The EOL values of USE for the plate and weld materials are 58 f t-lb transverse and 7 3 f t-lb, respectively.
These are D
above the Reference 1 allowable of 50 ft-lb, below which 1
annealing is required.
Therefore, provisions for annealing the reactor vessel before completing 32 EFPY of operation need not be considered.
W b.
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 3.
acceptable relative to the limits in Figures 2-1 through 2-3.
Therefore, the only operating conditions for which the operating limits are a concern are those involving ' operator interaction, such as hydrotest and initiation of core criticality.
2-5
l
'O 1600 0
CURVE A BOTTOM EFPY HEAD REGION 81012 1400
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/ / /N du~eCURvES ee f
f 1/4 T FLAW.
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ADJUSTED AS I
1000 SHOWN:
f F
EFPYI ART (*F) g
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10 1
/
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12 no m
O 800 SAFE E
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(([
OPERATING u
REGION R'
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600
'O y110*F 1
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400 312 psig g
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l 200 BOLT PRELOAD l
TEMPERATURE = 80'F I
' FLANGE REGION
[ RTNDT = 20*F j
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l 0
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100 200 300 MINIMUM VESSEL VCTAL TEMPERATURE (*F1 Figure 2-1.
Pressure Versus Minimum Temperature for Hydrostatic Pressure Tests for Cooper i
I 2-6 l
1000 l
VAUD TO 12 EFPY g
B 1400 ADJUSTED BELTINE,
'O 1/4T FLAW, ART = 110'F\\
N 1200 r
(a 1000 0
IP E
@ra 83) 2 8
sn
/
600 D
SAFE OPERATING REGION 3
NON-BELTLINE FW NOZ2LE LIMITS, 1/4T FLAW, RTNDT = 28 'F 200 3
/
BOLT PRELOAD TEMPERATURE = 80'F FLANGE REGION RTNOT = 2 0'F' 0
O 100 200 300 MINIMUM VESSEL METAL TEMPERATURE (*F)
Fig,ure 2-2.
Pressure Versus tiinimum Temperature for Non-Nuclear Heatup and Cooldown for Cooper 2-7
. CP r
1600 g
, (g.
VALID TO 12 EFPY C
1400 7
(*
ADJUSTED BELTINE, 1/4T FLAW, ART = 110'F 1
1200
!D l
1000 IP
(
f a
E 800 E
e o
600-NON BELTLINE SAFE FW N0ZZLE LIMITS OPERATING PLUS 40'F,1/4T FLAW REGION RTNDT = 2 B'F
('
400 FLANGE REGICN RTNDT = 20*F; g
MINIMUM PERMISSIBLE 200 TEMPERATURE = 80'F f
PER 10CFR50 r
APPENDIX G 5
0 0
100 200 300 MINIMUM VESSEL METAL TEMPERATURE l'F)
- ^
Figure 2-3.
Pressure Versus Minimum Temperature for Core Critical Operation for Cooper 2-8 1
._____________________________-______a
O' 3.
SURVEILLANCE PROGRAM BACKGROUND
' 09 i
3.1 CAPSULE RECOVERY The Cooper reactor was shut down in September, 1984 for refueling and O'
maintenance.
The accumulated thermal power output was 5,913,690 mwd or 6.,80 EFPY.
The 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
'D loaded specimen holder. Each capsule receives equal irradiation because of-core symmetry.
During the outage, Capsule 1 at 30' 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, O-California, j
Upon crrival at Vallecitos, the capsule was examined for identification.
The reactor code of 29 and the basket code'of 13 from J
O Reference 7 were confirmed on the capsule, as shown in Figure 3-1.
The capsule contained three Charpy specimen packets and four tensile specimen tubes.
Each Charpy packet contained-12 Charpy specimens and 3 flux wires.
]
The four tensile specimen tubes contained eight specimens.
The specimen E
gage sections were protected by aluminum sleeves, and during removal of the sleeves, the threaded ends of the specimens were slightly damaged.
The threads were later chased with a die-hex rethreading tool.
The gage sections of the tensile specimens were not damaged during removal.
j 3
1 1
3.2 RPV MATERIALS AND FABRICATION BACKGROUND j
l 3.2.1 Fabrication History j
3 The Cooper RPV is a
218-in.
BWR/4.
It was constructed by Combustion Engineering to the 1965 ASME Code with Addenda up to and including Winter 1966.
The shell and head plates are ASME SA-533 Grade B, j
Class 1 low alloy steel (LAS). The nozzles and closure flanges are 3-1
's
1 i
~
ASTM A508 Class 2 LAS and the closure flange bolting materials are D:
The fabrication process employed quench and i
temper ' heat treatment immediately af ter hot forming, then submerged arc welding and post-weld heat treatment.
The post-weld heat treatment was
. typically 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> at 1150*F 25'F.
The arrangement of plates and welds O
relative to the core beltline and various nozzles is shown in Figure 3-2.
3.2.2 Material Properties of RPV at Fabrication lk 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 lC records, and has been incorporated into DRF B13-01389.
Available data for the beltline welds was obtained from Combustion Engineering (Reference 8),
l but as-welded chemical composition data for the seam welds was not available.
The surveillance weld metal specimens were fabricated with the l9 same 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.
Therefore, the limiting chemistry combination of 0.35% Cu and 0.012% P was assumed for the beltline welds.
O 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.
O Properties of the belt 31ne materials and other locations of interest are presented in Table 3-2.
The Charpy data collected were used to establish the RT va ues fr ea vessel
- compoxnt, as descrned in NDT Subsection 3.2.4.
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3-2
G 3.2.3 Specimen Chemicci Composition i
O Samples were taken from tensile specimens J64 and J6L (base) and from J74 and J7D (weld) after they were tested.
The tensile specimens were fabricated from the same plates and weld as were the Charpy specimens, as O
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 O
calibrated with a standard solution containing 700 ppm Fe, 8 ppm Mn, 2 ppm Cu, 5 ppm Ni, 5 ppm Mo, 5 ppm Cr, 1 ppm Si, 1 ppm Co, and levels of perchloric acid and lithium consistent with the test.
The calibration for phosphorus was done by analyzing a series of seven National Bureau of 0
Standards steels with known quantities of phosphorus.
The chemical composition results are given in Table 3-3.
3.2.4 Initial Reference Temperatures
'"E*#
The requirements applicable to establishing the RTNDT 1966 Edition of the ASME Code can be summarized as follows for the RPV:
A a.
Test specimens shall be longitudinally oriented Charpy V-Notch specimens.
b.
At the RTNDT, n impact test result shall be less than 25 ft-lb, and the average of three test results shall be at least 30 ft-lb.
I c.
Pressure tests shall be conducted at a temperature at least 60*F l
above the acceptable RT fr e vessel, j
NDT I
The current requirements for establishing RT are significantly NDT l
different.
For plants constructed to the ASME Code after Summer 1972, the I
requirements are as fo11cvs:
Charpy V-Notch specimens shall be oriented normal to the rolling a.
1 direction (transverse).
l 3-3
\\
l j
O b.
RT is defined as the higher of the dropweight NDT or 60*F ET 0
below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion are met, c.
Bolt-up in preparation for a pressure test or normal operation 1 0 shall be performed at or above the RT r lowest service ET temperature (LST), whichever is greater.
Reference 1 states that for vessels constructed to a version of the
- G.
ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses must be supplemented in an approved manner. General Electr1e has developed methods for analytically converting fracture toughness data for vessels constructed before 1972 to comply with current requirements.
f.
The methods used have been presented to the NRC in about 10 Final Safety Analysis Report updates.
These were reviewed on a case by case basis and.
in ee-h case approved (e.g., La Salle 1 and 2 and Nine Mile Foint 2).
l These methods and example RT Calculati n8 f r Ve88e1 plate, weld, weld ET O
HAZ, forging, and bolting material are summarized in the remainder of this subsection.
C'alculated RT values for selected RPV locations are given NDT in Table 3-2.
Q.
For vessel plate material, the first step in calculating RT is to establish the 50 ft-lb transverse test temperature given longitudinal test 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 D
f t-lb energy to 50 f t-lb.
For example, for the six beltline plates the limiting combination of test temperature n;.d Charpy energy from Table 3-2 is 33 ft-lb at
+10*F for Plate G-2803-1.
The equivalent 50 ft-lb longitudinal test temperature is:
3 T
= 10*F + ((50 - 33) ft-lb
- 2*F/fc-1b] = 44*F 50L The transition from longitudinal data to transver::e data is made by adding 3,
30*F to the test temperature.
In this case, the 50 ft-lb transverse Charpy 74 F.
The RT s de greater of M or test temperature is T
=
50T NDT (T
- 60*F).
From Table 3-2, the NDT for G-2803-1 is -10*F.
Therefore, 50T the RT f r the core beltline plate is 14'F.
ET 3
3-4
M For vessel weld material, the Charpy V-Notch results are usually O
limiting in establishing RT The 50 ft-lb test temperature is NDT.
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 weld Charpy V-Notch energy from Ta' ole 3-2 is 56 f t-lb at 6
+10*F for veld 2-233, which is above the 50 ft-lb requirement, so T
= 10*F.
50T O
As shown in Table 3-2, there rre no NDT data available for the veld metal.
As long as the resultir.g RT s n ess tb.an
-M, de RT s
NDT NDT established as (T
- 60'F), or RT
= -50*F for the core beltline welds.
50T NDT 0
For the vessel weld HAZ material, the RT is assumed the same as for NDT the base material since ASME Code veld procedure qualification test requirements and post-weld heat treatment data indicate this assumption is va'id.
7)
For vessel forg:Ing material, such as nozzles and closure flanges the method for establishing RT is the same as for vessel plate material.
NDT Plate G-2806-1 of the lower head bottom torus, listed in Table 3-2, has a k
calculated value of T
= 88'F.
The NDT is -10*F.
Therefore, with the 50T RT being the greater of NDT or (T
- 60'F), the RT is 28'F.
This g
50T NDT is the highest RT f the non-beltline regicns of the vessel.
/
For bolting materiel, the current Code requirements define the LST as the temperature at which transverse Charpy V-Notch energy of 45 f t-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 A3ME Code at construction are applied.
As shown in Table 3-2, the limiting Charpy V-Notch energy for bolting material is 35 ft-lb at
+10'F.
The current requirements are not rne t, so the construction Code requirements are used.
The LST is defined as 60*F above tne tettperature at which 30 ft-lb Charpy V-Notch energy is achieved.
Therefore, the LST of the closure bolting l
material is 70*F.
l 3-5 i
I lb
G 3.3 SPECIMEN DESCRIPTION The surveillance capsule contained 36 Charpy specimens:
base metal (12), weld metal (12), and HAZ (12).
There were 8 tensile specimens: base metal (3), weld metal (2), and HAZ (3). The 9 flux wires recovered were iron (3), nickel (3) and copper. (3).
The chemistry and fabrication history for the Charpy and tensile specimens are described in this section.
3.3.1 Charpv Specimens The fabrication of the Charpy specimens is described in the Surveillance Test Program description given in Reference 9.
All materials used for specimens were beltline materials from the lower intermediate shell course.
The base metal specimens were cut from plate G-2802-2 from the beltilne.
The chemical analysis of this heat of low alloy steel is in Table 3'1.
The. test plate was heat treated for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 1150*F 1 25'T 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.
Specimens were machined from the 1/4 T and 3/4 T positions in the plate, in C'
the longitudinal orientation (leng 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.
3 The weld metal and HAZ Charpy specimens were fabricated from trim-off pieces of plates G-2802-1 and G-2802-2 that were welded together with the I
same weld process used for longitudinal seam weld 1-233 in the RPV beltline.
The chemical analyses of the plates are given in Table 3-1.
As-welded chemistry data were not reported in Reference 8 for the veld metal.
The weld metal chemistry from two specimens is presented in Table 3-3.
?
3-6
O The welded test plate fer the weld and HAZ Charpy specimens received a 0+
l heat treatment of 1150*F 2 25'F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> to conservatively simulate the fabricated condition of the RPV. The weld specimens and HAZ specimens were fabricated as shown in Figures 3-4 and 3-5, respectively.
The base metal orientation in the veld and HAZ specimens was longitudinal.
Contained in Table 3-4 are the identifications of the veld metal and HAZ Charpy specimens from the surveillance capsule.
3.3.2 Tensile Specin' ens
,(>
Fabrication of the surveillance tensile specimens is described in Reference 9.
The chemical composition and heat treatment for the base, weld and HAZ tens 11es are the same as those for the corresponding Charpy specimens.
The identifications of the base, veld, 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 section.
I?*
The base metal specimens were machined from material at the 1/4 T and 3/4 T depth in plate G-2802-2.
The epecimens, oriented along the plate i
rolling direction, were machined to the dimensions shown in Figure 3-6.
l0 The gage section was tapered to a minimum diameter of 0.250 inch at the center.
The veld metal tensile specimen material was cut from the velded 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 Figure 3-8 The specimen blanks were cut from the welded test plate such that the gage seccion minimum diameter was machined at the weld fusion approximately half weld metal and line.
The finished HAZ specimens are
- ).
i l
half base nietal oriented along the plate rolling direction 3-7 y
_. O o
556 687 6
M 444 444 4
000 000 0
O i
838 583 7
N 676 657 9
t 000 000 0
necre Pi 1 1 2 420 1
S 222 222 2
t O
h 000 000 0
g e
e S
i l
l L
e A
W b
b a
a I
R y
857 674 l
l 7
0 E
b 1 1 1 1 1 1 i
i 0
1 0
0 T
S 000 000 a
a A
n v
v M
o 000 000 a
a 0
0 i
O E
t t
t o
o N
i I
s n
n L
o T
p 099 000 3
6 L
m 1 00 1 1 1 1
1 E
oP 000 000 0
0 B
C 1
000 000 0
0 V
3 PR C
e l
F b
O 6
c n
1 02 555 T
N M
333 332 1
O I
1 1 1 1 1 1 1
T I
S O
P C
M 7
O 033 21 3 C
C 222 222 1
AC I
M E
x x
x x
H u4 u8 h u4 u9 l 2l 0 tl 2 l 6 C
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47 4
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222 222 2
2 7 2 1
t CCC CCC 1
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m T) 000 000 a
a 0
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n n
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yyb 3, 5, 4, 6,5 7, 5,
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7, 3
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004 803 9
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6 9
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6 4, 7, 4, 6,7 7, 6,
5, 4,
7, 8,
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C CE(
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ean) 555 503 2
l t o%
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899 999 8
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r 42 96 7
8 A
t e 47 4 71 7 2 7 1 8 7
6 6
0 7
F ab 707 030 1 3 23 2
6 3
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6 1
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W X
2 2
3 9
it o t o C
A A
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O N
CCC CCC L
HL l
STLU S
1 23 7 1 2 4
1 E
.r 3
0 9
0 2
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t e 333 1 22 3
4 1
0 1
1 6
nb 000 000 2
2 0
8 8
8 0
e m 888 888 8
2 2
2 8
d u 222 2 2 2 2
1 2
2 I N G
G G
o GGG GGG G
C s
e e
t d
e s
t a
l t
u a
i e
a r
l d
W l
s o
n P
e P
e u
T o
ms l
e g
r i
l r e a
n l
n e
o d
t l
et n
d i
l a
g T
a a
e t a i
l l
e l
n e
c e
h nl d
e t
h F
a d
H o
n S
I P u
W l
S l
a L
i t
e l
F e
m l
r rl i
h B
r e
H o
t e
el g
t e
s d
t l
w we n
r n
p s
a p
t B
L LS L
G N
U V
le o
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o oh o
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T B
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0 l
Table 3-3
. Base Metal Weld Metal Weld Metal-Base Metal
' Element Tensile J64 Tensile J6L Tensile'J74 Tensile J7D Mn 1.23 1.30 1.26 1.37 P
0.007 0.006 0.012 0.012 Cu 0.22 0.22 0.23 0.22 Ni 0.77 0.78 0.75 0.74
. C..
Mo 0.51 0.50 0.55 0.52 Cr 0.11 0.11 0.04 0.04 G
a 9
5 I
3-10
')
_______________o
O Table 3-4 0
IDENTIFICATION OF CHARPY AND TENSILE SPECIMENS REMOVED FROM SURVEILLANCE CAPSULE a
Charpy Specimens Base Weld HAZ O
EP3 EYA J3A EP7 EYJ J44 EPA EYL J4A EPC EYM J4E
(?
EPM EYU J4L ETB EYY J4M ETK J16 J4U ETT J1B J4Y
.EU1 J23 J51 EU6 J24 J52 EUC J2B J5L EUD J2K J5M O
a Tensile Specimens
~,:)
~
J67 J7D JA6 J6L JAP 3
All specimens have a dot over the center identification digit.
3-11 4
O i
O HE ACYOR CODE
{
16 e e' 8 e e 4 e e > 1 + 4 + 8 + 16 = 29 2 e 1
e'
(/
)
U I
< s,#
' 'C D,~~,9[TQ9y yf%f
}
l l-
- ^i f!:
., E. >.i.v-1, G
7
."Y lyf 1%Wepf@ '.
t
=-
..,Q:&g g ff$
%y 4 n'% MMik$ N; l
, ~..,
'e*nb lh'j vvW.%4rpi.tk
,f*'8-g h
uJ<
irth 4
!~
- t o
(;
4
.gp jez
- b i
2, b
4lr.., jggj{sgy,'%
)
i
.? x i
i y';.
r.7;
- v. a
.s lv$
n
- 3. w -
h[k,, !?hl,,..e "
' /
s
- f[?0 f
- e. %,dc'u f W 'N --
j (l:
f a m % a r @ v $ %j g/ f kl!
aA$S$ ddf
$}
if k r G ! h ;u_m,e sl$}f?f~Ejq h }rK G.g.g g yl
- y zj'i'.lf'~
c h aef m ai a WSe m e s -
0 h
1 CAPSULE CODE 8
oI L 1 + 4 + 8 = 13 2
1 e e, b
CAPSULE 1 g
Figure 3-1.
Sur.'e111ance Capsule Recovered from Cooper Reactor 3-12
l G.
l$
4
=
lcp TOP HE AD ENCLOSURE CL U*I l
gy risiriisi/// - ://
- ///sirisizisi,1,1
///,11/1:11, REG lON Lwwwwwwwwwwwi'wwwnwwn I
l I
Iis Q
l Q
l s
UPPER SHELL z
h O
. ur n l
a i
k f h g
l
'b UPPE R INTERMEDIATE l
SHELL l
l N
l LOWER l
.g b
{
thTER MEDI ATE l
o.2802 2 G 28017 s SHELL l
r G 28021 CORE l /
BELTLINE LONG1TUDINAL l
l 1 233 b
h.
CIRCUMF EREhTa AL
[
l [ 2 233
(
A l @B>
N l
G 2803 2
'ER g
} G 2803 3
- ~d$.1 l
l BOTTOM HEAD f
I ENCLOSURE N
/_
r i
Figure 3-2.
Schematic of the RPV Showing Arrangement of Vessel Plates and Welds 1
4
~
3-13 i
l
C9 A
D' gc;n "
pidM o
b p,Ot b
G-2802 2 1/2 in, s\\
(HE AT C2307 2)
O gg
\\ g\\
K s\\\\
L n.
s \\, \\k 5/8 i
\\
\\\\
b
\\ \\\\
\\\\ \\\\
4 SLABS g\\ g TOTAL g\\
\\
il r
g 3/4T 15 PIE CES
\\gg,g PER SLAB
\\\\
jr O
sff
=
=
\\
9 O
F M
9C oeg s O deg 10 min +
MARKED END y UNM ARKED EhD j
\\
4 2.165t 0.015 221(2 deQ O 010:0.001R 45 deg
+ 0.394
+
4tg j;
90 deg i 0 deg 10 min sW ALL (, ORN E R S 0 394 j
o 3,3 4
6 n
3 Figure 3-1.
Fabrication Method for Base Meta: Charpy Specimens
'4 3-14
l i
M ROLLING DIRECTION G 28021 l
(HE AT C23312) i V
5/8 en.
o WELD 1233 a
y G 2802 2 p/
(HEAT C2307 21 O
i 1
\\
h 6 SLABS I
~
b.
i U
l
=
- 21/4 in.-
q N
\\\\
\\\\ \\
xN
\\\\\\
x N
N O
so SPECIMENS
- ERsLAB
\\
\\
wACHINED AS N
\\\\
8% FIGURE 3 3
\\\\\\
x N
N
\\
\\
\\
Y\\
g\\
N N_
\\ 7Ng s/s.n\\
l
-^-
3 Figure 3-4.
Fabrication Method for Weld Metal Charpy Specinens i
3-15
(h b'I ROLLING DIRECTION C 28021 l
(HE AT C23312)
E-WE LD 1233 c.
1 G-2802 2 I
(HEAT C2307 2) p SCRAP # s N
6 SLABS -
=
]
I, c
N k
l 6-1/2 in, f
{
{
SCRAP N
f
- i/2 e M
a
\\
g
\\
\\' %
\\
\\\\
L N
NN x
xN ne x
N N
N N
N x
N N
\\
\\\\
N x
x x
$?B e xv v it i i i
10 SPECIMENS PER SLAB MAcmNED M 21/4 in.
AS IN FIGURE 3-3 WITH NOTC=
l CENTERED ON WE LD FUSIO:s LINE l ' i
',*1 Figure 3-5.
Fabrication Method for HAZ Charpy Specimens 3-16
' D l
D ROLLING DIRECTION 5
\\
g G 2802 2
\\
(HE AT C2307-2)
\\
\\
)
\\
\\
]
o
\\
\\
\\
\\
10 SLABS
\\
x x\\
\\
\\5/8 in.
o
\\
h u N N u
~
s
- '2 in.
N 1/4T b s
N G1/2 in, 3/4T N
o s
N s
n C:157_?
0
(\\
31/8 in.
o o
1.00020.005 GAGE LENGTH 7/16 - 14 UNC - 2A 3/8R
~
BOTH ENDS (TYP) 1/2 d c...
fi AGE MARKS
! l II j
il r
T 5
i o
j j D' D
D.
3 30DEG i ; ij q
g g
)
T TYP r
374 s
1/16 +
T/2
,1 1/2 ra. R E DUCE D,8 c
r-gg fp jg SE CTION 4
3:1/16 j
'?
NOTES:
- 1. D = 0.250:0.001 DIAM AT CENTER OS
- EDUCED SECTION
- 2. D' = ACTUAL "D" DI AM + 0.002 TO O 005 AT ENDS OF REDUCED SECTION, TAPEA!NG TO *C AT CENTER Figure 3 6.
Fabrication Method for Lue Metal Tensile Specimens 3-17
O i-
~
i
(
i l
l C-1
- b
" CUT WELD OUT d
1 in.
OF TEST PLATE
(,
WELC.233
\\
c 3
, (,A p
\\
)
Y,s 31/4 in.
6 SPECIMENS 3
PER SLAB q3
\\
DIMENSIONS 3 SLABS q
AS IN FIGURE 3-6 gf i
M SCRAP 61/2 an, D
J2 o.
L JL_
1 I
h Figure 3-7.
Fabrication Method for Weld Metal Tensile Spec 6 ens 1
l J
3-18
_ _ _ _ - _ _ - __- __--_ _ - a
O ROLLING DlHECTION G 2B221 (HEAT C23312)
G 2802 2 (HE AT C2307 2)
xx N">r
'O l
61/2 an.
N I
O
{
5 3 SLABS
}
l k
)
10'
~
c
'(
c 31/8 an. M g
g 1/2o g
_I_
A
[ 6 SPECIMENS PER s'
s SLAB, DIMENSIONS 3
/
AS IN FIGURE 3-6 t
N g SCRAP
(
\\
~
l
(
h Figure 3-8.
Fabrication Method for HAZ Te.sile Specimens
?)
3_19 i
- CY 4.
PEAK RPV FLUENCE EVA'LUATION
- o Flux vires were analyzed to determine flux and fluence received by the surveillance capsule.
An analysis combining two-dimensional.and one-dimensional flux distribution computer calculations was evaluated to iO 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 WIRE ANALYSIS
'O 4.1.1 Procedure The surveillance capsule conta1ned nine flux wires: three each of
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 content and each copper wire for Co-60 at a calibrated 4-cm 3
source-to-detector distance with 100-ce and 80-ce Ge(L1) detector system.
The gamma spectrometer was calibrated using NBS material. The nickel wires were not analyzed since the 70.8 day half-lived Co-58 in the nickel wires had decayed to insignificant levels since capsule removal.
Since the D
nickel wire results are generally the 2 east reliable, the lack of nickel wire data does not significantly affect the overall results.
To properly predict the flux and fluence at the surveillance capsule 3
from the activity of the flux vires, 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.
):
.y.
4-1 3
~
O From the flux wire activity measurements and power history, reaction C) rates for Fe-54 (n,p) Mn-54 and Cu-63 (n,a) Co-60 were calculated.
The
>l MeV fast flur reaction cross sections for the iron and copper wires were estimated to be 0.186 barn and 0.0032 barn, respectively.
These values were obtained from measured cross section functions determined at c
Vallecitos from more than 65 spectral determinations for BWRs and for the General Electric Test Reactor using activation monitor and spectral unfolding techniques.
These data functions are applied to BWR pressur.e vessel locations based on water gap (fuel to vessel wall) distances.
The
'O 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 Resu?.ts 0
The measured activity, reaction rate and determined full-power flux results for the surveillance capsule are given in Table 4-2.
The >l MeV 9
9 and >0.1 MeV flux values of 1.05x1'0 and 1.70x10 n/cm -see f rom the flux 3
monitors were calculated by dividing the reaction rate measurement data by the appropriate cross sections.
The corresponding fluence
- results, 17 17 2
2.3x10 and 3.6x10 n/cm for >l MeV and >0.1 Mev, respectively, were obtained by multiplying the full-power flux density values by the product 8
9 of the total seconds irradiated (2.82x10 see) and the full-power fraction (0.760).
Generally, for long-term irradiations, dosimetry results from copper g
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 results fairly consistent with the copper reaction despite the shorter half-life of 312.5 days for Mn-54.
Consistency in results by indicates en accurate power-history evaluation and a consistent core radial power shape.
n 4-2
6 The accuracies of the values in Table 4-2 for a 2a deviation are
. (3 estimated to be:
1 5% for dps/g (disintegrations per second per gram)
't 10% for dps/ nucleus (saturated) 9 25% for flux and fluence >l MeV 1 35% for flux and fluence >0.1 MeV A set of flux vires from Cooper was evaluated by General Electric 2
,g in 1977.
The
>l MeV flux was 2.1x10 n/cm -sec.
Differences in the current test inputs and methods would decrease that number to 1.6x10' n/cm -sec.
The result from this study of 1.05x10 n/cm -see is 2
significantly lower than even the adjusted flux value.
Past experiences with dosimeters removed after one fuel cycle have generally shown high g
results when compared to the ten-year dosimetry. The first 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 y
fuel bundles had unusually high power levels. The ten-year results reflect a long period of typical operation, so the results from the current test are considered to be more accurate.
l 4.2 DETERMINATION OF LEAD FACTORS 9
The flux vires detect flux at a single location.
The wires will therefore reflect th power fluctuations associated with the operation of g
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 the peak. flux. These lead factors are a function of the core and vessel geometry and of the distribution of bundles in the core.
Lead factors were generated for the Cooper geometry and operating s
history. The methods used to calculate the lead factors are discussed below.
+
4 -3
~
p, 4.2.1 Procedure 0
Determination of the lead factors for the RPV inside wall and at 1/4 T L
depth was done using 'a combination of one-dimensional and two-dimensional l
finite element computer analysis. 'The two-dimensional analysis established' l
gb'
.the relative fluence in the azimuthal direction at the vessel surf ace and 1/4 T depth.
A series-of one-dimensional analyses were done to determine 3
the core height of the ' axial flux peak and its relationship to the surveillance capsule height.
The combination of azinutl.a1 and axial lg 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 computer program was used to solve the Boltzman
(,
transport equation using the discrete ordinate method on an (R,0) geometry, assuraing a fixed source.
Eighth core symmetry was used with periodic boundary conditions at 0*
degree and 45*.
Neutron cross sections were determined for 26 energy groups, with angular-scattering approximated by a 9
- third-order Legendre expansion.
A schematic of.the two-dimensional vessel' model is shown in Figure 4-1.
A total of 43 radial elements and 30 azimuthal elements were used.
The model consists of an inner and outer core region, the shroud, water regions inside and outside the shroud, and the vessel wall from the inside surface to the 1/4 T depth.
Flux as a.
p function of azimuth was calculated, establishing the azimuth of the peak flux and its magnitude relative to the flux at the wires' location of 30*.
,g-The one-dimensional computer code (SN1D) was used to calculate radial flux distribution at several core elevations at the azimuth angle of 45',
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
- i the surveillance capoule elevation.
y',
4-4 9
_______...______________________._________j
0 4.2.2 Results O
The two-dimensional calculation indicated the flux to be a maximun 45' on either side of the RPV quadrant references (0*,
90*, etc.).
The peak closest to the 30* location of Capsule 1 is at 45*.
The distribution calculations establish the lead factor between the surveillance capsule g,,
i location and the p~eak location at the inner vessel wall. This lead factor is 0.64.
The fracture toughness analysis done is based on a 1/4 T depth flaw in the beltline region, so the attenuation of the flux to that depth is considered.
The resulting lead factor from the capsule to the 1/4 T (3
depth at the peak location is 0.87.
4.3 ESTIMATE OF END-OF-LIFE FLUENCE C
The fluence at end-of-life (EOL) is estimated by taking the upper bound of the measured flux frem Table 4-2 and the 1/4 T lead factor.
The 1.01x10' seconds.
The period assumed to represent EOL is 32 EFPY, or resulting EOL fluence is:
g, 2
9 18 2
(1.05x10 n/cm -s)(1.25)(1.01x10 s)/0.87 - 1.5x10 n/cm,
(s 0
4-5 s
i
lO I
Table 4-1 1
]
G I
SUMMARY
OF DAILY POWER HISTORY l
Operating Percent of Days Between Cycle Cycle Dates Davs Full Power Cycles 1
7/1/74 - 9/17/76 809 0.627 j
56 l
2 11/12/76 - 9/18/77 310 0.760 34 3
10/22/77 - 4/1/78 161 0.750 29 0,'
4 4/30/78 - 4/7/79 342 0.839 5
5/7/79 - 3/1/80 299 0.820 94 6
6/3/80 - 4/21/81 322 0.892 7
6/7/81 - 5/22/82 349 0.766 46 8
7/7/82 - 4/30/83 297 0.903 125 9
9/2/83 - 9/15/84 379 0.700 3268 0.760 (average)
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1 CORE INTERIOR:
2 ELEMENTS
[
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//((f VESSEL WALL 1/4 T
]
45 4 ELEMENTS
\\
I J
30 ELEMENTS IN AZIMUTHAL DIRECTION L
0 1 = CORE INTERIOR FUEL 2 = CORE EXTERIOR FUEL
)
l3 Figure 4-1.
Schematic of Model for Two-Dimensional Flux Distribution Analysis
-s
~
4-8 J
0 1.8
.Gt-1.7 OJ 1.6 1.5
($
-I a
i.4 g
r h
f 1.3 25' A
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1 1,1-I U
- 3 j
1.o E
0.9 l
- G o.s 0.7
.')
0.6 0
to 20 30 l -l ANGUL AR POSITION (deg) l Figure 4-2.
Relative Fast Neutron Flux Variation With Angular Position at the Vessel 4-9 L
(b 5.
CHARPY V-NOTCH IMPACT AND HARDNESS TESTING G
The 36 Charpy specimens recovered from the surveillance capsule were impact tested at temperatures selected to establish the toughness transition and upper shelf of the irradiated RPV materials.
Testing was b
conducted in accordance with ASTM E23-82 (Reference 10).
After impact testing, Rockwell C hardness testing was performed on the broken specimen halves per ASTM E18-79 (Reference 11).
O 5.1 1MPACT TEST PROCEDURE The testing machine used was a Riehle Medel PL-2 impact machine, serial number R-89916. The pendulum has a maximum velocity of 15.44 ft/see i
D 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 74.1 f t-lb and 13.9 f t-lb at a test temperature of
-40*F.
According to Reference 10, the test apparatus O
averaged results must reproduce the Watertown design values within an accuracy of 15% or 11.0 ft-lb, whichever is greater.
The successful qualification of the Riehle machine and operator is summarized in Table 5-1.
O Charpy V-Notch tests were conducted at temperatures between -20*F and 400*F.
For tests between 32*F and 212*F, the temperature conditioning fluid was water.
Dichloromethane was used at temperatures below 32'F.
3 Above 212'F, a silicone oil was used.
Cooling of the conditioning fluids i
was done with liquid nitrogen, and heating by an immersion heater.
The j
fluids were mechanically stirred to maintain uniform temperatures.
The fluid temperature was measured by a chromel-alumel thermocouple and a O
copper-constantan thermocouple.
These were calibrated with boiling water (212*F), and ice water (32*F).
Once at test temperature, the specimens vere manually transferred with centering tongs to the Riehle machine and impacted within 5 seconds.
I u
i 5-I
.__.____.___.___________d
QF For each Charpy V-Notch specimen tested, test temperature, energy absorbed, lateral expansion, and percent shear were evaluated.
Lateral expansion and percent shear were measured according to Reference 10 methods.
Percent shear was determined with method two of Subsection'11.2.4.3 of Reference 10, which is a comparison of the fracture Ik surface appearance with the reference fracture surfaces in Figure 15 of
~
Reference 10.
5.2 IMPACT TEST RESULTS j
O 1
Twelve Charpy V-Notch specimens each of base metal, weld metal and HAZ i
were tested at tempera *.ures selected to define the toughness transition and upper shelf portions of the fracture toughness curve.
Absorbed energy.
I 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 freehand fit with best-estimate S-shaped curves characteristic of fracture toughness transition curves.
Photographs were b
taken of the fracture surfaces for each specimen.
The fracture surface photographs were used to evaluate percent shear.
The photographs and a summary of test results for each specimen are contained in Appendix A.
U 5.3 IRRADIATED VERSUS UNIRRADIATED CHARPY V-NOTCH PROPERTIES As a part of the RPV fabrication test program, Charpy V-Notch testing was done at various temperatures on the unf rradiated RPV plate materials.
Data for the beltline plate from which the base metal specimens were fabricated, (place G-2802-2) were recovered from QA records.
The impact energy and lateral expansion data are plotted in Figures 5-1 and 5-4, respectively, along with the irradiated data from this test.
The 3
surveillance weld material was tested in the unitradiated condition as part of an EPRI effort in 1983 (Reference 12).
The unirradiated and irradiated data for impact energy and lateral expansion are plotted on Figures 5-2 and 5-5, respectively.
J 5-2 i
i
- g The curves of irradiated and unirradiated Charpy V-Notch properties are used to estimate the values in Table 5-3:
30 ft-lb, 50 f t-1b and 35 MLE index temperatures and USEs.
The RT shift values are determined NDT as the change in the temperature at which 30 ft-lb impact energy is achieved, as required in Reference 4.
In previous experience, the shift in the transition curve has been approximately equal at the 30 ft-lb and 50 ft-lb levels. The values in Table 5-3 show this to be the case for base metal. However, the 50 ft-lb index temperature shift for the veld. metal is significantly higher than at 30 ft-lb.
This is due to the decrease in USE altering the shape of the Charpy curve.
The shifts from initial to adjusted RT for the base and weld metals g
are compared in Section 7 to analytical values calculated according to Regulatory Guide 1.99, Revision 1.
5.4 ROCKWELL HARDNESS TESTING After the Charpy specimens were tested, one broken half of each specimen was subjected to Rockwell hardness
- testing, according to ASTM E18-79 (Reference 11).
The test used for the surveillance materials was the Rockwell C test, which employs a diamond sphero-conical penetrator with a minor load of 10 kgf and a major load of 150 kgf.
The machine used was a calibrated Wilson Rockwell Hardness Tester. As a further calibration before testing, a test block with a reference hardness of 35.0 1 1.0 HRC was tested.
The three values taken vern 34.2, 34.4 and 34.2 HRC, for an average of 34.3 HRC, which is acceptable.
Three indentations were made on one half of each specimen and the results were averaged to develop the hardness values in Table 5-4.
The half chosen for testing was the half with the specimen identification stamped in the end.
The indentations were made on the same side of each specimen in a group approximately 3/8 inches from the fracture surface.
The results show little difference in hardness between veld, base metal or HAZ.
5-3 7
a
- (s Table 5-1 O
QUALIFICATION TEST RESULTS USING l
U.S. ARMY WATERTOWN SPECIMENS (TESTED IN FEBRUARY 1986)
C Energy Absorbed Qualification Test Test Temperature Mechanical Gage Specimen Identification
(*F)
(ft-lb)
Co EE30427
-40 72.0 EE30075 74.5
.O EE30233 80.0 EE30956 75.8 EE30365 74.2
()
Average 75.3 Allowable
-40 74.1 ! 3.7 Acceptable DD50442
-40 15.0 i
DD50905 13.8
)
O DD50486 14.0 DD50977 14.0 j
14.0 DD50831 3
Average 14.2 1
Allowable
-40 13.9 1.0 Acceptable 7
i 5-4 J
h O.
L, Table 5-2
'O'.
CHARPY V-NOTCH IMPACT TEST RESULTS FOR IRRADIATED.RPV MATERIALS Test Fracture Lateral Percent Shear
- g)
Specimen Temperature Energy Expansion (Method 2)
Identification
("F)
(ft-lb)
(mils)
(%)
Base:
EUD
-20 13.5 22 10 ETK 0
27.5 31 10 (9
EPH 10 32.5 39 10 EPC 20 38.5 42 15
' EPA 40 45.0 49 30 ETT 60 55.0 49 40 EUC 80 73.0 64 50 EU1 120 86.5 64 85 g,
EP7 160 112.0 88 80 EP3 200 117.7 78 90 EU6:
300 121.7 93 90 l
ETB 400 125.3 95 100 Weld:
ep J16
-20 6.3 19 0
EYL 20 37.0 37 0
J1B 30 22.0 26 15 EYJ 40 14.0 19 20 J2B 60 34.7 52 70 EYA=
80 47.0 47 60 g
J24 100 65.5 64 90 EYM 120 65.0 69 85 EYU 160 56.0 55 65 J2K-200 70.3 60 90 i
J23 300 85.5 78 80 l
EYY.
400 85.5 64 80 lp HAZ:
J4L
-20 21.0 26 20 J4U 0
17.5 20 10 J4A 20 23.7 29 40 6
J5M 30 32.0 34 40 J5L 40 34.0 38-50 J4Y 80 39.0 47 60 J51 100 45.5 41 65 J3A 120 82.5 72 85 J44 160 69.5 61 85 J4M 200 66.0 71 95 3
J52 300 85.7 68 100 J4E 400 85.3 69 90 5-5 3
'G Table 5-3 09 -
SIGNIFICANT RESULTS OF IRRA') LATED AND UNIRRADIATED CHARPY V-NOTCH DATA O
Upper
- Shelf Index Temperature ('F)
Energy (ft-lb) p, Material E=30 ft-lb E=50 ft-lb MLE=35 mil L/T Unirradiated Plate
-66~
-32
-10 129/84 l (7 Irradiated Plate-8 45' 10 112/73 Difference 74 77 20 17/11 (13%)
Unitradiated Weld
-10 17 10 112/112
- (r,
- Irradiated Weld 45 90 40 85/85 Difference 55 73 30 27/27 (24%)
iO 2
Longitudinal (L) USE is read directly from Figures 5-1 and 5-2.
4 Transverse (T) USE is taken as 65% of the longitudinal USE, accordiag to Reference 6.
L/T USE values are equal for veld q
metal, which has no orientation effect.
5-6 3
O Table 5-4 O
ROCKWELL C HARDNESS TEST RESULTS Speciren Rockwell C Identification Type Hardness (HRC)
O EU6 Base 14.2 ETK 16.5 EPA 16.2 EP7 13.9 EP3 14.6
(,,
EUC 17.4 ETB 15.6 EUD 15.9 EPM 15.8 ETT 14.8 EU1 13.9
(,
EPC 15.2 1,verage = 15.3 J2B Weld 16.3 EYU 17.4 J16 15.4 g
J2K 16.3 EYJ 16.6 EYM 16.2 J23 17.6 J24 17.3 EYA 14.7 g,
EYY 13.3 J1B 16.7 EYL 15.5 Average = 16.1 J4M HAZ 16.6 3
J4A 15.6 J4U 13.8 J3A 15.2 J44 16.4 I
J4Y 15.9 J5L 15.7 j
g J4L 16.7 J52 14.9 i
J5M 16.4 J4E 13.2 J51 15.9 Average = 15.5 3
4 i
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O 6.
TENSILE TESTING G
Eight round bar tensile specimens were recovered from the surveillance
}
capsule. Uniaxial tensile tests were conducted 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).
g, 6.1 PROCEDURE All tests were conducted using a screw-driven Instron test frame p
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 p
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.
Thermocouple were spot-welded to the top, middle, and g
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 thermocouple were within 15*F of each other, the temperature of the clip-on thermocouple was noted and subsequently used as g,
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 j
to 0.05 inch / min until fracture. A 1 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 specimens each of base metal and HAZ were tested at room temperature (RT = 76* F), onset of upper shelf temperature (estimated at 185'F), and RPV operating temperature (550*F).
6-1
)
1 b
j J
J
- G' t
The ~ two weld metal specimens were tested at room temperature and 550*F.
O The yield strength (YS) and ultimate tensile strength.(UTS) were calculated by dividing the nominal area (0.0491 in.2) into the 0.2% offset load and into the maximum test load, respectively.
The values listed for the uniform and total elongations were obtained from plots that recorded load O
versus specimen extension and are based on a 1 inch gage length. Reduction of area (RA) values were determined from post-test measurements of the necked specimen diameters using a calibrated blade, micrometer and employing the formula:
O RA = 100% * (A - A )/A f
9 Af ter testing, each broken specimen was photographed end-on showing the D
fracture surface and lengthwise showing fracture location and local necking behavior..
6.2 RESULTS Tensile test properties of YS, UTS, RA, uniform elongation (UE) and total elongation (TE) are presented in Table 6-1.
Shown in Figure 6-1 is a stress-strain curve for a 550*F base metal specimen typical of the (b
strens-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
'6-6 for base, weld and HAZ specimens respectively. The base, weld, and HAZ 6
materials generally follow the trend of decreasing properties with increasing temperature.
)
C9 I
3 6-2
____.-.....__.a
O 6.3 IRRADIATED VERSUS UNIRRADIATED TENSILE PROPERTIES O
Unirradiated tensile test data were recovered from QA records for the surveillance specimen plate (G-2802-2). Data for 0.505 inch diameter gage tensile specimens from the fabrication test program were used to get
(*
unirradiated room temperature YS, UTS, RA and TE properties.
These are coepared in Table 6-2 to the irradiated base metal specimen RT data to determine the degree of irradiation effect.
The trends of increasing YS r
and UTS and of decreasing TE are characteristic of irradiation G
embrittlement. The RA is essentially unchanged.
C' l
0 C
O 6-3
GL Table 6-1 g<
TENSILE TEST RESULTS FOR IRRADIATED RPV MATERIALS Test Yield Ultimate Uniform Total Reduction I'
Specimen Temp Strength -Strength Elongation Elongation of Area Number Material
(*F)-
(ksi)
(ksi)
(%)
(%)
-(%)
J6L Base 76 77.8 100.1 9.6 19.3 68.2 l()
J67 Base 185 74.0 94.7 9.1 18.3
-65.0 J64 Base' 550 68.3 94.0 9.7 18.2 57.4 1
"J74 Weld 76 80.7 98.2 10.3 19.4 53.2 l@
J7D Weld 550 73.7 94.5 8.7 15.9 50.5 JAP.
HAZ 76 80.9 98.9 8.5 17.2 63.5 I
JA3' HAZ 185 76.4 91'.9 8.2 18.0 64.0
!d>
JA6 HAZ 550 71.3 93.6 8.1 15.3 55.4 G
.:t 6-4
)
l O
Table 6-2 O'
COMPARISON OF UNIRRADIATED AND IRRADIATED TENSILE PROPERTIES AT ROOM TEMPERATURE O
Yield Ultimate Total Reduction Strength Strength Elongation of Area (ksi)
(kst)
(%)
(%)
(;,
Base (G-2802-2):
Unirradiated
- 71.4 91.9 28.3 67.4 b
t,..
Irradiated 77.8 100.1 19.3 68.2 Difference "
8.2%
8.2%
-46.6%
1.2%
l l
L U
- G fj
" Specimens have 0.505 inch gage diameter.
b Specimens have 0.250 inch gage diameter.
" Difference = [(Irradiated - Unirradiated)/ Irradiated]
- 100%
')
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J64 5500F Figure 6-4.
Fracture Location, Necking Behavior, and Fracture Appearance for Irradiated Base Metal Tensile Specimens v' -
6-9 l
1
l-1 t 0; i
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Figure 6-5.
Fracture Location, Necking Behavior,.and Fracture Appearance for Irradiated Weld Metal Tensile Specirnens I
l 6-10
C):
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JA6 5500F Figure 6-6.
Fracture Location, Necking Behavior, arid Fracture Appearance for Irradiated IIAZ Metal Tensile Speciraens 3.
6-11
.__._.__..________.____________.___________.______-_______________________a
f
'br 7.
DEVELOPMENT OF OPERATING LIMITS CURVES
]
\\
7.1 BACKGROUND
Operating limits for pressure and temperature are required for three O
categories of. operation:
(a) hydrostatic pressure tests and leak tests, 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 W.
operating limits:
the closure flange region, the core beltline region, and the remainder of the vessel, or non-beltline regions.
The closure flange region limits are controlling at lower pressures primarily because 1
of Reference 1
requirements.
The non-beltline and beltline region q
I operating limits are evaluated according to procedures in References 1 and 2, with the beltline region minimum temperature limits increasing as the vessel is irradiated.
O 7.2 NON-BELTLINE REGIONS Non-beltline regions are those locations that receive too little
)
Non-beltline components include the increase.
fluence to cause any RTNDT M
nozzles, the closure flanges, some shell plates, top and bottom head l
plates and the control rod drive (CRD) penetrations.
Detailed stress The analyses of the non-beltline components were performed for the BWR/6.
all mechanical loadings and thermal transients analyses took into account D
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 Cooper vessel components, since the Cooper geometries are not significantly lO different from BWR/6 configurations and the mechanical and thermal losdings are comparable.
e 7-1 u
I
"/
G' I
F The non-beltline. region results were established by adding the 0
highest RT r the n n-beltline l discontinuities to the' P.versus (T -
ET RTET) curves.for the most-limiting BWR/6 components, which are the CRD
)
penetration and feedwater nozzle.
Shown in Figures 7-1.through 7-3 ar'e.the-
~
Cooper unique calculated non-beltline operating limits for Curves A, B, and.
' D' C, respectively.
The BWR/6 component curves are adjusted to reflect the limiting. non-beltline RT f the Cooper. vessel, which is 28'F for the NDT bottom head torus plates.
O 7.3-CORE BELTLINE REGION The pressure-temperature (P-T) limits for the unirradiated beltline region. are shown in Figures 7-1 through 7-3.
As the beltline fluence 6.
Increases during operation, these curves shift to the right by an amount discussed in Subsection 7.6.
Eventually, the beltline curves shift to become more limiting than the non-beltline curves.
The stress intensity factors calculated for the beltline regien according to Reference 2 a combination of pressure and thermal stresses.
O procedures are. based on The pressure stresses were ulculated 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.
G-The initial RT f 14*F for the limiting beltline plate was used to ET adjust the (T - RTET) values from Figure G-2210-1 of Ref erence 2.
7.4 CLOSURE FLANGE REGION O~
i 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 3
exceed these requirements and they do not affect the shape of the P-T curves.
However, some closure flange requirements from Reference 1 do l impact the curves.
In addition, General Electric recommends 60*F margin on the required bolt preload temperature.
1 7-2 L.. ~ - -... _
g, i
i As ' stated-in Paragraph G-2222(c) of Reference 2, for application of
.g.
full bolt preload and reactor pressure up to 20% of hydrostatic test pressure. - the RPV metal temperature must be at RT r greater.
. practice is to require (RTET + 60*F). for bolt preload, for two reasons:
' p,,
a.
The original ASME Code of construction requires (RTET + 60*F);
and b.
The highest stressed region during boltup is the closure flange O
region, and the flaw size assumed in that region (0.24 inches) is less than 1/4 T.
This flaw size is detectable using ultrasonic testing (UT) techniques.
In fact, References 14 and 15 report that a flaw in the closure flange region of 0.09 inch p'
can be reliably detected using UT.
is not a current ASME Code requirement; it provides extra (RTET + 60*F) margin for'~ Curves A and B.
However, (RTET + 60*F) is a requirement for Curve C, as described in paragraph IV.A.3 of Reference 1.
(3 Reference 1, paragraph IV. A.2, sets temperature minimum requirements for pressure above 20% hydrotest pressure.
Curve A temperature must be g,
no less than (RT
+ 90*F) and Curve B temperature no less than ET (RTET + 120*F).
The Curve A requirement causes a 30*F shift at 20%
hydrotest pressure (312 psig) as shown in Figure 7-1.
The Curve B requirement has no impact on Figure 7-2 because the analytical results for 3
the feedwater nozzle require that temperature be greater than (RTNDT + 120*F) at 312 psig.
7.5 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G b'
Curve C, the core operation curve shown in Figure 7-3 is generated from Figure 7-2, accounting for the requirements of Reference 1,
paragraph IV.A.3.
Essentially paragraph IV.A.3 requires that core critical P-T limits be 40'F above any Curve A or B limits. Curve B is acre a
limiting than Curve A, so Curve C is Curve B plus 40*F.
The (RTNDT +
7-3 5
f l
l minimum permissible temperature mentioned in Subsection 7.4.for
'O Curve C is. an exception for BWRs, allowing critical operation at temperatures below the hydrostatic test temperature.
7.6 EVALUATION OF RADIATION EFFECTS O
The shift in fracture toughness properties in the beltline materials is a function of neutron fluence and the presence of certain elements, such as copper (Cu) and phosphorus (P).
The specific relationship from O
Reference 5 is:
SHIFT ('F) = [40 + 1000(%Cu-0.08) + 5000(%P-0.008)]*(f/10 ')
(7-1)
I lO where:
%Cu = wt % of Cu present,
%P = wt % of P present, f = fluence (n/cm ) at selected'EFPY.
O 2
The limiting beltline plate and weld are determined based on the Cu-P content and initial RT f the materials.
Calculations based on the ET
!O information in Tables 3-1 and 3-2 show the following:
1 14 F Limiting plate:
0.20% Cu, 0.010% P, Initial RT
=
ET Limiting weld:
0.35% Cu, 0.012% P, Initial RTNDT "
lO Surveillance plate: 0.21% Cu, 0.010% P Surveillance veld: 0.22% Cu, 0.012% P Note that the maximum values specified in Reference 5 are chosen for the O-weld, where actual chemistry data for the seam welds was not available.
This material information is used to evaluate irradiation shift versus flizence.
g 7-4 9
' C9 1
7.6.1 Measured Versus Predicted Surveillance Shift
)
O Table 5-3 presents a measured shift for the base metal of 74'F.
The measured shift for the veld metal is 55'F.
The predicted shif ts of the surveillance plate and weld, calculated according to Equation 7-1, assume l'
an upper bound fluence of:
17 1
2 2
f - (2.3x10 n/cm for capsule)(1.25 uncertainty) = 2.9x10 n/cm,
O The predicted shifts are 31*F for the plate and 34*F for the veld.
7.6.2 Modification of the Shift Relationship I
Since the measured shifts exceed the predicted shifts, Jteference 5 methods for predicting the limiting beltline shift may be non-conservative.
The shift relationship is therefore adjusted to reflect the surveillance test results.
As seen in Equation 7-1, the shift is proportional to the O
material characteristics and to the square root of the fluence.
Assuming that the fluence relaticinship is correct, the coefficient representing the materials in Equation 7-1 is increased for the plate materials by the factor (74/31) or 2.39.
Likewise, the weld shift is calculated including a N
factor of (55/34) or 1.62.
7.6.3 Radiation Shift Versus EFPY b
Equation 7-1 can be simplified and expressed as a function of EFPY for the base metal and weld metal.
Subsection 4.3 concludes that the EOL 18 2
(32 EFPY) 1/4 T fluence is 1.5x10 n/cm.
Therefore, in terms of EFPY, the fluer.ce is
?)
f = 4.69x10 6
- EFPY (7-2)
Equation 7-2 is used in Equation 7-1 with the appropriate Cu, P values and factors. For the limiting weld metal: (including factor of 1.62)
= 36.61 * (EFPY)4
~
(7-3)
SHIFT 7-5
O For the limiting base metal:
(including 2.39 factor)
O
=27.82*(EFPY)f.
(7-4)
SHIFTb The adjusted reference temperature (ART) is defined as the initial RT plus the irradiation shift. The initial RT values of the ET NDT limiting plate and weld materials are 14*F and
-50*F, respectively.
Figure 7-4 shows the ART for each material based on these initial RTGT values and the shif ts of Equationa 7-3 and 7-4.
As shown la the figure, The ART values i
O the plate is limiting, because of its higher initial RTET.
of Figure 7-4 are used to develop operating limits curves.
7.6.4 End-Of-Life Conditions Paragraph IV.B of Reference I 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
'D values at 32 EFPY of 157'F for the veld and 171'F for the. plate are acceptable.
Calculations of decrease in USE, using Reference 5,.are summarized in C.
Table 7-1.
The equivalent transverse USE of the plate materials.is taken as 65% of the longitudinal USE, according to Reference 6.
The weld metal USE is not ad' justed because weld metal has no orientation effect.
Surveillance results for the base and weld metal in Table 5-3 show USE O
decreases of 13% and 24%, respectively, versus calculated values of 13% and 16% for the same plate and weld.
Reference 5 provides a good estimate for the plate, but underpredicts the weld decrease in USE by a factor of 1.5.
. O!
Original fabrication testing done by Combustion Engineering developed Charpy data for the beltline plates up to 160*F.
The values at 160*F are averaged to estimate USE for the unirradiated beltline plates.
There are no USE data for the beltline seam welds, aside from the surveillance weld 1
data.
It is
The weld USE decrease in Table 7-1 includes a factor of 1.5 to reflect the measured surveillance weld results.
2 7-6
O The minimum E0L plate and weld USE values are estimated as 58 f t-Ib p
and 73 ft-lb, respectively, which are above the minimum limit.
Therefore, irradiation effects are not severe enough to necessitate RPV annealing before 32 EFPY.
7.7 OPERATING LIMITS CURVES VALID TO 12 EFPY g
The ART selected for the core beltline curves depcnds on the amount of operation for which the curves will be valid.
Twelve EFPY was selected in an effort to keep the leak test required temperature below 200'F.
The g
beltline ART for 12 EFPY estimated with Figure 7-4 is 110*F, based on the plate material.
Adjusting the unirradiated beltline curves in Figures 7-1 through 7-3, with their initial RT f 14 F, and considering the NDT g
non-beltline curves, gives the operating limits valid to 12 EFPY, as shown on Curves A, B and C in Figure 7-5, 7-6 and 7-7, respectively.
The values plotted on Figures 7-5 through 7-7 are tabulated in Table 7-2 fot Curve A and Table 7-3 for Curves B and C.
Curve A was expanded to provide more flexible operating limits during pressure tests.
The non-beltline curve for the bottom head region was included so that the bottom head and beltline regions can be monitored sept.rately.
This reduces the problem during pressure testing of lower g
bottom head temperatures due to cold CRD injection.
The bottom head curve does not shift with time, as it is outside the beltline.
4 The beltline portion of Curve A was expanded to allow more flexibility y
in selecting the beltline target test temperature over the next few years.
Beltline curves for 8 and 10 EFPY were added to the 12 EFPY curve.
Upon evaluating the status of the vessel in EFPY, operators can select the appropriate curve and determine the minimum required test temperature.
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 most severe unplanned transient is an upset condition
~
.9 1
\\
\\
' consisting ' of several' transients which result in a SCRAM.
The worst O
combination of pressure and temperature.is 1180 psig with 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 be applied.
In this case, the core'is (e
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 Curvo A in Figure 7-5, at 1180 psi the I
minimum transient temperature in - the vessel of 250*F lies in the safe
- Q' operating zone.
Therefore, violation of the operating limits curves is I
only -a concern in cases where operator interaction occurs, such as hydrostatic pressure testing and initiation of criticality.
x y
lG l
l
!b 3
ih i
l l ']
9 7-8 p-
, O
.e Table 7 -
l
.Q[
ESTIMATE OF UPPER SHELF ENERGY FOR BELTLINE MATERIALS Upper Shelf (ft-lb) iO
Longitudinal / Transverse IO Identification Uu
'f=0 f=1.5x10 Lower Shell:
8 0
'G-2803-1 0.20 112 92/59 G-2803 0.21 115" 92/60 a
G-2803-3' O.20 til 90/58
.O; Low-Int Shell:
a G-2801-7 0.13 129 111/72 G-2802-1 0.17 111" 92/60 a
G-2802-2 0.21 129 103/67 C8 Surveillance Weld:
a b
1-233 0.22 112 73/73
'C:
O i
f
'i
- USE values taken from test data. Other USE I
values are calculated using Reference 5.
l b
Decrease in USE includes factor of 1.5.
i 7-9 y-l
-n
- x.
~
U;lp
~
Table 7-2 Q:
PRESSURE-TEMPERATURE VALUES FOR FIGURE 7-5 (CURVE A)-
Bottom BELTLINE CURVE TEMPERATURES Head Pressure 12 EFPY 10 EFPY 8 EFPY.
Region
!O-(esi) ~
(der F)
(der F)
(der F)
(der F) 0 80 80 80 80 312 80 80 80 80 312 110 110 110 110 550 110 110 110 110 O!
560 113 110 110 110 570 117 110 110 110 580 120 112 110 110 590 123 115 110 110 600 126 118 110 110 610 129.
121 112 110 0
620 132 124 115 110 630
'134.5 126.5 117.5 110 640 137 129 120 110 650 139.5 131.5 122.5 110 660 142 134 125 110 670 144~.5 136.5 127.5 110 O'
6.80
.146.5 138.5 129.5 110 690 149 141 132 110 700
.151 143 134 112 710 153 145 136 114 720 155 147 138 115.5 730 157 149 140 117 6
740 159 151 142 118.5 750 160.5 152.5
.143.5 120 760 162.5 154.5 145.5 121.5 770 164 156 147 123 780 166 158 149 124.5 790 167.5 159.5 150.5 126 O
800 169 161 152 127 820 172 164 155 130 840 175 167 158 132 860 178 170 161 134 880 181 173 164 136 900 183.5 175.5 166.5 138 920 186 178 169 140.5 3
940 188.5 180.5 171.5 143 960 191 183 174 145 980 193 185 176 147 1000 195.5 187.5 178.5 149 1020 197.5 189.5 180.5 151 1040 199.5 191.5 182.5 153 3
1060 204.5 196.5 187.5 154.5 1080 206 198 189 156 lA 7 10 I.
6l f
Table 7-2 (continued)
'G Bottom BELTLINE CURVE TEMPERATURES Head Pressure 12 EFPY 10 EFPY 8 EFPY Region (usi)
(der F)
(dee F)
(der F)
(deg F) 1100 208 200 191 158 1120 210 202 193=
159.5
.O 1140 212 204 195 161 1160 213.5 205.5 196.5 163
-1180 215 207 198 164.5 1200 217 209 200 166 1220 218.5 210.5 201.5 167.5 1240 220 212 203 169 1260 221.5 213.5 204.5 170.5 1280 223 215 172' 1300 224.5 216.5 173 1320 226 218 174.5 1340 227.5 219.5 210.5 176 0-1360 229 221 212 177 1380 230 222 213 178 1400 231.5 223.5 214.5 179 16 l
2
- y 0
g.
~
7-11
($
Table 7-3 E'
PRESSURE-TEMPERATURE VALUES FOR FIGURE 7-6 (CURVE B) AND 7-7 (CURVE C)
Pressure Curve B Curve C (psi)
Temp. (*F)
Temp. (*F)
Remarks (4
0 60 89 Boltup Temperature 50 80 Non-beltline limits Curve C 60 92 70 100 80 108 90 115 (S
97 80 120 Non-beltline limits Curve B 100 82 122 110 87 127 120 92 132 130 97 137 140 102 142 E
150 106 146 160 110 150 170 113.5 153.5 180 117 157 190 120 160 200 123 163 CD 210 125.5 165.5 220 128 168 230 130.5 170.5 240 133 173 250 135 175 260 137 177
' US 270 139 179 280 141 181 290 143 183 300 145 185 Increment pressure 20 psi 320 148 188 340 151 191 C'
360 155 195 380 158 198 400 160.5 200.5 420 163 203 440 166 206 460 168 208 b'
480 170 210 500 172.5 212.5 520 174.5 214.5 530 175.5 215.5 Beltline becomes limiting 540 177.5 217.5 560 181 221 7
580 184.5 224.5 7-12
,1 I
.__________.__________________.__J
(V
)
Table 7-3 (continued)
(D Pressure Curve B Curve C (psi)
Temp. (*F)
Temp. (*F)
Remarks 600 188 228 620 191 231 640 194 234 CA 660 197 237 680 200 240 700 202.5 242.5 720 205 245 740 207.5 247.5 760 210 250
(*
780 212.5 252.5 800 214.5 254.5 820 217 257 840 219 259 860 221 261 880 223 263 0
900 225 265 920 226.5 266.5 940 228.5 268.5 960 230.5
~270.5 980 232 272 1000 233.5 273.5 9
1020 235.5 275.5 10/:0 237 277 1060 240.5 280.5 1080 242 282 1100 243.5 283.5 1120 245 285 (p
1140 246.5 286.5 1160 248 288 1180 249.5 289.5 1200 250.5 290.5 Increment pressure 50. psi 1250 254 294 1300 257 297 y
1350 260 300 1400 263 303 7-13
G 1600 O
l 1400 0
UNIS1 RADIATED BELTUNE REGION
[
1200 (RTNDT = 14*F l
l(+
1000 NON-BELTLtNE REGION
~
(RTNDT = 28'F) l l'
~
O h800 I
m G
E l
l
/
600 l
10CFR50, APPENDIX G, IV.A.2 REQUIREMENT (RTNOT = 20*F) g l
l 400 I
l p __.
_...=.
J 3
l 1
GE RECOMMENDED BOLT 200 1
PRELOAD TEMPERATURE OF l
60'F (INCLUDES 60'F MARGIN) i g
1 0
O 100 200 300
)
MINIMUM VESSEL METAL TEMPERATURE ('F)
Figure 7-1.
Components of Operating Limits Curve for i
Pressure Tests (Curve A) for Cooper
]
l 7-14 I
J
fQ 1600 O
[
1400 UNIRRADIATED BELTUNE REGION I
(RTNDT = 14*F) l 1200 c
1 1000
_1 I
Qr I
800 g
l I
M O
k 600 NON-BELTLINE REGION (RTNDT = 2 8'F)
I I
10CFR50, APPENDIX G,IV.A.2 l
~
7 _ _ _ _ _ _'F)y REQUIREMENT (RTNDT = 20 i
200
/
l GE RECOMMENDED MINIMUM TEMPERATURE A
g OF 80*F (INCLUDES 60'F MARGIN) f I
i t
0 0
100 200 300 3
MINIMUM VESSEL METAL TEMPERATURE (*F)
Figure 7-2.
Components of Operating Limits Curve for Non-Nuclear Heatup/Cooldown (Curve B) for Cooper 7-15 9
- (3 I
- 1630, O
t 1400 UNIRRADIATED BELTLINE REGION (RTNDT = 14'F) g I
1200 I
.o 1000 J
k 800 f
fI
.O
-E E
m 600 10CFR53, APPENDIX C, IV.A.2 AND IV.A.3 REQUIREMENTS (RTNDT = 20*F) g l
400 J/
a r----
1 200
/
NON BELTLINE REGION (RTNDT = 2 8'F)
O g
9 0
O 100 200 300 MINIMUM VESSEL METAL TEMPERATURE ('F)
Figure 7-3.
Components of Operating Limits Curve for Core Critical Operation (Curve C) for Cooper 3
7-16
'O O
2 3
/ *
/.
d l
9 eW
/
8 d
2 n
~
/
a e
t a
l r P e O
p
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eo 4
no 2
i C l
/
t r
)
l o P
ef y
B 2
s 1
gt 0
nl 9
/
0 S
i u E
0 t s U
L 2
S i e
'u
/
A R
mR C
V A
i E
E Lt 5
S Y
s 3
A R
f e C
E oT 0
W T
(
S O
en
- D L
R P
re E
O 6
L um Q
W W
L ti 1
U F
ac
/
re E
/
I ep V
pS T
)
C m
P E
ee F
Tc 0
F n
/
E 1
ea 0
2 cl e
0 1
nl M
/
ei uC re ev f r
/
0 eu 2
RS 0
d n
(
y E
eo T
t A
a 8
sd L
/
P 9
ue js d a
/
/
4
/
t 4
7 7
er v
/
ug i
[/
F l-O
~
0 0
0 0
0 0
0 0
0 0
0 0
0 8
6 2
0 8
6 4
2 2
4 6
w 1
1 1
1 5
FsIm ad $zM s$5 G*-
5hr$ gz=.$ 9$34 u
[w e
o 1600 O
CURVE A BOTTOM EFPY HEAD REGION 8 10 12 1400 I
f.fI
/
/ / /
c,
/
/ / /
1200 I
/ / /
f f
I i
A
/
O BELTLINE CURVES, I I I
1/4 T FLAW,
-[5 ADJUSTED AS 1000
)
/ / /
E
[
J
/
10 1
A I
800
/ [
ATING "cac" g
j jy/
i M
O c110'F 400 312 psig g h
O 200 BOLT PRELOAD TEMPERATURE = BO'F
' FLANGE REGION
/ RTNOT = 20'F j
0 0
100 200 300 MINIMUM VESSEL METAL TEMPERATURE ('F)
'a Figure 7-5.
Pressure Versus Minimum Temperature for Hydrostatic Pressure Tests for Cooper 7-18
O 1600 l
l VALID TO 12 EFPY O
B
')
1400 ADJUSTED BELTINE.
/4T FLAW, ART = 110*F
'O 1200 r
O 1000 Y
O E
Ea 800 0
E w
hw
/
600 D
SAFE i
OPERATING REGION 400 31 NON BELTLINE FW N0ZZLE LIMITS, 1/
FLAW. RTNOT= 28 'F 200
/
BOLT PRELOAD TEMPERATURE = 80'F
...J-FLANGE REGION RTNDT = 20*F 1
0 0
100 200 300 MINIMUM VESSEL METAL TEMPERATURE ('F) 3 Figure 7-6.
Pressure Versus Minimum Temperature for Non-Nuclear Heatup and Cooldown for Cooper 7-19 J
- C#
- 1600 g
VALID TO 12 EFPY O'
C 1400 ADJUSTED BELTINE,
- D 1/4T FLAW, ART = 110'F 1200
- O 1
1000 1
- 6 C
E 800 i
E I.
e 600 NON BELTLINE SAFE FW NOZ2LE LIMITS OPERATING PLUS 40'F, il4T FLAW REGION RTNDT = 28'F 400 1
MINIMUM PERMISSIBLE 3
200
" TEMPERATURE = 80'F PER 10CFR50 APPENDIX G O
o#s 0
100 200 300 MINIMUM VESSEL METAL TEMPERATURE l'F)
)
l 3
Figure 7-7.
Pressure Versus Minimum Temperature for Core Critical Operation for Cooper J
7-20
-O 8.
REFERENCES LO 1.
" Fracture Toughness Requirements," Appendix G to Part 50.of Title 10 of the Code of Federal Regulations, July 1983 (24FR24008).-
2.
" Protection Against Non-Ductile Failure," Appendix G to Section III
' (j.
of the ASME Boiler & Pressure Vessel Code, Addenda to and including Summer 1986.
tO 3.
" Reactor Vessel thterial Surveillance Program Requirements,"
Appendix H to Part 50 of Title 10 of the Code of Federal Regulations, July 1983 (48FR24008).
l(n 4.
" Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," Annual Book of ASTM Standards, E185-82, July 1982.
5.
" Effects of Residual Elements on Predicted Radiation Damage to O
Reactor Vessel Materials " USNRC Regulatory Guide 1.99, Revision 1,
April 1977..
6.
" Fracture Toughness Requirements," USNRC Branch Technical Position G
MTEB 5-2, Revision 1, July 1981.
7.
" Capsule Basket," General Electric Drawing 117C4190, Revision 1,
October 1969.
8.
Letter, R. A. Hillis of Combustion Engineering, Inc. to G. A. Barry of General Electric Company, " Welding Material Certification Data for Cooper Beltline," June 1986.
..y-9.
" Surveillance Test Program for Cooper Station Reactor Vessel,"
Combustion Engineering Specification (CE VPF 1837-231-2), liay 1969.
,?
8-1 2
P O
- 10. " Standard Methods for Notched Bar Impact Testing of Metallic O
Materials," Annual Book of ASTM Standards, E23-82, March 1982.
11.
" Standard Test Methods for Rockwell Hardness and Superficial Rockwell Hardness of Metallic Materials," Annual Book of ASTM Standards, l
'D E18-79.
- 12. Wang, M.
T.,
" Fracture Toughness of Reactor Pressure Vessel Steel Welds," General Electric Company (NEDC-30299), October 1983.
0 13.
" Standard Methods of Tension Testing of Metallic Materials," Annual Book of ASTM Standards, E8-81.
O 14.
" Ultrasonic Examination for Cracks in the Top Head Flange," CBI 1
Nuclear, Development Report 74-9047, December 1975.
- 15. " Ultrasonic Examination for Cracks in the Shell Flange," CBI Nuclear, Development Report 74-9056, November 1975.
i 1
t, i
- a 1
1 i
i
,)
J l
a l
1 8-2 2
.{
6P APPENDIX A
- O-CHARPY V-NOTCH TRACTURE SURFACE PHOTOGRAPHS Photographs of-each Charpy specimen fracture surface were taken O
to facilitate the determination of percent shear, and to comply with the requirements of ASTM E185-82.
The pages following show the fractbre surface photographs along with a summary of the Charpy test results for each specimen.
The pictures are arranged by increasing O
test temperature for each material, with the. materials in the order of base, weld and HAZ.
- O
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