ML20148C531

Analysis of Capsule V from Rochester Gas & Electric Re Ginna Unit 1 Reactor Vessel Surveillance Program

ML20148C531
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Site:	Ginna
Issue date:	04/01/1973
From:	Mager T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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R. E. GINNA UNIT NO. 1

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REACTOR VESSEL RADIATION L-SURVEILLANCE PROGRAM BY:

T. R. Mager S. E. Yanichko S. L. Anderson S. A. Legge D. J. Lege March 1, 1973 dm APPROVED:

b. N. Chirigos \\

Westinghouse Electric Corporation Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 2835

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TABLE OF CONTENTS Section Title Page 1.0

SUMMARY

OF RESULTS.......................................... 1 2.0 I NT RO DU CT I O N................................................ 2

3.0 BACKGROUND

.................................................. 3 4.0 D E S C R ' PT ION O F PR0G RAM....................................... 5 5.0 TESTING OF SPECIMENS FROM CAPSULE "V"........................ 7 5.1 Cha rpy V-Not ch Impac t Test Result s..................... 12 5.2 Tensile Te s t Re sul t s................................... 2 4 5.3 WOL Test Results.......................................

24 6.0 DOSIMETRY ANALYSIS..........................................

34 6.1 Measured Activity of Dosimeters........................

34 6.2 Analytical Methods.....................................

36 6.2.1 Activation Monitors.............................

36 6.2.2 Fission Monitors................................

38 6.2.3 Projections of Sample Fluence to the Reactor Vessel..................................

39 6.3 Results of Analysis....................................

41 RE F E RE N C E S............................................................ 4 3 APPEND 1X...

44 1.0 HEATUP AND C00LDOWN LIMIT CURVES IN ACCORDANCE WITH ASME BOILER AND PRESSURE VESSEL CODE, SECTION III.......... 44 l

2.0 FRACT URE T OUGHNE S S PROPERTIE S..............................., 4 4 2.1 Criteria For Allowable Pressure-Temperature Relationships......................................... 48 2.2 Hea tup and Cooldown Limit Curve s....................... 50 3.0 SURVEILLANCE CAPSULE REMOVAL SCREDULE...................... 56 l

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SUMMARY

OF RESULTS The analysis of the first reactor vessel surveillance capsule material removed from the R. E. Ginna Unit No.1 reactor pressure vessel led to the following conclusions:

1 1.

The capsule received an average fast fluence of 4.9 x 10 neutrons /cm (E > 1 MeV). The predicted fast fluence for the capsule at the end 18 of the 1st core cycle was 6.0 x 10 neutron /cm (nyt > 1 MeV).

18 2.

The fast fluence of 4.9 x 10 n/cm (E > 1 MeV) increased the 30 ft-lb "fix" nil ductility transition temperature (NDTT) of the veld metal 140*F.

The lower (125P666) and intermediate (125S255) pressure vessel shell forging materials exhibited essentially a 25'F and a O'F shift respective-ly in the 30 ft-lbs "fix" nil ductility transition temperature.

3.

The lower (125P656) and the intermediate (125S255) pressure vessel shell forgings and the weld metal exhibited a fracture toughness in excess of 100,000 psi in ! as measured using equivalent energy concepts.

11 2

4.

The fast flux of 1.0 x 10 neutrons /cm /sec (E > 1 MeV) received by 11 Capsule V was approximately 83.3% of the 1.2 x 10 neutrons /cm see (E > 1 MeV) predicted for the R. E. Ginna Unit No. 1 reactor pressure vessel capsule.

I 5.

Based on a ratio of 3.3 between the fast flux at the surveillance capsule location to that at the vessel wall and a 80 percent load factor, the projected f ast fluence which the R. E. Ginna Unit i reactor pressure vessel will receive after 40 calendar year operation is 3.7 x 10 ' n/cm2 1

(E > 1 MeV). This is the same fluence as assumed in the FSAR for 40 year operation.

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6.

The increase in the 30 f t-lb "fix" nil ductility transition temperature of 140'F for the weld metal was approximately 25 F greater than predict-ed by the Technical Specification 550 F trend curve. The projected shift in the 30 f t-lb "fix" transition temperature of the weld metal af ter 40 calendar year operation is 300 F.

7.

The upper shelf impact energy of the weld metal decreased from 74 f t-lbs to 50.8 ft-lbs during the 1st core cycle.

8.

The irradiated properties of forgings 125P666 and 125S255 and the weld metal are adequate to provide for continued safe operation of the Robert E. Ginna Nuclear Power Plant Unit 1.

A summary of the neutron fluence and 30 f t-lb "fix" nil ductility trane-ition temperature data obtained from specimens in Capsule V is as follows:

18 Neutron fluence average E > MeV 4.9 x 10 n/ca' 11 Flux (E > 1 MeV) 1.0 x 10 n/cm see Shif t in 30 f t-lb "fix" NDTT Forging 125P666 s 25 F Forging 125S255 s 0F Weld Metal 140 F Correlation Monitor Material N 90 F

2.0 INTRODUCTION

This report presents the results of the examination of Capsule V, the first capsule of the continuing surveillance program for monitoring the effects of l

neutron irradiation on the Rochester Gas and Electric R. E. Ginna Unit No. I reactor pressure vessel materials under actual operating conditions.

The surveillance program for the R.E. Ginna Unit No.1 reactor pressure

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vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented in Reference [1].

The surveillance program was planned to cover the 40-year life of the reactor pressure vessel and was based on ASTM E-185-66 (Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors).

Westinghouse Nuclear Energy Systems personnel were contracted for the prepar-ation of procedures for the removal of the capsule from the RGE reactor and its shipment to the Southwest Research Institute (SwRI) Laboratories in San Antonio, Texas. SwRI Laboratories' personnel, under the technical direction of E. B. Norris, performed the post-irradiation mechanical testing of the sur-veillance specimens.

The dosimeters used to measure the integrated flux were shipped to WpES for evaluation.

This report summarizes the post-irradiation data obtained from the first material surveillance capsule (Capsule V) removed from the R. E. Ginna Reactor Vessel and discusses the analysis of these data. Using current methods as well as those l

specified in the Technical Specifications heatup and cooldown pressure-tempera-ture operating limits were established for the R. E. Ginna Unit No.1 nuclear power plant. The heatup and cooldown pressure temperature operating limits are presented in the Appendix to the report.

3.0 BACKGROUND

The ability of the large steel pressure vessel that contains the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to neutron bombardment. The overall effects of fast neutron-irradiation on the 1

l mechanical properties of low alloy ferritic pressure vessel steels such as ASTM A508 Class 2 (parent material of the R. E. Ginna reactor pressure vessel) are well l

documented in the literature. Generally, low alloy ferritic materials show an l

increase in hardness and other strength properties and a decrease in ductility under certain conditions of irradiation.

In pressure vessel material, the most serious mechanical property change is the reduction in the upper shelf impact strength. Accompanying the decrease in impact strength is an increase in the l

I l

l temperature for the transition f rom brittle to ductile fracture. Any radiation embrittlement or changes in the mechanical properties of a given pressure vessel steel can be monitored by reacter surveillance programs such as the Rochester Gas And Electric R. E. Ginna Unit No. 1 reactor-vessel irradiation surveil-lance program where a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested.

Two general approaches have evolved for guarding against fast fracture in veld-ed structures.

The first of these is known as the transition temperature approach, while the second approach is the stress analysis or energy release rate approach.

The transition temperature approach was the earlier because at the time it was felt to be impossible to evaluate fast failure in terms of the principle stresses in the structure. The basic philosophy of the trr on temperature approach is relatively simple - a material has a characteristic temperature at which it is susceptible to low-stress, fast fracture in the presence of sharp defects or cracks, and above which low-stress fast fracture does not occur. Of the transition temperature tests, the Charpy V-notch impact test is one of the most commonly used.

Data from one set of Charpy V-notch specimens =ay be used to define several types of transition temperatures.

Both the tough frangible and fracture-mode types of transition temperatures are represented. Transition temperature may be defined, for example, as that temperature at which the energy required to fracture a specimen is chosen as some arbirarily low value, such as 30 ft-lbs.

Or it may be defined as some higher temperature where the sample reveals a certain arbitrar-ily defined percentage of brittle fracture (usually 50 percent brittle-50 percent shear). More recently, the use of lateral contraction at the notch root (lateral expansion measurements) has become a potential method for defining the transition te=perature.

The transition temperature philosophy was reflected in both the make-up of the R. E. Ginna surveillance capsule program and the ASTM E-185 recommended practice for surveillance progra=s.

That is, impact and tensile specimens were encapsulated in the R. E. Ginna surveillance capsules.

A second widely used method of defining a transition temperature is the NDTT (Nil-Ductility Transition Temperature) test.

This test technique utilizes a plate specimen, supported near the ends, and subjected to a load applied through a dropped weight. A crack starter is employed in the form of a weld bead on the tension side of the plate.

A stop under the specimen limits bending and assures'a brittle fracture. The highest temperature at which the plate f ractures is defined as the NDIT.

The test is cocconly called a drop weight test.

Because space limitations in most pressurized water reactor surveillance programs are restricted to approximately one (1) inch, drop weight specimens are not encap-sulated in surveillance capsules.

However, NDTT is generally obtained on the preirradiated material utilizing the drop weight specimen. It is generally agreed, that for mild steels such as A508 class 2, NDTT as determined f rom a drop weight test corresponds to the 30 f t-lb energy as determined from a Charpy V-notch impact test.

Although the transition temperature approach is a conservative approach when applied to reactor pressure vessels, it does not provide the basis for quantitative consideration of the many aspects involved in a thorough evaluation of the brittle failure potentials. During the past decade, a much more sophisticated and quantitative approach to the brittle fracture problem has been developed.

Research during the last ten years has been devoted to linear-elastic fracture mechanics (LEFM), the "forerunner" of modern f racture mechanics concepts. LEFM is based on the energy release rate criteria.

The theoretical and experimental aspects of LEFM are well documented and the technology suf ficiently advanced to where it can be successfully applied to considerations of the potential of fast fracture in heavy wall reactor pressure vessels.

Foreseeing the advance of the fracture mechanics technology, Westinghouse recommended that WOL (Wedge Opening Loading) fracture mechanics specimens be included in the R. E. Ginna surveillance program to evaluate radiation damage to pressure vessel =aterials.

4.0 DESCRIPTIOS OF PROGRAM Six surveillance capsules to monitor the effects of neutron exposure on the R. E. Ginna reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup.

The six capsules were pos-itioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in Figure 1.

The vertical center of the capsules is opposite the vertical center of the core..

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Capsule V was removed during the first refueling shutdown. This capsule con-tained Charpy V-notch impact, tensile (Figure 2) and WOL specimens (Figure 2) from the intermediate and lower shell ring forgings (Heats 125P666 and 125S255),

weld metal from the core region of the reactor vessel and Charpy V-notch spec-imens f rom weld heat af f ected zone (HAZ) material. The capsule also contained Charpy V-notch specimens from the 6-inch-thick correlation monitor material (A302 Gr. B) furnished by the U.S. Steel Corporation. The chemistry and heat treatment of the surveillance material is presented in Tables 1 and 2.

All test specimens were machined from the 1/4 thickness location of the forgings.

Test specimens represent material taken at least one forging thickness frem the quenched end of the forging. All Charpy V-notch ano tensile specimens were oriented with the longitudinal axis of the specimen parallel to the hoop direction of the forgings. The WOL test specimens were machined with the simulated crack of the specimen perpendicular to the surfaces and the hoop direction of the forgings.

Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the weld. Tensile specimens were oriented rich the longitudinal axis of the specimen parallel to the weld.

Capsule V contained dosimeter wires of copper, nickel, and aluminum-cobalt (cadmium shielded and unshielded).

In addition cadmium shielded dosimeters of Np and U were contained in the capsule and located as shown in Figure 3.

Thermal monitors made from two low melting eutectic alloys and sealed in pyrex tubes were included in the capsule and were located as shown in Figure 3.

The two eutectic alloys and their melting points are:

2.5% Ag, 97.5% Pb Melting Point 579'F 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 590*F 5.0 TESTING OF SPECIMENS FROM CAPSULE "V" The post-irradiation mechanical testing of the specimens was performed at Southwest Research Laboratories with consultation by Westinghouse NES personnel...

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l TABLE 1 CHEMISTRY AND HEAT TREATMENT OF SURVEILLANCE MATERIALS REPRESENTING TWO FORGING SHELL COURSES AND A WELDMENT FROM THE RGE UNIT NO.1 REACTOR VESSEL I

CHDtICAL ANALYSES (PERCENT)

SHELL FORGING COURSE C

Mn P

S Si Ni Cr Co Mo V

Cu Al N2 Sn 125P666 Lower 0.19 0.67 0.10 0.011 0.20 0.69 0.37 0.013 0.57 0.02 0.05 0.004 0.01 125S255 Intermediate 0.18 0.66 0.010 0.007 0.23 0.69 0.33 0.015 0.58 0.02 0.07 0.003 0.01 Weldment 0.075 1.31 0.012 0.016 0.59 0.56 0.59 0.001 0.36 0.001 0.23 0.020 0.015 i

HEAT TREATMEKr l

b HEATED TO 1550*F - 9 HRS.

WATER QUENCllED TEMPERED AT 1220*F - 12 liRS., AIRC00 LED 66 STRESS RELIEVED AT 1100*F - 11 HRS., FURNACE COOLED i

!!EATED AT 1550*F 1/2 HRS., WATER QUENCHED FORGING TEMPERED AT 1220*F - 18 liRS. AIR COOLED 125S255 STRESS RELIEVED AT 1100*F 1/4 liRS., FURNACE COOLED i

i Weldsent STRESS RELIEVED AT 1100*F 1/4 IIRS., FURNACE COOLED

TABLE 2 CHDilSTRY AND HEAT TREATMENT OF SURVEILLANCE MATERIAL REPRESENTING U.S. STEEL CORPORATION CORRELATION MONITOR MATERIAL CHDi1 CAL ANALYSES (Percent)

C Hn P

S Mo Si

._C u 0.24 1.34 0.011 0.023 0.51 0.23 0.20 HEAT TREATMENT The six-inch-thick plate was charged into a furnace operating at 110*F heated at a maximum rate of 63*F per hour, to 1650*F, held at temperature for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and water quenched to 300*F.

The plate was then recharged into a furnace operating at 700 to 750*F and heated at a maximum of 63*F per hour to 1200*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, t

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Upon receipt of the capsule at the SwRI Laboratories, the specimens and spacer blocks were carefully removed and placed in an indexed receptacle so that specimen location was identifiable.

Each specimen was inspected for identification number, which was checked against the master list in WCAP-7254. No discrep-ancies were found.

The pyrex tubes, housing the thermal monitors, were placed inside plastic bags before they were broken open so that the contents could not be lost.

Examin-ation of the two low melting (579'F and 590'F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579*F.

swr 1 Laboratories used a Riehle impact test machine to perform tests on the irradiated Charpy V-notch specimens.

Prior to initiating tests on the irrad-iated Charpy-V specimens, the accuracy of the Riehle impact machine was checked with a set of standard specimens obtained from the Army Material and Mechanics Research Center in Watertown, Massachusetts.

The results of the calibration testing showed that the machine was certified for Charpy V-notch impact testing.

A remotely operated specimen transfer mechanism permitted breaking the specimen within 4 seconds after removal from the heating bath.

A Dillon 10,000 pound capacity tensile test machine equipped with a strain gage extensometer, load cell, and autographic recording equipment was used to perform the post-irradiation tensile and WOL testing. When the WOL specimens were test-ed, the extensometer was replaced by a clip gage to measure the crack opening i

displacement. An electric laboratory furnace was used to heat specimens above ambient for elevated temperature testing.

The test specimens were instrumented with thermocouples wired to the test section.

5.1 CHARPY V-NOTCH IMPACT TEST RESULTS The irradiated Charpy-V specimens represented reactor pressure vessel beltline forging material, weld and heat-affected-zone (HAZ) material and the ASTM l

correlation monitor material, T't.e results are presented in Tables 3 thru 7 and Figures 4 thru 6.

The unirradiated data are also shown in Figures 4 thru 8 for comparison with the post-irradiation data.

A sum =ary of the increase in the 30 ft-lb "fix" transition temperature and the decrease in the upper shelf energy is presented in Table 8. -

TABLE 3 IRRADIATED CHARPY V-NOTCH IMPACT ENERGY FOR THE R. E. GINNA UNIT NO. 1 PRESSURE VESSEL SHELL FORGING 125P666 FLUENCE 4.9 x 10 j,2 (g,1g,y) 18 SPECIMEN TEST ENERGY LATERAL EXPANSION NUMBER TEMP.('F)

(PT-LBS)

(MILS)

P3

-50 5.8 4

P4

-10 32.5 27 P9

+10 41.3 36 P5

+40 84.5 69 P8

+40 94.3 72 P7

+75 99.5 71 P6

+140 165.0 92 P1

+210 146.0 70 P10

+210 155.5 90 P2

+500 160.0 95 i - - -.

TABLE 4 IRRADIATED CHARPY V-NOTCH IMPACT ENERGY FOR THE R. E. GINNA UNIT NO. 1 PRESSURE VESSEL SHELL FORGING 125S255 18 2

FLUENCE 4.9 x 10 g yg,y)

SPECIMEN TEST ENERGY LATERAL EXPANSION NUMBER TEMP.(*F)

(FI-LBS)

(MILS)

S1

-50 3.8 3

S7

-25 74.5 56 55

-10 51.3 41 S8

+10 66.5 64 S3

+40 80.5 58 S6

+75 107.0 85 S9

+140 129.0 88 S10

+175 140.5 86 S4

+210 136.0 91 S2

+500 125.5 92

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TABLE 5 IRRADIATED CHARPY V-NOTCH IMPACT ENERGY FOR THE R. E. GINNA UNIT NO. 1 PRESSURE VESSEL WELD METAL l0 FLUENCE 4.9 x 10 n/cm (D1 MeV)

SPECIMEN TEST ENERGY LATERAL EXPANSION NUMBER TEMP.(*F)

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W6

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+75 23.0 17 W5

+125 29.5 21 W8

+125 34.5 23 W7

+175 47.0 38 W10

+175 42.3 46 W3

+210 50.5 45 W9

+300 53.3 51 W4

+500 55.0 55 4

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TABLE 6 IRRADIATED CHARPY V-NOTCH IMPACT ENERGY FOR THE R. E. GINNA UNIT NO. 1 PRESSURE VESSEL HEAT-AFFECTED-ZONE METAL 1

FLUENCE 4.9 x 10 n/cm (D 1 MeV)

SPECIMEN TEST ENERGY LATERAL EXPANSION NUMBER TEMP.(*F)

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HS

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H8

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+10 50.5 35 H10

+10 61.0 53 H7

+40 57.5 46 H1

+75 36.0 27 H9

+75 86.3 63 H4

+125 76.5 40 H2

+210 120.0 92 H6

+500 132.0 85.... _ -.. -. -, _

TABLE 7 IRRADIATED CFJJtPY V-NOTCH IMPACT ENERGY FOR THE R. E. GINNA UNIT NO. 1 ASTM SA302B CORRELATION MONITOR MATERIAL (SUPPLIED BY U.S. STEEL)

ELUENCE 4.9 x 1018 @2 (p ggy)

SPECIMEN TEST ENERGY LATERAL EXPANSION NUMBER TEMP. ('F)

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+125 34.0 30 R4

+150 38.0 33 R7

+175 45.0 42 R8

+175 44.5 41 R2

+210 61.5 51 R6

+300 66.0 51

s TABLE 8 18 2

THE EFFECT OF 550'F IRRADIATION AT 4.9x10 n/cm (E>1MeV) ON THE NOTCH TOUGHNESS PROPERTIES OF THE R. E. GINNA UNIT NO.1 REACTOR VESSEL IMPACT TEST SPECIMENS Energy Absorption 30 FT.-LB. TEMP.(*F)

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CHARPY V-Notch Impact Energy versus Temperature for R. E. Ginna Unit 8l Pre.sure Vessel Weld Heat Affecte Zone Material 22

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The test resulte obtained on the vessel beltline material are presented in Tables 3 and 4 and Figures 4 and 5.

The data showed that the two ring forging heats 125P666 and 1255255 are relatively insensitive to irradiation based on the respective 30 f t-lb "fix" transition temperature shif ts of 25'F and 0*F 8

2 at a fluence of 4.9 x 10 n/cm., The relative small decrease in the upper shelf impact energy with the neutron exposure also tends to indicate that the two forging heats are insensitive to irradiation.

The test results obtained on the weld material are presented in Table 5 and Figure 6.

The data showed that the fluence received by Capsule V raised the 30 f t-lb "fix" transition temperature of the weld metal specimens 140*F.

The upper shelf impact energy was decreased approximately 30 percent.

The weldment HAZ material showed essentially no transition temperature shif t or decrease in upper shelf impact energy.

The test results for the HAZ material are summarized in Table 6 and Figure 7.

The test results obtained on the ASTM reference correlation monitor material are presented in Table 7 and Figure 8.

The 90*F increase in the 30 ft-lb "fix" transition temperature is in agreement with the reference trend band established by the Naval Research Laboratories ' for this material.

5.2 TENSILE TEST RESULTS The results of the tensile tests are presented in Table 9.

Tests were perform-ed on specimens from each of the two forgings and the weld metal at room tempera-ture, 212*F and 580*F.

In general, the yield strength of the materials increased approximately 20 percent with neutron bombardment. A typical load-displacement curve obtained for the tensile tests is shown in Figure 9.

5.3 WOL TEST RESULTS Of the various quantitative methods that have evolved for Fuarding against fast fracture within the past few years linear-elastic fracture mechanics (LEFM) is the most universally accepted. With appropriate information in the related areas of material properties, stresses, and potential def ects, the concepts and expres-sions of LEFM can be employed in established step-by-step procedures which.

TABLE 9 IRRADIATED TENSILE PROPERTIES FOR THE R. E. GINNA UNIT NO. 1 PRESSURE VESSEL SH' ELL FORGINGS AND WELP METAL 18

' FLUENCE 4.9 x 10 n/cm (D IMeV) 0.2% YIELD ULTIMATE TENSILE REDUCTION SPECIMEN TEST STRENGTH STRENGTH ELONGATION IN AREA NUMBER FORGING TEMP.('F)

(PSI)

(PSI)

(%)

(%)

P-13 125P666

+80 70,600 95,900 24.2 73.4 P-14 125P666

+212 64,900 88,500 24.6 72.5 P-15 125P666

+580 64,100 89,900 21.1 65.9 S-14 125S255

+80 74,900 96,800 23.7 69.7 S-13 125S255

+212 86,600 107,000 19.6 65.1 S-15 125S255

+580 84,400 106,000 15.7 58.4 W-2 Weldment

+80 95,400 112,000 21.2 59.7 W-1 Weldment +212 88,500 105,000 22.6 63.1 W-3 Weldment +580 82,100 102,000 21.0 56.7. _ _ _ - _

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will ensure immunity from fast fracture. The material parameter KIc, a asic material property, is dependent upon mechanical and metallurgical condition of the material. Thus, neutron bombardment is known to influence the fracture resistance of pressure vessel grade steels.

The fracture toughness, K can be, determined by testing a fracture mechanics k

specimen; WOL, Compact Tension, bend or spin disk. Mager

} has demonstrated the fracture toughness K f A508 Class 2 steel to be highly temperature Ic dependent with a rapid increase in toughness near the NDTT (nil ductility transition temperature).

Valid K data were obtainable only up to about 10*F, 7

even when using 8-inch-thick specimens. Furthermore, it has been demonstrated l

by Mager that, for the irradiation conditions of interest to the nuclear industry, specimens with thicknesses of 4 inches or greater must be tested to attain a fracture toughness K of 100,000 psi (in).

As previously mentioned, because of reactor space limitations, reactor vessel surveillance capsules are restricted to approximately one-inch in thickness.

IX-WOL f racture mechanics specimens were included in the R. E. Ginna reactor materials surveillance program.

For the temperature range of interest, it is impossible to obtain valid K fracture toughness data from reactor vessel sur-7#

veillance programs.

However, Witt{12] described a method to at least obtain quantitative bounding values for the fracture toughness. The basis for the method of data analysis is the equivalent energy concept.

A method is outlined in reference 12 for obtaining KQd (subscript d being i

the test specimen thickness) value from standard fracture toughness tests.

These steps are as follows:

1.

Measure the area under the load-deflection curve up to maximum load of a specimen of thickness d (Point A in Figure 10).

2.

Select any point on the linear portion of the load deflection curve.

(Point B in Figure 10).

Measure the area up to this point and divide this area into the area up to maximum load (Point A in Figure 10).

Call this ratio b.

3.

Using the load at Point B (Figure 10) as P calculate K from the K q

g 7

expression for a compact tension specimen.,

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The value thus determined is unique regardless of the point selected on the linear portion of the curve.

The testing of each of the nine irradiated WOL specimens resulted in the deter-mination of lower bound K data. The WOL test results are summarized in R1 Table 10.

Testing was performed at room temperature and/or 212'F.

The fracture behavior of the WOL specimens in the post-irradiation condition can be generalized by three types of load-displace =ent curves.

Type 1 is shown in Figure 11 where the load-displacement curve shows gross yielding at the crack tip. The specimen did not completely break in two.

This was typical for specimens S-2, P-1, P-2 W-2 and W-3.

Load-displacer.ent behavior type 2 is shown in Figure 12 where there is an intermediate amount of plasticity at the specimen crack tip. This was typical for specimen S-1.

The third type of load-displacement behavior is signif-ied in Figure 13; very little plasticity occurred at the crack tip.

The Charpy V-notch impact test results presented in Tables 3 thru 7 were indic-l ative of the WOL specimens results.

Forging 125P666 was relatively insensitive to the irradiation (%'F increase in transition temperature) and the lower bound fracture toughness K was relatively high (s310 ksi (in)1/2).

The weld g

metal which was sensitive to the irradiation exhibited a decrease in shelf impact energy and the lowest fracture toughness of the three materials. However, the lower (125P666) and intermediate (125S255) pressure vessel shell forgings and weld metal exhibited fracture toughness in excess of 100,000 psi (in)1/~'

as measured using equivalent energy concepts.

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Type 2 Load Displacement Curve Representative of a intermediate Amount c,f Plosticity at tt e Specimen Crack Tip

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TABLE 10 IRRADIATED FRACTURE TOUGHNESS PROPERTIES FOR THE R. E. GINNA UNIT NO. 1 PRESSURE VESSEL SHELL FORGINGS AND WELD METAL FLUENCE 4.9 x 10 n/cm (I> 1MeV)

ENERGY TO TEST CRACK MAXIMUM MAXIMUM ENERGY g

SPECIMEN FORGING TEMP.

LENGTH LOAD LOAD PO TO PQ Ic1 NUMBER NUMBER

(*F)

(IN.)

(LBS)

(IN. LBS) (LBS)

(IN. LBS) (ksi In.

)

S-1 125S255

+78 0.524 9,450 189.6 4,000 18.0 121.7 S-3 125S255

+78 0.528 10,450 234.8 4,000 25.8 120.0 S-2 125S255

+212 0.570 11,200 1115.2 4,000 29.2 244.7*

P-3 125P666

-54 0.535 10,180 589.6 4,000 25.2 185.4 y

P-1 125P666

+78 0.580 11,460 1792.8 4,000 32.6 312.4*

P-2 125P666

+18 0.572 9,500 1121.6 4,000 20.1 309.0*

W-1 Weldment

+78 0.614 7,940 80.9 4,000 13.1 110.1 W-3 Weldment +212 0.545 11,250 527.8 4,000 23.2 186.6*

W-2 Welduent +212 0.533 11,800 531.6 4,000 41.2 137.5*

• Specimen did not fracture i

6.0 DOSIMETRY ANAL.YSIS 6.1 MEASURED ACTIVITY OF DOSLETERS In order to ef fect a correlation between fast neutron (E > 1.0 MeV) exposure and the radiation induced property changes observed in the test specimens, a nu=ber of fast neutron flux monitors were included as an integral part of the Reactor Vessel Surveillance Program.

In particular, Capsule V contained detectors employing the following threshold reactions 54 54 Fe

)g Ni' (n.p) Co Cu63 (n.a) Co60 Np (n.f) Cs (cadmium shielded)

U (n f) Cs (cadmium shielded)

In addition, thermal neutron flux monitors, in the form of bare and cadmium shielded Co-Al wire, were included in order to assess the effects of isotopic burnup on the response of the fast neutron detectors.

The relative locations of the various dosimeters within capsule V are shown in Figure 3; while the radial and azimuthal position of the capsule with respect to the nuclear core and other reactor internals is illustrated in Figure 1.

It ciay be noted that the nickel, copper, and cobalt-aluminum monitors were in the form of wires placed in holes drilled in spacers at several axial levels within the capsule.

The iron dosimetry, on the other hand, was accomplished by drilling samples from the charpy test specimens. The cadmium shielded neptunium and uranium fission detectors were accomodated within the dosimeter block located near the center of the capsule.

The results of the neutron dosimetry removed from capsule V are summarized in Table 11.

The data as presented have been adjusted to account for decay in the period between reactor shutdown and time at which activity measurements were obtained.

The nomenclature defining the location of the iron monitors corresponds to the identification numbers of the individual charpy specimens f rom which the iron drillings were obtained.

TABl,E 11 RESULTS OF FAST NEUTRON DOSIMETRY Reaction Activity and Measured at Fast Neutron Dosimeter Activity Removal Fluence (E > 1 Mev)

Location (DPM/MR)

(DPM/Ma)

(n/cmb Fe (N,P) Mn 5

18 W-1 1.48 x 10 1.82 x 10 5.52 x 10 5

5 18 R-1 1.54 x 10 1.89 x 10 5.73 x 10 5

18 S-6 1.31 x 10 1.61 x 10 4.87 x 10 5

5 18 P-7 1.54 x 10 1.89 x 10

5. 73 x 10 5

5 18 W-2 1.22 x 10 1.50 x 10 4.55 x 10 5

5 18 R-3 1.17 x 10 1.44 x 10 4.35 x 10 5

8 S-8 1.21 x 10 1.49 x 10 4.51 x 10 5

18 P-9 1.26 x 10 1.55 x 10 4.70 x 10 S8(N,P) Co$

Ni 6

6 10 Center 1.43 x 10 3.5 3 x 10 3.73 x 10 Cu (N.a) Co 3

18 Top 4.4 3 x 10 4.28 x 10 5.42 x 10 3

3 8

Top - Middle 4.06 x 10 3.92 x 10 4.97 x 10 1

Bottom - Middle 4.49 x 10

4. 34 x 10 5.50 x 10 Botton 4.88 x 10 4.72 x 10 5.97 x 10 3

Np (n,f)

Dosimeter Block 5.64 x 10 '*

5.64 x 10 '*

4.40 x 10 1

1 0

U (n,f) 1*

Dosimeter Block 6.08 x 10 6.08 x 10 4.40 x 10

• Fission Dosimeter Response is Measured in Units of Fissions / Capsule 6.2 ANALYTICAL METHODS An accurate evaluation of the response of any given monitor removed from the surveillance capsule, is dependent on a knowledge of the energy spectrum of the neutron field to which the monitors were exposed. For this analysis, the PLMG one-dimensional multigroup diffusion code was employed to calculate the spectral data at the location of interest.

Briefly, PIMG utilizes a 55 neutron energy group scheme (54 groups in the energy range 0.625 ev to 10 MeV and one thermal group below 0.625 ev) along with a library of cross-section data to compute spectral distributions within a geometry of interest. The R. E. Ginna reactor geometry employed here included a description of the radial regions internal to the primary concrete (core barrel, thermal shield, pressure vessel, and water annuli) as well as an appropriate reactor core fuel loading pattern. Thus, distortions in the fission spectrum due to the attenuation of the reactor materials are accounted for in the analytical approach. The relative neutron energy spectrum calculated to exist at the capsule location is given in Table 12.

Having the neutron energy spectrum at the location of literest, the fluence levels experienced by the activation and fission monitors were deduced from the following equations.

6. 2.1 Activation Monitors Suppose the product activity was reported as D dPm/mg of product. The following relation holds n

60CN g--

- 1A

- A (T - t )

D=

f 2,e(E)4 (E) gi, F

(1 - e

3) e E

J=1 Tcblo 12 PIMG GROUP FLUXES AT SURVEILLANCE CAPSULE LOCATION Lowe r Lower Bound Relative Bound Relative Group Energy Neutron Group _

Energy Neutron No.

(MeV)

Flux No.

(MeV)

Flux

-2 1

7.79 0.0819 19 8.65 x 10 1.35 2

6.07 0.223 20 6.74 x 10~

1.29

~

3 4.72 0.540 21 4.09 x 10 2.33

-2 4

3.68 0.597 22 2.48 x 10 2.11

-2 5

2.86 0.785 23 1.50 x 10 1.93

-3 6

2.23 1.24 24 9.12 x 10 1.85

-3 7

1.74 2.66 25 5.53 x 10 1.84

~3 8

1.35 2.21 26 3.35 x 10 1.84

-3 9

1.05 1.81 27 2.03 x 10 1.82

-3 10 0.821 2.51 28 1.2 3 x 10 1.81 11 0.639 2.64 29 7.50 x 10 '

1.80

~

12 0.498 2.66 30 4.54 x 10 1.80 13 0.387 2.23 31 2.75 x 10

1.80 14 0.302 2.23 32 1.67 x 10

1.80 15 0.235 2.09 33 1.30 x 10

O.898 16 0.183 1.78 34 1.01 x 10 '

O.899

~

-5 17 0.143 1,63 35 7.87 x 10 0.898 18 0.111 1.52 i

l i,

o where:

C is a normalizing constant N

is Avogadro's number A

is the atomic weight of the target element f

is the weight fraction of the target isotope.

g o(E) multigroup activation cross-sections for the reaction of interest

$(E) =ultigroup neutron fluxes obtained from PIMG A

is the decay constant for the product isotope F) fraction of full reactor power during j time interval, t) th t) length of j irradiation period T

elapsed time between initial reactor startup and sample counting The values of F) and lengths of time intervals T and t) depend on the operating history of the reactor in question and were obtained from daily operating charts of the RGE reactor.

From Equation 1, the constant c may be calculated and the fluence above energy E is given by L

n

$ (E > () = C [

$ (E)

F) t)

(2)

E>(

j=1 n

where I F)t) is the total effective full power seconds converting neutron j=1 flux to fluence.

6.2.2 Fission Monitors The analysis of both the U and Np fission detectors are based on the production of Cs In gcneral, the Cs activity within e dosimeter follow-ing fission of Np or U is given by n

G = KyM {$(E)o (E)

F) (1 - e d) e (3) l E

j=1 l.-.

where G

is the C, activity measured in units of dPS/ capsule K

is a normalizing constant y

is the fractional yield of C in fast fission of N or U 237 238 P

N is the number of target atoms of N or U per capsule g

p o (E) are the multigroup fission cross-sections for N 237 238 f

and U p

In the same manner as for the activation monitors, the constant K may be obtained from Equation 3 and the fast neutron fluence derived from the following expression.

n

}{ F) t

$ (E > E ) = K 4 (E)

(4)

E>(

j=1 The reaction rate cross-sections employed in the dosimetry analysis are summarized in Table 13.

6.2 3 Projections of Sample Fluence to the Reactor Vessel For standard 2 loop reactor vessels and internals the dimensions given in Table 14 apply.

TABLE 14 SAMPLE CENTERLINE RADIUS VESSEL I.R.

Loops inches inches 2

62.375 66.0 l

The fact that samples are located within the reactor vessel means that they

~

accrue fast neutron (E > 1 Mev) exposure, nem at a faster rat, than the adjacent vessel. Where possible, samples are also placed at angular locations of relatively high fast neutron exposure so that specimens in the capsule are irradiated at a rate faster than the maximum rate on the vessel wall. The ratio of these rates is called the sample "lead f actor".

Lead factors for the samples in standard reactor vessels are quoted in Table 15.

l l l l

l

o TABLE 15 SAMPLE LEAD FACTORS Loops Sample Angular Location Lead Factor 2

57' 1.9 2

67*

1.7 2

77*

3.3 The data presented in Table 15 may be employed to correlate the measured exposure at the sample location with the peak exposure of the pressure vessel for corresponding irradiation period.

6.3 Results of Analysis The f ast neutron (E > 1.0 MeV) fluence levely derived from the activities of the various monitors are presented in Table 11.

Although the monitors employed in the surveillance program do not all respond with a 1.0 MeV threshold, equivalent 1.0 MeV fluence levels were derived from the measured data by means of the following expression.

4( )dE E > 1.0 MeV

$ (E > 1.0 MeV) = $ (E > E ) )

I (5) l [g,

• (E)dE l

g L

where: 4 (E > 1.0 MeV) is the neutron fluence above 1.0 MeV and the remaining variables have been defined in the preceding section.

The validity of the above expression is, of course, predicated on an f

accurate determination of the neutron energy spectrum over the energy i

range of interest. The accuracy of the energy spectrum as calculated by the PIMG code is an implied assumption underlying this entire analysis.

By comparing the calculated fluence levels on the basis j

of a common threshold, a check on both the accuracy of the fluence derived from the iron data and the validity of the assumed energy l

-41

TABLE 13 REACTION X-SECS USED FOR SPECTRUM ANALYSIS Reaction Cross-Section (Barns)

No Np (n.f)

CS U (n.f)

CS Fe (n.p) Mn f Ni (n.p)S8 Co 63cu (n.a)60Co 1

2.277 0.970 0.592 0.607 0.035 2

1.890 0.859 0.572 0.608 0.0098 3

1.419 0.553 0.464 0.535 0.00085 4

1.466 0.533 0.324

0. 388 5

1.59 0.512 0.145 0.222 6

1.685 0.505 0.0494 0.113 7

1.685 0.494 0.0194 0.0371 8

1.688 0.330 0.0089 0.0112 9

1.632 0.059 0.0015 0.0043 10 1.196 0.010 11 1.042 0.002 12 0.609 0.0008 13 0.3 0.00013 14 0.107 15 0.0402 16 0.0282 17 0.0295 18 0.0182 19 0.0131 20 0.0124 21 0.02 22 0.0194 23 0.0172 24 0.0149 25 0.0121 26 0.0114 27 0.0107 28 0.0102 29 0.01 30 0.01 31 0.01 32 0.01 33 0.01 34 0.01 35 0.01 spectru:n is achieved. The iron data may then be used with some degree of confidence to correlate fluence levels with radiation damage.

9 REFERENCES 1.

S. E. Yanichko, "Rochester Gas and Electric Robert E. Ginna Unit No.1 Reactor Vessel Radiation Surveillance Program," Westinghouse Nuclear Energy Systems - WCAP-7254, (May 1969).

2.

ASTM Designation E185-66, "Surveillance Tests on Structural Materials in Nuclear Reactors," Book of ASTM Standards Part 31 (1967).

3.

E.B. Norris, Southwest Research Institute Laboratories, unpublished test records and data submitted to Westinghouse Nuclear Energy Systems.

(November 1972).

4.

W. S. Hazelton, S. L. Anderson, and S-E. Yanichko, "Basis for Heatup and Cooldown Limit Curves." Westic house Nuclear Energy Systems -

WCAP-7924 (July 1972).

5.

Proposed Technical Specifications For The 1520 MW Rating R. E. Ginna g

Unit No. 1, November 18, 1971, License Application Docket No. 50-244.

6.

G.R. Irwin, "Fracture Mechanics," Structure Mechanics, Pergamon Press, New York City and London (1960).

7.

R. E. Johnson, Fracture Mechanics: A Basis for Brittle Fracture Pre-vention, WAPD-TM-505, AEC Research and Development Report (November 1965).

8.

Fracture Toughness Testing and Its Applications, ASTM STP 381 (April 1965).

9.

L. E. Steele and R. H. Sterne, Jr., "Steels For Commercial Nuclear Power Reactor Pressure Vessels, "Nuclear Engineering and Design, Vol. 10 (1969).

10.

T. R. Mager, Westinghouse Nuclear Energy Systems Unpublished data.

l 11.

T. R. Mager, HSST. Technical Report No. 9, "Post-Irradiation Testing of l

2T Compact Tension Specimens," Westinghouse Nuclear Energy Systems WCAP-7561 1

(August 1970).

12.

F. J. Witt and T. R. Mager, A Procedure for Determining Boundary Values On Fracture Toughness K At Any Temperature, 5th Symposium on Fracture 7

Mechanics, University of Illinois (August 1971).

13.

H. Bohls Jr., et al, "PIMG-A One Dimensional Multigroup P Code for the y

IBM-704 " WAPD-TM-135, July 1959.

i 14.

ASME Boiler and Pressure Vessel Code,Section III, Summer 1972 Addenda, l

Non-Mandatory Appendix G.

1 ;

i

APPENDIX HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION 1.0 HEATUP AND C00LDOWN LIMIT CURVES IN ACCORDANCE WITH ASME B0llER AND PRESSURE VESSEL CODE, SECTION III.

Heatup and cooldown limit curves are calculated using the most limitino value of RTNDT (Reference Nil-Ductility Temperature) detemined as follows.

Detemine the highest RT of the material in the core NDT region of the reactor vessel using the preservice reactor vessel material properties and estimating the radiation induced a RT RT is defined as the Reference Nil-NDT.

NDT Ductility Temperature.

It is designated as the higher of the drop weight nil-ductility transition temperature (NDTT) or the temperature where the material exhibits at least 50 ft-lbs impact energy and 35 mil lateral expansion (in the transverse direction) minus 60*F.

Given the copper content of the most limiting heat of material, the radiation induced a RT can be estimated from Figure A1.

Fast neutron NDT fluence (E > 1 MeV) at the 1/4 T (wall thickness) and 3/4 T (wall thickness) vessel locations are given as a function of full power service life in Figure A2.

The data for all other ferritic materials in the reactor coolant pressure boundary are examined to assure that no other component will be limiting with respect to RT NDT*

2.0 FRACTURE TOUGHNESS PROPERTIES The fracture toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the ASME Boiler 3 and the calculation methods of Reference 4 and Pressure Vessel Code The preirradiation fracture toughness properties of the R. E. Ginna reactor vessel materials are presented in Table Al.

The post-irradiation fracture tougnesss properties of the reactor vessel beltline material (given in the body of this report) were obtained directly from the R. E. Ginna Reactor Vessel Material Surveillance Program.

s TABLE A1 j

RGE REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED) 50 FT-LB/35 MILS UPPER SHELF HgANSION RT (FT-LB)

MATL

)

()

LON RANS (F

CL.HD. DOME A302,B NA 0.10

-20 52 125**

65 NA NA CL.HD.FLG A508,2 NA 0.007

-75

-50

-35**

-75 140*

91***

VES.SH.FLG A508,2 NA 0.007

-82

-30 8**

-52 115*

75***

INJEC.N0Z A508,2 NA 0.013 60*

65 95**

60 66 43***

INLET N0Z A508,2 0.09 0,012 60*

-20 35**

60 107 70***

g INLET N0Z A508,2 0.09 0.010 60*

-22 15**

60 115 75***

I OUTLET N0Z A508,2 0.09 0.011 60*

O 45**

60 109 71***

OUTLET N0Z A508,2 0.09 0.012 60*

22 60*

60 88 57***

UPPER SHL A508,2 NA 0.010 40 12 20**

40 173 112***

INTER SHL A508,2 0.07 0.010 20 40 105**

45 120 78****

LOVER SHL A508,2 0.05 0.010 40 10 30**

40 164 107***

TRANS. RING A508,2 NA 0.008 10

~30

-13**

10 182 118***

BOT.HD. DOM A302,B NA 0.010

-50 35 85**

25 NA NA WELD Weld 0.23 0.012 0*

NA 45 0

NA 74 HAZ HAZ NA NA NA 55 NA NA 77 NA I

• No data, conservative estimate

• • Estimated from longitudinal data, temperature at which minimum 77 ft-1b/55 mils lateral expanaton obtained

• • • Longitudinal value reduced 35%

3793 18 3Ts3-I OVER AND WELD r

0.25 n

Cu OVER 8A$E. T AND 0 30 ELD g wEte App oygg 3/4T m A$

B ASE. T Cu 0.10 OVER AND 0.15 WELD 8 A$f, 5 cu 0 05 0 10

~

I l

l l

l l

}'8 2

4 6

8 1038 2

4 6

8 1020

~

FLUENCE (N/CM2 >l MEV) l l

25 30 35 unction of Full e Al. Effect of Fluence and Copper Content on Shtft of RTNDT far Reactcr Venel - Steels Expesed to 5500F Temperature 46

ale Pressure-Temperature Relationships calculation temperature relationships for various heatup and ch the 3/4 T

alculated using methods derived fmm Non-Mandatory tuation at the an III of the ASME Boiler and Pressure Vessel Code.

'I scussed in detail in Reference 4.

'9

• • E 5, tend to es that the allowable tutal stress intensity factor i stresses are.

ing heatup or cooldown cannot be greater than that e time (or water ve (Reference 1) for the metal temperature at that e the thermal the approach applies explicit safety factors of sing heatup ess intensity factors induced by pressure and themal the preceeding oly. Thus, the governing equation for the heatup-

.erest must be 2Kgg + 1.25 kit

• KIR (1) is f r both the inal limit curves asite curve is ss intensity factor caused by membrane (pressure) he steady state is intensity factor caused by the thermal gradients.

e, the allowable g

by the code as a function of temperature relative

, of the material.

. hen adjusted to 4

iture sensing ilysis, Equation (1) is evaluated for two distinct 19 heatup limitations isure-temperature relationships are developed for h that over the course ero rate of change of temperature) conditions assuming from the 0.0. to the ode reference 1/4 T deep flaw at the ID of the pressure imes, be based on the

'act that, during heatup, the thermal gradients in the iroduce compressive stresses at the 1/4 T location, is hat for heatup, induced by internal pressure are somewhat alleviated.

always at the'I.C.

orature curve based on steady state conditions (i.e.,

represents a lower bound of all similar curves for to produce tensile

-hen the 1/4 T location is considered, ses at the 0.0. position.

or on kit represents additional conservatism above _

i 4

i 2600 2400 2200 2 2000 2

1800 HE ATUP RATES I

E 1600

' O " U ""

CRI TIC ALI TY U

LIMIT w

1400 E

E o

I200 5

1000 E

800 600 400 200

]

I l

l I

I

.0 0

50 100 150 200 250 300 350 400 INDICATED TEMPERATURE (OF)

Figure A-3. RGE Reactor Coolant System Hootup Umitations Applicable for 7.0 Effective Full Power Years. TERROR =

,PERROR" # OI

2600 2400 2200 E 2000 mi 1800 w

5 1600 1400 w

o.

(,,

o i200 3

1000 cootoowM R ATE S g

800 (o,fug) z 600

- O N

2o N-400 - foof 200 i

I I

I I

I o

O 50 100 ISO 200 250 300 350 INDICATED TEMPERATURE (OF)

F iguie A-4.

RGE Reacto. Coolont System Cooldown Limitations Applicoble y

for 7.0 Effective Full Power Years. T

~

ERROR

' ERROR

4 i

1 2600 --

2400 -

2200 -

i g 2000 -

E 1800 g 1600 HEATUP RATES TO 60'F/ttR E 1400 U

w l200 O

ca:TicAtiiv

1000

""'I b

800 E

i 600

/*

400 l

l 1

1 I

i 1

1 1

200

-0 0

50 100 150 200 250 300 350 400 450 INDICATED TEMPERATURE (OF) wy Figure A-5. RGE Reacter Coolant System Hootup Limitations Projected to End of Plant T

= 10*F, P

= 60 PSI ERROR ERRCt

i a.s.

3.0 Surveillance Capsule Removal Schedule

[

To date, Capsule V has been removed and the encapsulated specimens have been tested.

Based on the post irradiation test results of Capsule V, the reconnended removal schedule for the remaining capsules in the R. E. Ginna reactor vessel is as follows:

Factor By Which Capsule Removal Capsule Identification Leads Vessel Maximum Exposure Time End of 2nd R

3.3 Core Cycle S

1.9 10 years N

1.9 20 years T

1.7 30 years P

1.7 Standby All remaining capsules contain weld metal specimens. - - - _.

..