ML19093A512

06/24/1975 Final Report on Unit 1 Pressure Vessel Irradiation Capsule Program, Examination and Analysis of Capsule T

ML19093A512
Person / Time
Site:	Surry
Issue date:	06/24/1975
From:	Denning R, Farmelo D, Lowry L, Perrin J, Wooton R Battelle
To:	Office of Nuclear Reactor Regulation, Virginia Electric & Power Co (VEPCO)
References
Download: ML19093A512 (54)

Text

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SURRY UNIT NO. 1 PRESSURE VESSEL

~-#

IRRADIATION CAPSULE PROGRAt~:

EXAMINATION AND ANALYSIS OF CAPSULE T to VIRGINIA ELECTRIC AND POWER COMPANY*

June 24, 1975 by J. S. Perrin, L. M. Lowry, D. R. Fannelo, R. O. Wooton, and R. S. Denning JUL* 7 1975 PRODUCTION SERVICES BATTELLE Columbus Laboratories 505 King Avenue Columbus, Ohio. 43201 l_

TABLE OF CONTENTS

SUMMARY

• . * . . .. 1 INTRODUCTION. 2 CAPSULE RECOVERY AND DISASSEMBLY .* 4 SAMPLE PREPARATION * * * * *

• 5 Pressure Vessel Material . .. . 5 Correlation Monitor Material 7 EXPERIMENTAL PROCEDURES * * * * *

• 7 Dosimeter and Thermal Monitor Examination ** 7 Tensile Properties * ,

• 9 Charpy Impact Properties

• 12 RESULTS AND PISCUSSION. 15 Dosimeter and Thermal Monitor Examination ** 15 Tensile ~roperties * * *

• 21 Charpy Impact Properties, 21 CONCLUSIONS *** , 43 REFERENCES. * * * * , *

• 45 APPENDIX A PRESSURE VESSEL MATERIAL * * . . * *

• A-1 APPENDIX B ASTM CORRELATION MONITOR MATERIAL

• B-1

"' I SURRY UNIT NO. 1 PRESSURE VESSEL IRRADIATION CAPSULE PROGRAM:

EXAMINATJON AND ANALYSIS OF CAPSULE T by J. S. Perrin, L. M. Lowry, D.R. Farmelo R. O. Wooton, and R. S. Denning The irradiation conditions and the irradiation induced changes in mechanical properties of the Surry Unit No. 1 reactor pressure vessel have been determined from evaluation of specimens contained in Capsule T. This capsule contained base metal, weld metal, and heat-affected-zone metal specimens. The capsule was removed after 1.07\equivalent full-power years of operation. The irradiation temperat~re did not exceed,598. ~ and the 10 c&psule received a fluer..ce of 2.5 x 10 nvt (>l Nev). }!lt-:qtfq-'

The measured chang~ the 50 ft-lb Cl~arpy . i~pac@transition temperature for the three reactor materials were consistent with those observed for other programs involving similar materials and irradiation con-ditions. Of particular interest is the upper shelf behavior of the weld metal. The upper shelf energy level dropped from a preirradiation value of approximately 70 ft-lb to an irradiated value of 53 ft-lb. A trend band for the change in the 30 ft-lb and 50 ft-lb transition temperature with tncreasing /

exposure to fast neutrons was determined from the results of this program and :*

those of other programs, The tensile specimen examination showed that the yield}

and ultimate tensile strengths of-the base and weld metal increased, and that the reduction in area and total elongation decreased.

2 INTRODUCTION This report presents the results of the examination of Capsule T, the first capsule of the continuing surveillance program for monitoring t.he effects of neutron irradiation on the SA 533 Grade B Surry Unit No. 1 reactor pressure-vess_el material under actual operating conditions. This report contains experimental procedures, results, and discussion relating to the investigation.

Radiation damage studies initiated during the early days of nuclear pmier-reactor development revealed the deleterious effects of high energy neutrons upon the notch ductility of reactor vessel steels. The effect was characterized by a rapid rise in the transition temperature with increasing neutron exposure. In addition, the tensile properties show a significant in-crease in yield strength and tensile strength, accompanied by.a loss of uniform elongation and reduction of area with increasing neutron exposure.

Suffi::::.icnt data on the effects of radiation on the* mechanical properties of reactor pressure-vessel steel are now available to indicate the type and relative magnitude of property changes to be encountered during the expected lifetime of the reactor structure. This information is an integral part of the design basis for a nuclear reactor. During the reactor life the operating limitation curves (i.e., pressure and temperature) will be periodically ad-justed to incorporate the projected changes in mechanical properties.

To further ensure the continued safe operations of the plant, a reactor-vessel radiation-surveillance program is being conducted. The primary purpose of this program is to evaluate the specific changes in the mechanical properties of the pressure-vessel materials under the actual service conditions* (neutron fluence, time, and temperature) of the reactor plant. It is known that the.

magnitude and relationships of the property changes are functions of the specific material composition and metallurgical conditions; the amount, rate, and energy

' (1-7)*

spectrum of the radiation; and the exposure temperature

• The surveillance program is designed to provide information for determining whether the reactor pressure-vessel operating limitations are indeed conservative, _as is expected.

• References at end of text.

.. 'II 3

The surveillance program for the reactor was designed and recommended by the Westinghouse Electric Corporation and is based on ASTM El85, "Surveillance Tests on Structural Materials in Nuclear Reactors"(S). The details of this program and the preirradiation mechanical properties of the materials are presented in Reference (9). Prior to startup, eight capsules containing tensile, Charpy V-notch, and WOL fracture-mechanics specimens of the pressure-vessel materials were installed in the reactor. The capsules were located between the thermal shield and the vessel wall. In addition to these mechanical-property test specimens, the cap*sules contain thermal-monitor and neutron-fluence speci-mens for evaluation of the specific temperature and-radiation exposure conditions of the specimens.

The particular exposure condition variables evaluated are the total integrated fast fluence of the capsule and the maximum temperature encountered by the specimens during the exposure period. The temperature history of the surveillance capsule is fairly representative of that encountered by the.

press11r.e-vessel wall. However, the capsule j_s a f:i:.nite distance from the reactor pressu~e-vessel wall and, therefore, the capsule received an accelerated fluence as compared to the vessel wall.

The most essential mechanical properties evaluated by the test specimens in the surveillance capsule are the ductile-to-brittle fracture transition temperature and the conventional tensile-strength and ductility values. In this context, essential refers to those requirements of the current methods for establishing pressure-temperature operating limitations of the reactor pressure vessel. An essential requirement of the mechanical property measurements is that they be made on representative materialc For this sur-veillance program, the capsules contain test specimens of the SA 533 Grade B reactor-vessel steel from two 7-1/8 in. thick shell plates from the vessel intermediate and lower shell courses, and also weld metal and heat-affected zone (HAZ) metal. The thermal history of the material used to fabricate test specimens is as identical as possible to that received by the reactor pressure vessel during fabrication. In addition to the reactor materials, test specimens.

of a specially prepared correlation material (A 533 Grade B) made available by Subcorrnnittee II of ASTM Committee ElO were also contained in the capsule,

o* I 4

The data obtained from evaluations of the correlation material provide a valuable link with the surveillance programs of other nuclear-reactor pressure

• cJ

• vessels.

The main test of the report contains the results of the thermal

. monitor and neutron dosimeter examination, the Charpy impact specimen examinations, and the tensile specimen examination. The WOL specimen examination is presented in an appendix to the report.

CAPSULE RECOVERY AND DT SAS SENRI,Y Battelle's Columbus Laboratories (BCL) personnel went to the Surry Nuclear Plant to pick up the surveillance capsule assembly. They brought a pool-side jib crane, a specialized underwater cutting tool, and a shipping cask. The cutting head of the underwater cutting tool is a mild st~el casting.

The bea<l had been sand blasted, copper plated, and then nickel plated and_epoxy painted to prevent it from rusting and thereby contaminating the pool water.

To further avoid contaimination, pool water was used in the line leading from the pump intensifier unit to the cutting head.

The capsule assembly had an overall lengthof approximately 112 inches.

Surry personnel removed the capsule assembly from the pressure vessel and trans-ferred it underwater in a canal to the spent fuel pool. The upper lid and lower drain lid were removed from the shipping cask. Using an overhead crane, the shipping cask and associated cannister.were then raised from the receiving area, moved to a position over the spent fuel pool, and lowered into the pool so that the bottom end of the cask was resting on the floor of the pool. The bridge crane was then used to position the capsule and attached lead tube such that the capsule was in the cask except for the lead tube and about 6 inches of the capsule.

The cutting tool was lowered into *the pool using a stainless steel cable attached to the pool-side jib crane, The cutter was guided into position using the stainless steel pipe line leading to the cutting head. The initial cutting position was determined by knowing the distance from the bottom of

5 the cask c_anister to the top of the canister (41 inches) and the top of the canister to the cutting edge of the tool (2-1/2 inches). After lowering the capsule to the bottom of cannister, it was raised 3 inches and cut at 40-1/2 inches. The lead tube was thereby separated from the capsule assembly because this cut was about 2 inches above the top end of the capsule. The lead tube was placed into the cask along side the capsule after one additional cut was made to separate the lead tube into two pieces.

The cask was raised from the pool and its exterior was thoroughly rinsed with water. The. water inside the cask was allowed to drain into the gr~ting around the pool. The cask was decontaminated at pool-side to the level of removable contamination allowed for shipping, 2200 disintegrations/

2 2

• 100 cm /min ~y and 220 disintegrations/100 cm /mina.

• The cask was then shipped to the BCL Hot Laboratory Facility by commercial carrier *.

Upon arrival at BCL, the cask was placed in a hot cell. The capsule and lead tube sections were then removed from the cask. The capsule was examined to confirm the _cut separating the lead tube from the capsule *was in the lead tube.

The specimens were then removed from the capsule and inventoried.

The specimen inventory is shown in Table 1, which is in agreement.with the inventory in WCAP 7723( 9 ). Before testing, the mechanical property specimens were cleaned in the following manner.

1. Prewashed in a solution of Radiac and.water.

2. Ultrasonically cleaned for.a minimum of 45 minutes in a solution of detergent and water.

3. Removed from ultrasonic bath, rinsed with clean water, and then rinsed with reagent grade alcohol.

SAMPLE_ "PREPARATION Pressure Vessel Materjal Babcock and Wilcox supplied the SA 533 Grade B Class 2 steel used in the irradiation capsule program. Appendix A contains the heat treatments and chemical analyses of these materials.

6 TABLE 1. SPECIMEN IDENTIFICATION AND LOCATION IN THE SURRY UNIT NO. 1 IRRADIATION TEST CAPSULE Specimen Capsule Type T Charpy V H7, HS Cl.

0 W7, W8 f--1 Charpy V Specimen Code HS, H6 WS, W6 V - Plate Cli-415-1 w- Weld Metal Tensile Wl, W2 H - Weld Heat Affected WOL W2 Zone R - ASTM Correlation Charpy V l-13 > H4 Moniturs W3, W4 WOL Wl Specimen Orientation Charpy V Hl, H2

(!)

Wl, W2 VESSEL ffiill r-1

l Dosimeter Cl)

Cl.

64 co Charpy V C) V R39, R40 V47, V48

-CORE -

WOL Vl4 Charpy V R37, R38 V45, V46 WOL Vl3 Tensile V9, VlO Charpy V

• R35, R36 s0 J

V43, V44 J

0 Charpy V l:Q R33, R34 Vl~l, V42

7 Correlation Mon]tor Material The ASTM correlation monitor material used in the present program was supplied by Oak Ridge National Laboratory. Appendix B contains a description of this material including heat treatment and chemical analysis *

.EXE.ERIMRNTAL PROCEDURES This section describes the procedures employed in the testin~ of the impact and tensile specimens. Also included are the procedures used to examine the dosimeters and thermal monitors. All experimental examination and evalu-ations were conducted at Battelle's Columbus Laboratories *

.llilliimeter and Thermal Noni to 1: Exc;;miJ:ill.tiQn The capsule contained two kinds of low-melting-point -eutectic alloy thermal monitor wires for determination of the maximum temperature attained by the test specimens during irradiation, One alloy was 2.5% Ag - 97.5% Pb with a melting point of 579 F. The other alloy was 1.75% Ag - 0.75% Sn -

97.5% Pb with a melting point of 590 F. These thermal monitor alloys *were sealed in Pyrex tubes and inserted in spacers in the capsule. During capsule disassembly the thermal monitor wires were removed from the spacers and Pyrex

.tubes. The wires were then visually examined for evidence of melting at a magnification of l~X using a stereomicroscope.

The capsule contained a total of 14 dosimeters of copper, nickel, cadmium-shielded aluminum-cobalt alloy, unshielded aluminum-cobalt alloy, neptunium 237 and uranium 238 in three locations. In addition, the Charpy impact specimens provided material for iron dosimeters. The reactions used for the dosimetry calculations were as follows:

8 Iron 54Fe (n,p) 54Mn Nickel 58Ni (n,p) 58Co Copper 63Cu (n,a) 60Co Cobalt 59Co (n,Y) 60Co Uranium ~38u (n,f) 137Cs Neptunium 237Np (n, f) 137 Cs All 14 dosimeter samples were analyzed.

After removal from the capsule, the individual samples were placed in via.ls for transfer to the radiochemistry laboratory. Radiation readings at 1 meter and on contact were recorded. The nickel, copper, and cobalt wires ~vere decontaminated by wiping with dilute acid, distilled water, and reagent gra de acetone. Th e iron samp 1 es, an d 238U an_d 237'1\T

~P capsu 1 es were wiped with dilute acid and distilled water to remov3 major cont:amination and then cleaned ultrasonically in a solution of Radiac and water.

The copper, nickel, and Al-0.15 Co wires were weighed to +/-0.0001 g, and the activation product intensities were detennined directly by gamma ray spectrometry. For the iron samples drillings were taken through a complete cross section near the impact area of the designated Charpy impact specimen.

The drillings were then mounted on a standard counting ring and garrrrna counted.

238 u and 237 Np capsu 1es were opene. d in

. an a 1 par~

h d.iation containment box by specially prepared tools used to grip the small 1/4 in. diameter x 3/8 in.

long cylinders and cut off the tops. The tool used for cutting off the tops 238 was a modified tubing cutter. The u and 237 Np were present in the form of oxide powders. The two samples were poured into small tared primary contain-ment via.ls and then into clean tared secondary vials for weighing to +/-0.0001 g on an analytical balance. They were dissolved in 81:1 HN0 3

(u3 o8 ) and 137 8.M H so -0. l.M NaBro (Np0 ), and diluted to appropriate volumes._ Cs analyses 2 4 3 2 were performed in duplicate after purification by the chloroplatinate method.

For assurance of complete fission product decontamination, Zr and Ru holdback carriers were employed, and an extra scavenge precipitation step was performed,

9 All the activation products were analyzed by gamma-ray spectrometry utilizing a 3 in. diameter x 3 in. long Nal (Tl) scintillation crystal detector and model 401D 400 channel analyzer (Technical Measurements Corp) capable of 137 137 ~a 7 percent resolution FWHM (full width half maximum) at the 0.663 Mev cs-54 60 137 gamma ray energy level. The Mn, Co, and Cs samples were counted directly NBS stan d ar d s. Th e 58 co activity

. . *was ob taine

. d f rom comparison against . with theoretical efficiency curves prepared from NBS standards.

The procedure.s used in the evaluation* of the dosimetry samples followed the appropriate ASTM recommendations(lO-l 6).

Tensile Properties The design of the tensile specimens is shown in Figure 1; the gage section has a nominal 0.250-in. diameter and a nominal 1.000-in *. length. The tensile tests were conducted on a screw-driven Instron testing mc1chine* havirig a 20,000 lb. capacity. A crosshead speed of 0.05 in. per min was used. The deformation of the specimen was measured using a strain gage extensometer.

The strain gage unit senses the differential movement o.f two extensometer extensio*n arms attached to the specimen gage length 1 in *. apart. The extension arms are required for thermal protection of the strain gage unit during the elevated temperature tests. Figure 2 shows the extensometer extension arms and strain gage assembly used for tensile testing. A tensile specimen is shown at the top of the.figure next to the region of the extension arms where the specimen is loaded for testing. The strain gage unit is shown at the bottom of the figure next to the region of the extensometer arms where the unit is attached during testing. The extensometer was calibrated before testing using an Instron high-magnification drum-type extensometer calibrator.

Curves were run in a load-elongation mode until the vicinity of maximum load.

The curves we*re then finished in a load- time mode so that a complete curve would be generated in case of extensometer slippage after the specimen started to neck down locally.

1.005 GAGE LENGTH

.995

_ _,-,1._ _ .*

.25G DIA

_*2 _1 x j-. . ~,

~'..\---\.:.........-:.-..-+-1-1

.25!

r2c.9 DIA 1

r~~DIA.,

i Ii 4_Lt *

. t;; .

llii:. .

+.

I .

l1_250 REDUCED

.250 R11 81/

.255 TYP I~ * .

• 1** 4 9_:: . ;)0"'- - - ~ - ,

'197 J.2GO SECTION 14 8 1-4----------4.250~----------N 4.2T6

.630

~620 ~

JG) t. *.

,790

~ .786

,395 I

,3;3_ t I

<t_ OF HOLES TO BE ,375 D1A(2l FOR WITHIN .002 OF ,377 IT, 2 ONLY SECTION A-A TRUE ct_ OF SPECIMEN NOTES:

!*LATHE CENTERS REQUIRED*

. 2.12W ALL OVER UNLESS OTHERWISE SPECIFIED,.

FIGURE 1, TEN*s ILE SPECIMEN

P4973 FIGURE 2. EXTENSOMETER EXTENSION ARMS AND STRAIN GAGE ASSEMBLY USED FOR TENSILE TESTING

.. 1 12 Elevated temperature tensile tests were conducted using a three-zone split furnace. The irradiated tensile specimens were tested at room temperature

• and 550 F. The specimens were held at temperature before testing to stabilize the temperature. Temperature was monitored using a Chromel-Alumel thennocouple in direct contact with the gage section of the specimen. Tempera-ture was controlled within+/- 3 F.

The load-extension data were recorded on the testing machine strip chart. The yield strength, ultimate tensile strength, and total elongation were determined from these charts. The reduction in area was determined from specimen measurements made using a vernier caliper.

Charpy Impact Properties The impact properties were determined using a standard 240 ft-lb Satec-Bald-win Model SI-lC impact machine in accordance with the recormnendations of the pertinent ASTM standard (l?). The accuracy of the machine was verified with standards purchased from the U.S. Army Materials Research Agency. The results are given in Table 2. The design of th~ Charpy imp2.-:!t specimens iB ~h.0wn i.n Figure 3. The velocity of the hammer at the striking position is 17.0 ft/sec.

The 240 ft-lb range was used for all tests. The energy loss due to friction of the machine was determined daily during use of the impact machine, This was done by the following; (a) releasing the pendulum from the 240 ft-lb upright

.position with no specimen in the machine and determining the indicated energy value is Oft-lb; (b) without resetting the pointer, again releasing the. pendulum from the 240 ft-lb upright position and permitting it to swing 11 half cycles, After the pendulum starts its 11th cycle, the pointer is moved to between 12 and 24 ft-lbs and it is determined that the indicated value, divided by 11, does not exceed *o.4 percent (9.6 ft-lb) of the 240 ft-lb capacity.

TABLE 2. CALIBRATION DATA FOR THE BCL HOT LABORATORY CHARPY IMPACT MACHINE Average Standar~a)

BCL Energy, Energy, Variation Group ft-lb ft-lb Actual Allowed Low Energy 12.3 12.4 -0.1 ft-lb +/-1. 0 ft- lb Medium Energy 51.1 52.9 -3 .Lr% +/-5.0%

High Energy 68.8 71.6 -3.9% +/-5.0%

(a) Established by U.S. Army Materials and Mechanics Research Center.

.. ~

.Olt R

.009 I

]

J I

.31G

~

.314 1.063 1.053 '*

. 2.125 2.105 6

. * }' ALL OVER UNLESS OTHERWISE SPECIFIED FIGURE 3. CHARPY v.. NQ'I'CH IMPACT SPECIMEN .

14 ASTM procedures for specimen temperature control were utilized. The low temperature bath consisted of agitated methyl alcohol cooled with additions of liquid nitrogen. The container was a Dewar flask which contained a grid to keep the specimens at least 1 in. from the bottom. The height of the bath was enough to keep a minimum of 1 in. of liquid over the specimens. The Charpy specimens were held at temperature for a minimum of at least the ASTM recommended time.

The tests above room temperature were conducted in a similar manner except that a metal container with a liqu{d bath was used, The bath used for temperatures from 70 to 200 F was water, and the bath used for tempera-tures above 200 F was oil. The baths were heated to temperature using a hot plate.

The specimens were manually transferred from the temperature bath to the anvil of the impact machine by means of tongs that had also been hr00.ght to ter:-:pe.rature. in the bath. The specirr,ens were removed frorn the bath and impacted in less than 5 sec. The energy required to break the specimens was recorded and plotted as a function of test temperature as the testing proceeded.

Lateral expansion was determined from measurements made with a lateral expansion gage. Fracture appearance was estimated from observation of the fracture surface, and comparing the appearance of the specimen to an ASTI1

."fracture appearance chart (lS).

r

15 RESULTS AND DISCUSSION Posimeter and Thermal Monitor Examination The capsule contained two 579 F (2.5 percent Ag, 97.5 percent Pb) and one 590 F (1.75 percent Ag, 0,75 percent Sn, 97.5 percent Pb) thermal moni_tors. The monitors were in the form of wire with a square or rectangular cross section, The 579 F monitors were located in the top and bottom regions of the capsule. The 590 F monitor was located in the middle region of the c9-psule.

Monitors were examined at a magnification of 4X using a stereomicroscope.

None of the three monitors showed any evidence of melting. Based on the visual examination of the thermal monitor wires~ it appears that the capsule was not above 579 For 590 Fin the r~gions of the three monitors for any period of time long enough to cause melting of the thermal monitors.

The surveillance capsule ,;,,as in the reac:tor for 391 equivalent. full power days or 1. 07 equivalent full power years. (l 9 )

• The average power level (including shutdown.periods) was calculated from the power-time histograms to be 42.3 percentof the full po,;,er level of 2441 Mwt. The capsule was located at a core position of 285 degrees( 9 ), and had an exposure lead factor 2

with respect to the inner surface of the pressure vessel wall.of 1.52( 0).

Eight fast neutron monitors and six thermal neutron monitors were counted for gamma ray activity to determine integrated fast fluence and thermal fluence, respectively. The fast flux monitors are iron, nickel, copper, uranium and neptunium, and the results are shown in Table 3, Fast

. 18 18 fluence values (E > 1 MeV) ranged from 2,50 x 10 (Fe) and 2.64 x 10 (Ni) to 3.40 x 10 18 2 *

(Cu)n/cm

• Neptunium and uranium values were 2.80 x 10

-:1-s---- and 18 2 2.99 x 10 n/cm , respectively. Because the iron samples are from actual Charpy specimens and the nuclear constants are well established, the average 18 2 iron fast fluence (E > l MeV) of 2 .50 x 10 n/ cm is considered most representative of the five monitors. The nick.el monitor confirms this value.

Copper dosimetry results tend to run high due to uncertainty in nuclar constants

16 and the need for ultra-high purity. One ppm of cobalt impurity will introduce an error approximately 10 percent high since the same activation product, Co-60, is produced. Reasonable agreement is seen from the Np-237 and U-238 monitors.

The Surry Unit No. 1 Technical Specificatiqn indicates that after 40 years of operation at an 80 percent load factor (32 equivalent full power years) the pressure vessel wall will have received an exposure of 4.30 x 10 19 2 . 18 2 n/cm (>1 MeV). This is equivalent to a prediction of 1.44 x 10 n/ cm *-

(>1 MeV) after 1. 07 effective full power years. In the present examination, 18 2 the average capsule fluence for the iron specimens is 2.5 x 10 n/cm

(>1 MeV). Based on a capsule lead factor of 1.52, this corresponds to a wall

• 18 2 18 2 exposure of 1.65 x 10 n/cm (>l MeV). This fluence of 1.65 x 10 n/cm

(>1 MeV) cba-s~iLon-t~e;~-1nvestigation is inv2ood. ~ement with the 18

• 2

• predicted value of 1.4t~ x 10 n/cm (>1 MeV)..

• The thermal neutron fluences for the bare and cadmium covered \

\

cobalt dosi:ne.tC.rs urc ~hown in Tc:.ble 4. The. true thenna.l flu1::11ce was equal to 2.56 x 10 18 2 n/cm , .calculated from.

R - 1 nvtt rue == Cob are x -R--. where ).

ca rnium ratio ) = Cob are / Co Cd covere d == 2 . 7 0

• R( . d .

• Constants used in the calculations are summarized in Table 5 where the effective cross sections, crR, are of prime importance. The neutron flux and reaction cross section are defined in terms of the "fast flux" or neutron flux above 1.0 MeV as 0:,

J O"(E)N(E)dE

~ J cr(E)N(E)dE = Qf 0 ro (1) 0 J

1.0Mev N(E)dE

17 TABLE 3. FAST NEUTRON DOSIMETRY RESULTS (E > 1 MeV) FOR SURRY UNIT NO,. 1 V

2 Location Fast Fluence. nLcm

• E ~ l M~V in Capsule Iron Nickel Copper Uranium Nepiunium 18 18 Top 2.87xlo 3.36xl0 18 18 18 18 Middle 2.40xlo 2.64xl0 2.99xl0 2 .80xI0 18 18 Bottom 2. 23xl0 3.43xl0 18 18 18 18 18 Average 2 .50xl0 2.64xl0 3.40xl0 2.99x10 2. 80x10 TABLE 4. THERMAL NEUTRON DOSIMETRY RESULTS FOR SURRY UNIT NO. 1 2

Location Thermal Fluence, n/cm in Capsule Co Coed-Covered Bare 18 18 Top 4.35xl0 1. 64xlo Middle 18 18

3. 77xlo l.35x10 Bottom l~. 08xlo 18 1. 53x10 18 Average 18 18 4.07xl0 l.51xl0 True thermal fluence(l) 18 2.56xlo (1) True thermal fluence = Co x .E.-=---1R~

Bare Co Bare  !+/-..sn_

where R (Cadmi~ Ratio)=~~~~ - - 2 70 Coed.

a mium ov. C - 1.51 - *

. 18 TABLE 5, VALUES USED IN DOSIMETRY CALCULATIONS Target Threshold Fission Product (1)

Isotope crR Energy Yield Half-Reaction Target  % Abundance barns' MeV  % Life 63 60 0.000631 ,

. Cu(n,O:') Co 100% Cu 69.17 5 5.26y 58N.J. ( n,p ) 58C o 100% Ni 67. 77 0.111 1.0 71.3d 54 5 Fe (n, p) \m 96.7% Iron 5,82 0.0832 1.5 314d

-?3A~U(n,f) 137***

• Cs U308. - >99. 9 0.356 0.8 6.2. 30.0y 237 Np (n, f) 137 Cs Np0 2 >99.9 2.55 0.4 6.0 30,0y 59 60 co(n,Y) co Al-0.15% Co 100 37.1 5,26y (1) For fast reactions,  % is for neutrons >1 MeV at capsule location.

19 where g 2

= Neutron flux at full reactor power, n/cm -sec co Q

f = Q J N(E)dE = Flux above 1.0 MeV 1.0 MeV CYR = Average reaction cross section above 1.0 MeV

. . 2 a (E) = Reaction cross section, cm, at energy E.

Calculations have been perfonued to determine the reaction. cross sections, crR, and the activation correction factors, C, for five dosimeters placed in the Surry Unit No. 1 reactor. The ANISN computer program was used to calculate the neutron energy spectrum at the dosimeter positions and to perform the integrations indicated in Equation (1). The dosimeter activation at the time of removal from the reacto.r is then A = n Qf crR C ,

where A = *

• Disintegrations* / cm3/ sec 3

n = Dosimeter nuclei concentrations, atoms/cm.

J

..-, -A T * -A (T - t . )

C = l F.J (1 - e. J) e J = Activation correction factor j=l

t. = The elapsed time at the end of the jth time interval, sec J

J.= Number of time intervals F. = Fractional power level during a time interval T.

J -1 J A= Dosimeter decay constant, sec T = Length of time the dosimeter is in the reactor, sec.

Figure 4 is a comparison of the ANISN-calculated neutron spectr.u1n with the fission spectrum, It' is seen that the. ANISN spectrum contains far fewer high-energy neutrons than the fission spectrum. This causes the ANISN-calculated values of CYR to be considerably smaller than the fission-spectrum-averaged cross sections. The activation correction factors were calculated

20 0.4 n2 r-4 I

~ ANISN Spectrum

~

r--

r::.1 0.1 z

. 0.08 5 . L_

~

.µ tl

(!.)

0.06[ r p,

Cl) bO

~

a.I p

r::.1 0.04f

. I 0,02 0.01.-...*~~~~....il~~......;.I~~*=__;.l~.......~*.._.......i..!~':.....iJ..,..:i..--~~~__..~.........__,..~~----.&-,!._....__...._.

0.1 0.2 0.4 0.6 0.8 1.0 2.0 4.0 6.0 8.0 Neutron Energy, MeV FIGURE 4. COMPARISON OF THE NEUTRON FISSION SPECTRUM WITH THE ANISN-CALCUL.i\TED SPECIRUN AT THE DOSD1ETER LOCATION FOR SURRY UNIT NO. 1

21 for Surry Unit No. l operation through December 31, 1974. The power history.

237 238 .

was taken from Reference (19). The decay for the Np and U dosimeters was calculated for the fission product Cs-137.

Tensile Properties The tensile properties are listed in Table 6 for the base and weld metals. The table lists temperature, 0.2 percent offset yield strength, ultimate tensile strength, uniform elongation, total elongation, and reduction in area. Posttest photographs at 4X of, the tensile specimens are shmm in Figures 5 through 8. These photographs show the necked down region of the gage length and the fracture. A typical tensile curve showing stress as a function of strain is shown in Figure 9; the particular test shown is for base*metal specimen VlO tested at 88 F.

Tensile tests for the irradiated metals were run at room temperature and 550 F. i'he results are shown comparecl Lo the unirradiated propc:rtie:;

from Reference (9) in Figures 10 and 11, which are plots of ultimate tensile strength, 0.2 percent offset yield strength, reduction in area, and total elongation. The yield strengths and ultimate tensile strengths have increased appreciably as a result of irradiation at 88 and 550 F for both the base metal and the weld metal. The total elongation has decreased slightly at 88 and 550 F for both metals. The reduction in area has decreased at both temperatures for the two metals, with the weld metal showing greater decreases then the base metal. The increases in yield strength and ultimate tensile strength accompanied by decreases in reduction in area and total elongation as a result of irradiation represent typical behavior for these base and weld metals.

Charpy Impact Properties The impact properties determined as a function of temperature are listed in Table 7 through 10. In addition to the impact energy values, the tables also list the measured values of lateral expansion and the estimated fracture appearance for each .specimen. The lateral expansion is a measure

22 TABLE 6. SURRY UNIT NO. 1 TENSILE PROPERTIES 0.2%

Offset Untimate, Yield Tensile Total Reduction Temp., Strength, Strength, *Elongation, in Area,

• Material Specimen F psi psi percent percent Base VlO 88 81,760 103,890 20.8 67.0 Base V9 550 69,310 95,930 15.8 63.8 Weld W2 88 89,860 105,480 21.6 56.0 Weld Wl 550 77,940 96,150 17 .0 51.2

P5895 FIGURE 5. TENSILE SPECIMEN VlO TESTED AT 88 F P5894 FIGURE 6. TENSILE SPECIMEN V9 TESTED AT 550 F

P5893 FIGURE 7. TENSILE SPECIMEN W2 TESTED AT 88 F P5892 FIGURE 8, - TENSILE SPECIMEN Wl TESTED AT 550 F

25 2001000-*----r----,.-----ii-----r---~-.--...........___,.

l,oo,oo)t U) r L

0 t'---***--"__.....

1 _.________. --~__.,_-~--,__----------~--~--=-i 0 5 10 15 20 25 30

-Percent Elongation, in/in FIGURE 9. TYPICAL STRESS-STRAIN CURVE FOR SURRY UNIT- NO. 1 Curve shown is for base metal tested at 88 F.

/

t f, 26

.120,000 r-----..--------,...f----,-----,----.,....-----...,.1--~

100,000 fsA - Ultimate --~ ~ ~ ~ - - - --£ Ten"Sile Slrength er,

a. ~ ~ I - - - - - * -_ _ _ -fs (f)

Cf)

.W er::

I-80,000 I en '*

60,000

.r--------e b.., 0::. Unitrodiote4 Ai~- lrrcdioted 40,000 I -- I J sor A -r-* I /' r -~"'1--~

I D.1r----~------J4 Reduction in Area; ft 1*.

~ 6T w

u 40 0::

w 0..

Toiol EfongoHon *

&(~:------_J-Bt,__ _ _ ____,..-----0 20 0

I

~

1 0-Unincdioled

&,._/fiP-lH;diotad L.,., ~~-~-L= I ~w* .L c = j l

-=---ao<'600 a~~

C 100 200 300 400 500

• TEMPERATURE, F FIGURE 10. COMPARISON OF UNIRRADIATED AND IRRADIATED TENSILE PROPERTIES FOR SURRY UNIT NO. 1 BASE METAL

. PLATE C4415- l

27 120,000r-----r*----r"----r----r--~~----r------..,....,....~-**--*

I00,000 Ultimate Tensile Strcnglh

(()

0..

0 60,000, ----~6---*--------£

~,Q :Unitrodiai!H:~

A.,f/!J ... lucdioted 80 -~-, J u,*....-..41::a,,,,;,..

~,-

l .

~

60

~

IJ.

---

~"B:---------~-

Reducfion in Arao. .J I-

• Z w

<...)

0:::

40 -

w

. CL 5---.

• Total Elongolico 20 ~

--~-~--=--=--=-~-8':::!_--.:_-=-=-=-=-=-=-=-=-=-=--=--=;. -a

.6.,0-Unirrodi(J!ed

/J~,G)- ltrodiot~d *

o. ~_,..,_______,_t~~------~~......,_,~~~__,,,~~----1........__,.

0 JOO 200 300 400 500 600 TEMPERATURE, f FIGURE 11. COMPARISON OF UNIRRADIATED AND IRRADIATED TENSILE PROPERTIES FOR SURRY UNIT NO. 1 WELD NETAL

,.J

28 TABLE 7. CHARPY V-NOTCH IMPACT TEST RESULTS FOR BASE NETAL PLATE C4415-l Test Impact Lateral Fracture Temperature, Energy, Expansion, Appearance, Specimen F ft lb mils Percent Shear V46 0 13.0 13 .o 5 V44 35 31.5 27.5 15 V47 77 li-8. 0 41.5 30 V41 120 64.0 52.5 65 V43 150 81.5 63.5 55 V42 212 106 . 5 82.0 100 V48 295 125.0 85.5 100 V45 350 115.0 68.0 100 TABLE 8. CRA.RPY V-NOTCH IMPACT TEST RESULTS FOR WELD HETAL Test Impact Lateral Fracture Temperature, Energy, Expansion, Appearance, Specimen F ft lb mils Perc.ent Shear W3 0 8.5 13.5 0 W4 77 20.0 21.0 10 W6 120 23.5 26.0 30 W8 212 41.0 /14.0 90 W2. 295 49.0 52.5 100 W7 350 53.0 45.5 100 HS 350 54.0 55.5 100 Wl 390 51.0 53.0 100

29 TABLE 9*. CHARPY V-NOTCH Il1PACT TEST RESULTS FOR HEAT AFFECTED ZONE METAL

• Test Impact Lateral Fracture Temperature, Energy, Expansion, Appearance, Specimen F ft-lb mils* Percent Shear Hl} 0 21.5 22.5 10 Hl 35 16.0 17.5 10 H3 -35 80.0 56.0 '90 HS 77 83 .o 56.5 80 HS 120 39.0 39.0 60 H2 212 113 .5 81.-0 100 H7 295 131.0 86.5 100 H6 350 52.0 55.5 100 TABLE JD. CHARPY V- NOTCH IHPACT TEST RESULTS FOR ASTM CORRELATION MONITOR METAL Test Impact Lateral Fracture Temperature, Energy,

• Expansion, Appearance, Specimen F *ft-lb mils Percent Shear R33 15 4.0 7.0 0 R40 77 13.0 17.0 20 R38 120 33.5 37.0 25 R36 150 50.0 49.0 45 R39 212 68.0 62.0 70-.

R37 295 107.0 89.0 100 R34 350 111~.s . 71.0 100 R35 390 98.0 74.0 1.00

, 30 of the deformation produced by the striking edge of the impact machine hammer "lvhen it impacts the specimen; it is the change in specimen thickness of the section directly adjacent to the notch location. The fracture appearance is a visual estimate of the amount of shear or ductile type of fracture appearing on the specimen fracture surface.

The impact data listed in Tables 7 through 10 are graphically shown in Figures 12 through 15. These figures show the change in impact properties as a function of temperature. The data for the irradiated curves *was detennined 9

in the present program, and the unirradiated curves are from WCAP 7723< ). Of particula::: interest are the temperatures corresponding to the impact energies of 30 and 50 ft-lb. The energy level of the upper shelf is also of interest.(~£ the upper shelf is relatively low (e.g., 50 ft-lb or lower), the possibility of failure by low energy ductile tearing is greater*)..__

The curves for the four irradiated metals are well defined with little data scatter with onl:r one exception, The exception is the curve for the irradiated HAZ metal, The reason for this is that it is difficult to cut specimens out of a heat affected zone in a plate between base metal and weld metal, and be assured that the HAZ specimens all have the identical microstructure and thermal history, The unirradiated impact data for the HAZ metal also showed substantial data scatter, and the actual data points from WCAP 77 23 for the unirradiated 1-L.<\.Z metal are shown with the data points for the irradiated RAZ metal. A dashed line is drawn for the irradiated F.AZ metal. It is not necessarily the best fit to the data,. but rather is a reference curve to which the actual irradiated data points can be compared.

Figures* 16 through 19 show the fracture surfaces of the Charpy specimens.

Figure 16, as an example, shows how* the fracture surface changes as the test temperature is increased for base metal spec~mens. The OF- specimen (V46) shows an almost flat fracture surface, with only 5 percent shear fracture appearance,

.This specimen absorbed 13.0 ft-lb of energy during the impact test, a typically low value for the low temperature, brittle region of the Charpy curve. As can be seen in the figure, the amount of lateral expansion is small, and was measured as being 13.0 mils. As the test temperature is increased, specimens show an

. 31 l~o v -~-:---* r *= r-----i~-----r-----

- . Unirradiotad

-o- lrrodialed 0

---

0 1

• -100 . 0 100 200 300 400 TEMPERATURE, F FIGURE. 12. CHARPY UIPACT .ENERGY VERSUS TEHPERATURE FOR SURRY UNIT NO. 1 BASE METAL PLATE C4415-l .

32 100:-----r-----.----i-------.-------i-------

Unirrodiated lrrndia.ted

.* 80

., -o -

{Db cc w

z I

w 20 o--__,____,_____..,__~_._____.__________

-200 * * -100 0 100 200 , 300 400 TEMPERATURE, F

• FIGURE 13. CRI\.RPY IMPACT .fillERGY VERSUS TEMPERATURE FOR SURRY UNIT NO. 1 WELD :METAL

33 150i----r-----r-------.----~----,------

--1:r- Unirrndioted lrrnciiotad

• 0 0

...alOO I .

/~

& // . fl

/

. Unirrodiar~d- /

.I 0

/-. lrractiohHr I

6. .I 0 I

I I

0 I

/o O*--~~~...--i..~~-~--1.~~~--!.~~~--'~~----~-'--~~~~

-*200 -100 **0 100

• 200 300 400 TEMPERATURE 1 F FIGURE lli., CHAR.PY IMPACT ENERGY VERSUS TfiliPERATURE FOR SURRY UNIT NO. 1 HR.-4.T AFFECTED ZONE :METAL

34 150-r-----.------.-----.-----.-------r-------.

- Unirrodioted Irradiated 0

~100 I_

~2,,_0_0____~10_0_ ____,o'-----,n--L-Jo___2_._0_0___-::_*,o---Jo_ _ _ 400 TEMPERATURE) F FIGURE 15. CHARPY IMPACT ENERGY VERSUS TEMPERATURE FOR SURRY UNIT NO. 1

• ASTM CORREL..\TION HONI TOR liETAL

V46 V44 V47 V41 V43 V42 V48 V45 FIGURE 16. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR SURRY UNIT NO. 1 BASE METAL W3 W4 W6 W8 W2 W7 W5 Wl FIGURE 17. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR SURRY UNIT NO. 1 SAW WELD METAL

H4 Hl H3 H5 H8 H2 H7 H6 FIGURE 18. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR SURRY UNIT NO. 1 HEAT AFFECTED ZONE METAL R33 R40 R38 R36 R39 R37 R34 R35 FIGURE 19. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR SURRY UNIT NO. 1 ASTM CORRELATION MONITOR METAL

,l 37 increasing amount of shear fracture appearance, The +350 F specimen (V45) fracture surface is typical of the type seen at the higher temperature end of the Charpy transition curve, The fracture surface shows large *shear lips with a 100 percent shear fracture appearance. The specimen absorbed the relatively large amount of 115.0 ft-lb during impact, The substantial amount of plastic deformation occurring during this test is reflected in the large value of 68,0 mils lateral expansion.

Table 11 summarizes the unirradiated and irradiated 30 and 50 ft-lb transition temperatures and the upper shelf energy levels for the four metals, Table 12 lists the 30 ft-lb and 50 ft-lb transition teriiperature shifts due to irradiation. The 30 and*50 ft-lb transition temperatures are difficult to determine for the HA.Z metal due to the scatter in the data.

The ASTM correlation monitor metal shows a 30 ft-lb shift of 30 F and a 50 ft-lb shift of 80 ft-lb, Irradiation has caused the upper shelf energy to drop from 123 to 105 ft-lb.

For the three reactor vessel metals, the irradiated 30 ft-lb transition temperature ranges from 40 F (base) to 150 F (weld). The FAZ value is difficult to define due to data scatter, but it is below the 150 F weld metal value. The largest shift due to irradiation is an increase from -15 F to 150 F, a total of 165 F, for the weld metal. The irradiated 50 ft-lb transition temperature ranges from 80 F (base) to 300 F (weld). The 50 ft-lb HAZ value is difficult to define due to data scatter, but it is below the 300 F weld metal value. The largest shift due to irradiation of.the 50 ft-lb transition temperature is an increase from 50 F to 300 F, a total of 250 F, for the weld metal.

The upper shelf energy levels for the base and weld metals have been lowered by irradiation. The base metal upper shelf dropped slightly from 125 to 120 F, a change of 5 F. The weld metal upper shelf dropped from 70 to 53 F, a change of 17 F, The upper shelf energy level change for the RAZ metal is difficult to determine due to the large amount of data scatter; however, the three HAZ specimens showing 100 percent shear fracture appearance have an average impact energy of 99 ft-lb.

I I

I

38 TABLE 11. CH..lill.PY INPACT PROPERTIES FOR SURRY UNIT NO. 1 Tx:ans;ition Temperature. F Upper Shelf, Material Condition 30 ft-lb 50 ft-lb ft-lb Base Unirradiated -10 20 125 Base Irradiated 40 80 120

• lLA.Z Unirradiated rv-50 ""'-15 (a)

RAZ Irradiated (a) (a) (a)

Weld Unirradiated -15 50 70 Weld Irradiated 150 300 53 ASTM Correlation Unirradiated 45 75 123 Monitor ASTI1 Correlation . Irradiated 115 155 105 Monitor (a) See text for discussion of HAZ metal.

TABLE 12. C0r1PARIS0N OF SURRY UNIT NO. 1 30 FT-LB AND 50 FT-LB TR.l>,.NSITION Tfil1PERATURE SHIFTS 30 ft-lb Transition 50 ft-lb Transition Material Temperature Shift, F Temperature Shift, F Base 50 60 RAZ (a) (a)

Weld 165 250 ASTH Correlation 70 80 Monitor (a) See text for discussion of HAZ metal;

39 The significant drop in upper shelf energy level for the weld metal from 70 ft-lb to the relatively low value of 53 ft-lb represents a decrease of 17.ft-lb or 24 percent. Drops of this magnitude have been (22) associated with high residual element levels

• The chemical compositions of the base metal and weld metal used for the surveillance capsule specimens is presented in Appendix A of this report. The copper content is quite high, having a value of 0.25 weight percent. It is this relatively high copper level which is most pro.bably causing the low irradiated upper shelf level of the weld metal.

The transition temperature shifts .for the* three reactor vessel metals are shown plotted in Figures 20 and 21 as a function of fluence.

The first figure is the 30 ft-lb transition temperature shift, and the second figure is the 50 ft-lb transition temperature shift values obtained in the present survei* 11 ance program an d in

• other survei' llance programs. <23 3

- o)

R4.Z values fi:*om the present program are not included. The values used ta form the tread band are those from programs where the irradiation temperature was between 550 and 590 F. The apparent large scatter in data among the various progr~ns is not unusual. Note that the weld metal values generally determine the upper bound of the trend.band. It can be seen that the tran-sition temperature shift values for the base metal of the present program falls within the upper and lower bounds determined by metals of the other programs, but the weld metal does noL This is due to the relatively low upper shelf of the weld metal, which leads to large transition temperature shift values.

As mentioned in the Introduction of the present report, the surveillance capsule being examined contained not only Charpy specimens machined from pressure vessel metal used in the reactor under examination, but also ASTM correlation monitor specimens. These specimens are in numerous conunercial power reactor surveillance capsules. By comparing the results of the correlation monitor specimens from numerous surveillance.programs, further knowledge should be gained concerning the effect of differing nuclear irradiation conditions (neutron spectrum and flux intensity) on the radiation response of reference correlation monitor steels(S).

40 I I 11 T Base Weld HAZ Point Beach No. I <> ~ .

~

Connecticut Yankee 0 @ e 300 Big Rock Point V wt ~

Yankee Surveillance 6 Yankee Special 0 Hum bolt Bay [> I}?., . ~

Son Onofre <] ~ 4 Surry Unit No. ! G l§.,

ti.. 1 w 200 Elk River . X V>

a Q) t-u c.:::

j,--f t ~

(l.)

-~

l a

l-Cl.)

0...

E

~

I-

• o z

100

• Trend Bond For 550- 590 F I I I *1 I ___ j~*----'-'-.!--.J.......!----!-...1....l.-J---"""

1019 ,020 Neutrnn Fluence 1 nvt FIGURE 20. COMPARISON OF 30 FT-LB TRANSITION TEMPERATURE SHIFT VALUES FROM VARIOUS SURVEILLANCE PROGRAMS FOR A302 GRADE BAND A533 GRADE B PRESSURE VESSEL STEELS *I

41 Base Weld HAZ Point Baach No. I <> + ~.

300-r Connecticut Yankee 0 e G 1-1. 8. Robinson D Surry Unit No. I D LL Cl)

-5 200 0

l-o.>

n.

E C) l-e:::

0

• I--

v, C

Ci F

...a

~

0

. G>O 0 <9a

~

D I

l D D

,..,.,L_J_L ~J !. I 1619 L t I 111l _

1020

_J Neuhon Fluence nvt I FIGURE. 21. COMPARISON OF 50 FT-LB TRANSITION TEMPERATURE VALUES FROM VARIOUS SURVEILLANCE PROGRAMS FOR A302 GRADE BAND A533 GRADE B PRESSURE VESSEL STEELS

42

<> ~ Point Beach No. I o - , Connecticut Yankee 300 \I - Big Rock Point D - Yankee

<J - San Onofre

~ - Surry Unit No. I LL.... 200 (l.) .

(f)

I 0

C,)

~

u c::

--

.

Cl) I

• • -E

=:}.

(t) 0..

E*

-~

i--a 0

z -100

"'- Trend Band For 550- 590 F 1019 1020 Neutron Fluence, nvt FIGURE 22. COMPARISON OF 30 FT-LB TRANSITION TE.MPERATURE SHIFT VALUES FROM VARIOUS SURVEILLANCE PROGRANS FOR A302 GRADE B AND A533 GRADE B.

ASTM CORREIATION-:MONITOR :MATERIAL

,l.

43 The 30 ft-lb transition temperature value for the ASTM correlation monitor material is shown in Figure 22. Alt,;o shown in this figure are the 30 ft-lb transition temperature values obtained from other programs using

• this metal. The 30 ft-lb transition temperature value determined in the present program is consistent with other values*shown.

The Figure 20 30 ft-lb transition temperature trend band can be used to estimate the 30 ft-lb transition temperature values of the pressure vessel metals after various periods of reactor operation. According to Reference (9), the capsule was at a core position of.285 degrees in the reactor. The exposure lead factor of the surveillance capsule in this position (20) with respect to the pressure vessel wall is 1.52. Because the capsule received an exposure of 2.50 X 1018 nvt (> 1 MeV), the vessel wall therefore, 18 received an exposure of 1. 65 X 10 nvt (>l MeV). As discussed earlier in the report, the reactor had been run for 1.07 equivalent full power years (EFPY) at time of capsule removal. The 10, 20, and 32 EFPY pressure vessel 19 19 exposures woul<l therefore be predicted to be 1.54:;;: 10 , 3.08 x io , and 19

4. 93 x 10 nvt (>1 MeV). By observing where these values are on Figure _20, the trend band can be used to predict behavior of pressure vessel metals at these exposures. However, this does not apply to the weld metal because of its relatively low upper shelf energy and resultant relatively high 30 ft-lb transition temperature.

CONCLUSIONS*

',Che maximum irradiation temperature of the irradiation surveillance capsule did not exceed 590 F. The .neutron fluence experienced by the capsule 18 was 2.5 x 10 nvt (>l MeV) which was attained after 1.07 equivalent full-power years of operation, Based on a capsule lead factor of 1.52, this is equivalent to a maximum fluence of 1.65 x 10

~

18 _____________________

nvt (>l MeV) for the pressure vessel after 1.07 equivalent full power years. For a pressure vessel life of 40 years operation at an 80 percent load factor (32 equivalent full power years), the maximum fluence experienced by t;he pressure vessel would there-19 fore be predicted to be 4.93 x 10 nvt (>l MeV).

44 Tensile tests were conducted at 88 F and 550 Fon base and weld metal. The yield strength and ultimate tensile strength of both metals increased due to irradiation. The corresponding ductilities (reduction in area and total elongation) decreased in both metals at both temperatures.

The Charpy impact behavior was determined for base, weld, and RAZ metal. The lowest upper shelf observed was for the weld metal, which decreased from 70 to 53 ft-lb due to irradiation. The highest 30 ft-lb and 50 ft-lb irradiated transition temperatures observed were for the weld metal, which yielded values of 150 F and 300 F, respectively

• 45 REFERENCES (1) Reuther, T. C., and Zwilsky, K. M., 11 The Effects of Neutron Irradiation on the Toughness and Ductility of Steels 11 , in )?roceedings of Toward

.Improved Ductility and Toughness Symposium, published by Iron and Steel Institute of Japan (October, 1971), pp 239-319.

(2) Steele, L. E., 11 Major Factors Affecting Neutron Irradiation Ei-nbrittle-ment of Pressure-Vessel Steels and Weldments 11 , NRL Report 7176 (October 30, 1970).

- . \~**

(3) Berggren, R. G., 11 Critical Factors in the Interpretation of Radiation Effects on the Mechanical Properties of Structural Metals 11 , Welding Research Council Bulletin, frl.., 1 (1963).

(4) Witt, F. J., 11 Heavy-Section Steel Technology J?rogram Semiannual Progress Report for Period Ending February 29, 1972 11 , OR.l'i"L Report No. 4816 (October, 197 2).

(5) Hawthorne, J. R., 11 Radiation Effects Information Generated on the ASTM Reference Correlation-Monitor Steels 11 , American Society for Testirig ar..d Materials Data Series Publice.tion DS54 (1974).

(6) Steele, L. E., and Serpan, C. Z., 11 Neutron Embrittlement of- Pressure Vessel Steels - A Brief Review11 , Analysis of Reactor Vessel Radiation Effects Surveillance Programs, American Society for Testing and Mater:ials Special Technical Publication 481 (1969), pp 47-102.

(7) Integrity of Reactor Vessels for Light~Water Power Reactors, Report by the USAEC Advisory Committee on Reactor Safeguards (January, 1974).

(8) ASTM Designation E185-73, 11 Surveillance Tests on Structural Materials in Nuclear Reactors 11 , Book of ASTM Standards, Part 45 (1974), pp 621-627, (9) Yanichko, S. E., 11 Virginia Electric and Power Co. Surry Unit No. 1 Reactor Vessel Radiation Surveillance Program11 , WCAP 7723, Westinghouse Electric Corporation (July, 1972).

(10) ASTM Designation E320-69T, 11 Ra.diochemicc!-l Determination of Cesium-137 in Nuclear Fuel S0lutions 11 , Book of ASTM Standards, Part 30 (1970),

pp 1004-1009.

(11) ASTM Designation E261-70, 11 Measuring Neutron Flux by Radioactivation Techniques!!, Book of ASTI1 Standards, Part 30 (1970), pp 762-772.

(12) ASTM Designation E262-70, 11 Measuring Thermal-Neutron Flux by Radioactivation Techniques 11 , Book of ASTM Standards, Part 30 (1970), pp 773-780.

1*

46 (13) AS' Designation E263-70, "Measuring Fast-Neutron Flux by Radioactivation of Iron, Book of AS11'I Standards, Part 30 (1970), pp 781-783.

(14) ASTM Designation E264-70, "Measuring Fast-Neutron Flux by Radioactivation of Nickel", Book of ASTM Standards, Part 30 (1970), pp 787-791.

(15) ASTM Designation E343-67T, "Fast-Neutron Flux by Activation of Molybdenum-99 Activity from Urantum-238 Fission", Book of ASTI-'1 Standards, Part 30 (1970),

pp 1078-1084. .

(16) AS'IM Designation E393-69T, "Measuring Fast-Neutron Flux for Analysis for Barium-lL~O Produced by Uranium-238 Fission", Book of ASTM Standards, Part 30 (1970), pp 1174-1180.

(17) ASTM Designation E23-72, "Notched- Bar Impact Testing of Metallic Materials,

Book of AS' Standards, Part.IO (1974), pp 167;-183.

(18) ASTM Designation A370-71, "Mechanical Testing of Steel Products'! Book of ASTM Standards, Part 10 (197L~), pp 1-52.

(19) Private communication fran J. T. Benton of VEPCO to J. S. Perrin of BCL (January 31, 1975).

(20) Privai:e cormnunic:ation from J. T. Benton of VEPCO to J. S. Perrin of BCL (February 21, 1975).

(21) Private corrununication from J. Buhl of VEPCO to J. S. Perrin of BCL (June 24, 1975).

(22) Bush, S. H., "Structural Materials for Nuclear Power Plants", AS' Gillette Memorial Lecture (J.974).

(23) Serpan, C. Z., Jr., and Watson, H. E., "Mechanical Property and Neutron Spectral Analyses of the Big Rock Point Reactor Pressure Vessel", Nucl.

Eng. Design, ll, 393-415 (1970).

(24) Serpan, C. Z., Jr., and Hawthorne, J. R., "Yankee Reactor Pressure-Vessel Surveillance: Notch Ductility Performance of Vessel Steel and Maxinmm Service Fluence Determined from Exposure During Cores II, III, and IV, NRL Report 6616 (September 29, 1967).

(25) Brandt, F. A., "Humboldt Bay Power Plant Unit No. 3 Reactor Vessel Steel Surveillance Program", GECR-5492 (May, 1967).

(26) "Analysis of First Surveillance Material Capsule from San Onofre Unit I", Southern California Edison Company (July, 1971).

(27) Perrin, J. S,, Sheckherd, J. W., and Scotti, V. G') Examination and i Evaluation of Capsule F for the Connecticut Yankee Reactor Pressure-Vessel Surveillance Program", Final Report to-Connecticut Yankee Atomic Power Company (March 30, 1972).

!~-~----------------------------------------

47 (28) Perrin, J. S., Sheckherd, J. W., Farmelo, D.R., and Lowry, L. M.,

11 Point Be.ach Nuclear Plant Unit No. 1 Pressure Vessel Surveillance Program: Evaluation of Capsule V11 , Final Report to Wisconsin Electric Power Company (June 15, 1973).

(29.)

• Ireland, D. R., and Norris, E. B., 11 Influence of Neutron Irradiation on the Properties of Steels and Weld Typica1 o"f the ERR Pressure Vessel After Two Power Years Operation11 , SWRI-1228-P-9-15 (March, 1968).

(30) Sterne, R. H., Jr,, and Steele, L. E., 11 Steels for Cormnercial Nuclear Power Reactor Pressure Vesselsr:, Nucl. Eng. Design, 10, 259-307 (1969).

APPENDIX f.

PRESSURE VESSEL MATERIAL

• APPENDIX A*

PRESSURE VESSEL MATERIAL Babcock and Wilcox Co. supplied material from two 7-7/8-inch-thick plates (Heat C4326-1 and C4415-l) of SA 533 Grade B Class 1 steel used in the Surry Unit No, 1 reactor pressure vessel intermediate and lower shell course, respectively. In addition, a weldrnent which joined sections of the plates and represents the vessel core region girth weld was also supplied.

The two plates were produced by Lukens Steel Co. The chemical composition of the two plates and the weldment are shown in Table A-1. The heat treat-ment history of the above material follo~s:

Intermediate shell 1650°-1700° F - 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Water-quenched (Plate C4326- l) 1210° F - 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Air-cooled 1125° F 1/2 hours Furnace-cooled to 600 F Lower shell 1650-1700° F - 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Water quenched (Plate C4l~l5- l) 1200° F 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Air-Cooled 1125° F 1/2 hours Furnace-cooled to 600 }'

Weldment 1125° F 1/2 hours Furnace-cooled to 600 F

• The information in this Appendix is from Yanichko, S. E.,

"Virginia Electric and Power Co, Surry Unit No, 1 Reactor Vessel Radiation Surveillance Program", WCAP 7723, Westinghouse Electric Corporation (July, 1972).

TABLE A-1. CHEMICAL COMPOSITION OF SHELL PLATES Af..1D WELD HETAL FOR SURRY. UNIT NO. 1 Intermediate Lower Shell Plate Shell Plate Element C4326-l C4415-l

• Weld Metal C 0.23 0.22 0.10 Mn 1.35 1.33 1.49 p 0.008 0.014 . 0. 011 s 0.015 0.014 0.010 Si 0.23 0.23 0.37 Ni 0.55 0.50 0.68 Cr 0.069 0.078 0.076 V 0.001* 0.001* 0.001 Mo 0.55 0.55 0.46 Co 0.014 0.015 0.001 Cu 0.11 0.11 0.25 Sn 0.008 0,008 - -

Zn 0.001* 0.001*

Al 0.037 0.036 0.013

• N2 0.007 0.007 0.008 Ti 0.001* 0.001*

Zr 0.002 0. 002 .

As 0.007 0.007 B o. 0031: 0.003*

• Not detected, The number indicates the minimum limit of detection.

NOTE: On qualitative spectrographic analyses without standards for impurities, no other elements over 0.01% were found.

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APPENDIX B ASTM CORRELATION MONITOR MATERIAL

APP~DIX B>',

ASTM CORRELATION MONITOR MATERIAL Correlation monitor material for Surry Unit No. 1 was supplied by the Oak Ridge National Laboratory from plate material used in the AEC-sponsored Heavy Section Steel Technology (HSST) Program. This material was obtained from a 12-inch-thick A533 Grade B, Class 1 plate (HSST Plate 02) which was provided to Subcommittee II of ASTM Committee E 10 on Radioisotopes and Radiation Effects to serve as correlation monitor material in reactor vessel surveillance programs. The plate was produced by the Lukens Steel Co.

and heat-treated by Combustion Engineering, Inc. The following is a tabulation of the heat treatment history and plate chemistry:

Heat Treatment History 1675 +/-25° F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Air Cooled 1600 +/-25° F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Water- qu'?.nched 1225 +/-25° F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Furnace-cooled 1150 +/-25° F 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> - Furnace-cooled to 600° F Plate Chemistry C Mn p s Si Ni Mo Cu Ladle 0.22 1.45 0.011 0.019 0.22 0.62 0.53 Check 0.22 1.48 0.012 0.018 0.25 0.68 0.52 0.14

• The information in this Appendix is from Yanichko, S. E., "Virginia Electric and Power Co. Surry Unit No. 1 Reactor Vessel Radiation Surveillance Program", WCAP 7723, Westinghouse Electric Corporation (July, 1972).