Text

Analysis of Capsule T from Pse&G Salem Unit 2 Reactor Vessel Radiation Surveillance Program
	ML18092A231
Person / Time
Site:	Salem
Issue date:	03/31/1984
From:	Baggs R, Cheney C, Yanichko S; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML18092A230	List: ML18092A229; ML18092A231; ML18092A234;
References
	WCAP-10492, NUDOCS 8407110271
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{{#Wiki_filter:! ( I WCAP-10492 . WESTINGHOUSE CL.a 3 CUSTOMER DESIGNATED DISTRIBUTION ANALYSIS OF CAPSULE T FROM THE PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM R.S. Boggs S. E. Yanichko C. A. Cheney W. T. Kaiser March 1984 Work perf)rmed under Shop Order No. PUOP-106 APPROVcD: ~;f' )"r'I _ _) T. R. Mager, M ager Metallurgical and NOE Analysis Prepared.by Westinghouse for the Public Service Electric and Gas Company Although _information con.ained in this report is nonproprietary, no distribution shall be mad.~ outside Westinghouse or its licensees 9401110211 a*4o7o3 PDR ADOCK 05000311 p PDR without the customer's approval.* WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems 1 P.O. Box 355 I Pittsburgh, Pennsylvania 15230

I I I PREFACE This report has been technically reviewed and verified. Reviewer Sectic1ns 1 thru 5 and 7 M. K. Kunka Sectk,n 6 S. L. Anderson Appe1.dix A F. J. Witt

=

TABLE OF CONTENTS Section Title 1

SUMMARY

OF RESULTS 2 INTRODUCTION 3 BACKGROUND 4 DESCRIPTION OF PROGRAM' 5 TESTING OF SPECIMENS FROM CAPSULE T 6 7 8 5-1. Overview 5-2. Charpy V-Notch Impact Test Results 5-3. Tension Test Results

54.

Wedge Opening Loading Test Results RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1.

Introduction s.:2. Discrete Ordinates Analysis

6-3.

Neutron Dosimetry 6-4. Transport Analysis Results

6-5.

Dosimetry Results SURVEILLANCE CAPSULE REMOVAL SCHEDULE REFERENCES .. Page 1-1 2-1 3-1 4-t 5-1 5-1 5-3 5-4* 5-4 6-1' 6-1 6-1 6-6 6-11 5;..18 ' 7;..1 8-1 Appendix SALEM UNIT 2 HEATUP AND COOLDOWN LIMIT CURVES A-1 A FOR NORMAL OPERATION A-1. Introduction A-1 A-2. Fracture Toughness Properties A-2 A-3. Criteria for Allowable Pressure-Temperature A-2 Relationships A-4. Heatup and Cooldown Limit Curves A-5

r Table 4-1 5-'1 5-2;;...* 5-3 5-4 5-5. s-s: 5-7 5-8 5-9 5-10 6-1 . 6-2.. 6-3

-----~----*=======:::--:---=::=====

LIST OF TABLES Title Page Chemical Composition and Heat Treatment of The 4-3 Salem Unit 2 Reactor Vessel Surveillance Materials Charpy V-Notch Impact Data for The Salem Unit 2 5-5 Intermediate Shell Plate 8. 4712-2 (transverse) Irradiated at 550° F, Fluence 2.56 x 1018 n/cm 2 (E > 1 MeV)

Charpy V-Notch Impact Data for The Salem Unit 2 5~6 Intermediate Shell Plate B 4712-2 (lon~itudinal) :

Irradiated at 550°F, Fluence 2.56 x 101 (E > 1 MeV) Charpy V-Notch Impact Data for The Salem Unit 2 5-7 Pressure Vessel Weld Metal tlrradiated at 550° F, Fluence 2.56 x 1018 n/cm2 (E> 1 MeV) Charpy V-Notch Impact Data for The Salem Unit 2 5-8 Pressure Vessel Weld Heat Affected Zone Metal Irradiated at 550° F, Fluence 2.56 x 1018 (E > 1 MeV) Instrumented Charpy Impact Test Results for 5-9 Salem Unit 2 Intermediate Shell Plate 8 4712*2 (transverse) Instrumented Charpy Impact Test Results for The 5.-10 Salem Unit 2 Intermediate Shell Plate B 4712-2 (longitudinal) Instrumented Charpy Impact Test Results for 5-11 Salem Unit 2. Weld Metal Instrumented Charpy Impact Test Results for 5-12 Salem Unit 2 Heat Affected Zone Metal The Effect of 550°F Irradiation at 2.56 x 1018 5-13 (E > 1 MeV) on the Notch Toughness Properties* of The Salem Unit 2 Reactor Vessel Materials Tensile Properties for Salem Unit 2 Reactor Vessel 5-14 Material Irradiated to 2.56 x 1018 n/cm 2

47 Group Energy Structure 6-5.

Nuclear Constants for Neutron Flux Monitors 6-7 Contained in The Salem Unit 2 Surveillance Capsules Calculated Fast-Neutron Flux (E > 1.0 MeV) and 6-H~ Lead Factors for Salem Unit 2 Surveillance Capsules ii

~. --*-- _____ :.._ _____________... --- -~-- ~-----*----- *---*----*-*-----..- ----~-~~:;..:.,,.---*-- ..... :.:.~'..,,.::;.:~~-* ~~.. ::.~:.:_--*---* LIST OF TABLES (cont.) Table -*E:;o Title 6-4 6-5 6-7 6-8 6-9 6-10 ' 7-1

A-1

.. Calculated Neutron Energy Spectra at the Center of $alem Unit 2 Surveillance Capsules Spectrum-Averaged Reaction Cross Sections at the Center of Salem Unit 2 Surveillance Capsules Irradiation History of Salem Unit 2 Reactor Vessel Surveillance Capsule T

. Comparison of Measured and Calculated Fast-Neutron Flux Monitor Saturated Activities for Capsule T . Results of Fast:.Neutron Dosimetry for Capsule T Results of Thermal-Neutron Dosimetry for Capsule T Summary of Neutron Dosimetry Results for Capsule T Salem Unit 2 Surveillance Capsule Removal Schedule Salem Unit 2 (PNJ) Reactor Vessel Toughness Data iii Page 6-19 6-20 6-21 6-22 6-24 6-25 6-26 7-1 A-7

-~-~~~-* -**---- LIST OF ILLUSTRATIONS Figure'~;.:::;*~ Title Page* 4-1 Arrangement of Surveillance Capsules in The Salem 4-4 Unit 2 Reactor Vessel (Updated Lead Factors for I The Capsules Shown in Parentheses.) 4;_2 Capsule T Diagram Showing Location of Specimens, Thermal Monitors, and Dosimeters 4-5 5-l Irradiated Charpy V-Notch Impact Properties for 5-15 Salem Unit 2 Reactor Vessel Intermediate Shell Plate B 4712-2 (transverse orientation) 5-2 Irradiated Charpy V-Notch Impact Properties for 5-16 Salem Unit 2 Reactor Vessel Intermediate Shell Plate B 4712-2 (longitudinal.orientation) 5-3 Irradiated Charpy V-Notch Impact Properties for 5-17 Salem Unit 2 Reactor Pressure Vessel Weld Metal. 5-4 Irradiated Charpy V-Notch Impact Properties for 5-18 Salem Unit 2 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5..;5 Charpy Impact Specimen Fracture Surfaces for 5-19 --~- Salem Unit 2 Pressure Vessel Intermediate Shell Plate 8 4712-2 (transverse orientation) 5-6. Charpy Impact Specimen Fracture Surfaces for

5-20 Salem Unit 2 Pressure Vessel Intermediate Shell Plate 8 4712-2 (longitudinal orientation).

5-7 Charpy Impact Specimen Fracture Surfaces for 5-2i Salem Unit 2 Weld Metal 5-8 Charpy Impact Specimen Fracture Surfaces for 5-22 Salem Unit 2 Weld Heat Affected Zone Metal 5-9 Comparison of Actual versus Predicted 30 ft lb 5-23 Transition Temperature Increases for The Salem Unit 2 Reactor Vessel Material based on the Prediction Methods of Regulatory Guide 1.99, Revision 1 5-10::. - Tensile Properties for Salem Unit 2 Reactor Vessel 5..;24 Intermediate Shell Plate B 4712-2 (transverse orientation) 5-11 Tensile Properties for Salem Unit 2 Reactor 5-25 Vessel Weld Metal iv

Fig~re. :--> 5-12 5-13 5-14 6-1 6-2 6-3 6-4 6-5 6-6 6-7 A-1 A-2 A-3 A-4 .. =*-=-=========-===== -~~== LIST OF ILLUSTRATIONS (cont.) Title Fractured Tensile Specimens of Salem Unit 2 Reactor Vessel Intermediate _Shell Plate 84712-2 (transverse orientation) Fracture Tensile Specimens of Salem Unit 2 Reactor Vessel Weld Metal Typical ~tress-Strain Curve for Tension Specimens Salem Unit 2 Reactor Geometry Plan View of a Reactor Vessel Sunieillance Capsule Calculated Azimuthal Distribution of Maximum Fast"."Neutron Flux (E > 1.0 MeV) within the Pressure Vessel Surveillance Capsule Geometry Calculated Radial Distribution of Maximum Fast-Neutron Flux (E > 1.0 MeV) within the Pressure Vessel Relative Axial Variation of Fast-Neutron Flux . (E > 1 ~O MeV) within the Pressure Vessel Calculated Radial Distribution of Fast-Neutron Flux (E > 1.0 MeV) within the Reactor Vessel Surveillance Capsules Calculated Variation of Fast-Neutron Flux Monitor Saturated Activity within Capsules Located at 40 Degrees Calculated Variation of Fast-Neutron-Flux Monitor Saturated Activity within Capsules Located at 4 Degrees Effect of Fluence and Copper and Phosphorus Contents on aRT.NDT for Reactor Vessel Steels Fast Neutron Fluence (E > 1.0 MeV) as a Function of Full Power Service Life Salem Unit 2 Reactor Coolant System Heatup Limitations Applicable Up to 7 EFPY Salem Unit 2 Reactor Coolant System Cooldown Limitations Applic?ble Up to 7 EFPY v Page 5-26 5-27 5-28 6-2 6-4 6-12 6-13 6-14 6-15 6-16 6-17 A-8 A-9*

A-10 A-11

-1 I ~

SECTION 1

SUMMARY

OF RESULTS The anal~sis of the reactor vessel material contained in surveillance Capsule T, the first capsule to be removed from the Salem Unit 2 reactor pressure vessel, led to the fol lowing_ conclusions: _ The capsule received an average fast neutron fluence (E > 1.0 MeV) of 2.56 x 1018 n/cma:. lrr~_d_iation of the reactor vessel intermediate shell plate B 4712-2 to 2.56 x 1018 n/cm2 resulted in 30 and 50 ft lb transition temperature increases of 70 and 65° F, respe~tively, for specir:nens oriented normal to the principal rolling direction of the plate and 50 and 60° F, respectively, for specimens oriented parallel to the plate principal rolling direction-. Weld metal irradiated to 2.56 x.1018 n/cm2 resulted iri both 30 and 50 ft lb transi-tiq_n _1~mperature increases of 155 and 180° F, respective.ly. Co_mparison of the 30 ft lb transition temperatur~ increases for the Public, Sefiitce Electric and Gas Company's* Salem Unit 2 surveillance material with

predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 1 shows that the transition temperature for the plate and weld rnaterial shifted mqr~~than predicted. Since the shifts were greater than predicted and the intermediate and lower shell vertical weld seams chemistries were estimated, the future operating limits for the vessel, shown in Appendix A, are being based on:tJ:~ upper limits of the Regulatory Guide 1.99 prediction curve. --

1-1

}"

~---~--

SECTION 2 INTRODUCTION This repor{presents the results of the examination of Capsule T, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Public Service Electric and Gas Company Salem UnH 2 reactor pressure vessel materials under actual operating conditions. - ~ The. surveillance program for the Public Service Electric and Gas Company Salem Unit 2 r~actor pressure vessel materials was designed and recommended by the Westingh~use Electric Corporation. A description of the surveillance program and the preirraqiation mechanical properties of the reactor vessel materials are presented by Dav!d.~O.n and Smith. [1l The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E-185-73, "Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". [21 . Westing_h_o_use Nuclear Energy Systems personnel were contracted for the prepara-tion of p*rocedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory, where the postirradiation mechanicaJ testing of the Charpy V-notch impact and tensile surveillance specimens was performed. This repo~ summarizes the testing of and the postirradiation data obtained from surveillance Capsule T removed from the Public Service Electric and Gas Company Salem Unit 2 reactor vessel and discusses the analysis of these data. 2-1.

-""...;,;:~ ~~*-... _-*:0. -__... ___ . ---~=-** -- --- -*-* --.....,, __ _ SECTION 3 BACKGROUND* The ability of the large steel pressure vessel containing the reactor core. and* itS primary coolant to resist fracture constitutes an importantfactor in ensuring safety in the. nuclear industry. The beltline region of the reactor pres.sure vessel is the most - critrcal' *region of the vessel because it is s_ubjected to §igQificant fas~ neutron bombardment. The overall effects of fast neutron irradiafion on the rr,echanical properties of low alloy ferritic pressure vessel steels such as.SA 533 GradE B Class 1 (base material of the Public Service Electric and Gas Company Salem Unit 2 reactor pressure-vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility* and toughness under certain conditions of irradiation. A method-for performing analyses to guard against fast fracture in reactor pressure

vessels:has been presented in "Protection Against Nonductile Failure,"Appendix G to Section: Ill of the ASME Boiler and* Pressure Vessel Code. The method utilizes fractur..e mechanics concepts and is based on the reference nil-ductility temperature (RTN[jf}~

RT NOT is defined as the g.reater of either the drop weight nil-ductilHy transition temperature (NOTT per ASTM E-208) or the temperature 60° F less than the 50 ft lb (and* 35,,.mil lateral expansion) temperature as determined from Charpy specimens . oriented normal (transverse) to the major working direction of the material. The RT NOT of a given material is used to index that material to a reference stress intensity factor curve (KIA curve) which appears in Appendix G of the ASr\\/jE Code. The KIA curve is a lower bound of dynamic, crack arrest, and static fracture tough-. ness results obtained from several heats of pressure vessel steel. When a given _material is indexed to the KIA curve, allowable stress intensity factors can be obtained for th is material as a function of temperature. Allowable operating Ii n 1 its can then b.e determined utilizing these allowable stress intensity factors. 3-1

RT NOT and, in turn, the operating limits of nuclear pow.er plants can be adjusted to

account-for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program, l1l in which a surveillance capsule is periodically removed from the operattng nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft lb temperature (A RT NOT) due to irradiation is added to the original RT NOT to adjust t~e RT NOT for radiation embritt!ement. This adjusted RT NOT (RT NOT initial + A RT NOT) is used to index
the materiai to.the K IR curve and, in turn, to set operatingJimits for the nuclear power I

plant which take into account the effects of irradiation on the reactor vessel ma.terials. 1:**... ** 3-2

./ I SECTION 4 DESCRIPTION OF PROGRAM* Eight surveillance capsules for monitoring the effects of neutron exposure on the Salem Unit2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The eight capsules were positioned in the reactor

vessel"oetween the thermal shield and the vessel wall as sh9vm in Figure 4-1. The *
verticarcenter of the capsules is opposite the vertical center of the core.

_Capsule Twas removed from the reactor after 1.16 effective full poweryears*of plant operation*. This capsule contained Charpy V-notch impact specimens from the limiting* core region plate (intermediate shell plate B 4712-2), 1.:ore region weldment and weld heat affected zone (HAZ) material. All heat affected mne specimens were_ obtained from within the HAZ of plate B 4712-2. The capsule also contains wedge opening*:1oading (WOL) and tensile specimens from the weld r.1etal that joined inter-_ mediate sfiell plates B 4712-1 and B 4712-2 and additional te*1sile specimens from intermediate shell plate B 4712-2. The chemistry and heat treatment of the program

surveillance materials are presented in Table 4-l.

Alf test specimens were machined from the 1/4.thickness location of the plate. Test specimens represent material taken at least one plate thickness from the quenched end of tflef plate. Base metal Charpy V-notch and tensile specimens were oriented with the: l'ongitudinal axis of the specimens normal to (transverse orientation) the major worRing direction of the plate. Additional base metal Charpy V-notch speci-mens were' oriented with the longitudinal axis parallel to (longitudinal orientation) the. major working direction of the p!ate. The WOL specimens in Capsule T were machinea such that the simulated crack in the specimen would propagate parallel to the weld-direction. All WOL specimens were fatigue precracked per ASTM E399-73 requirements. Charpy V-notch specimens from the weld metal were oriented with the longitudinal -axis of the specimen's normal to (transverse orientation) the weld direction. Tensile specimens were oriented with the longitudinal axis of the specimen normal to (transverse) the weld direction. 4-1

Capsule T *contained dosimeter wires of pure copper, iron, r:ickel, ~nd aluminum - 0.15% cobalt (cadmium-shielded and unshielded). In addition cadmium shielded dosimeters of neptunium (Np237) and uranium (U 238) were contained in the capsule. Thermal monito~s made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. The composition of the two alloys and their melting points are as follows: 2.5% Ag, ~7.5% Pb

1.75% Ag,-0.75% Sn, 97.5% Pb Melting point: 5?9° F (304° C)

Melting point: 590° F (310° C) The arran!;ement of the various mechanical test specimens, dosimeters and thermal. monitors c-Jntained in Capsule Tare shown in Figure 4-2. 4-2 - *~

~ I I TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE SALEM UNIT 2 REACTOR VESSEL SURVE!LLANCE MATERIALS Plate* B 4712~2 Weld-Metal Westinghouse Lukens Steel - Westirighouse Element Analysis Cal Analysis -Analysis [a] c ---- _0.23 0.22 -0.10 _s --:__ 0.010 0.015 0.011 N2 0.004 [b) 0.007 Co 0.015 [b) 0~024 Cu~-- 0.10 [b) a*.23 Si-0.30 0.24 0.29 Mo - 0.55 0.55 0.45 Ni 0.61 0.60 0.71 Mn---~ 1.34 1.37 1.27 ~. Cr-:::-_ 0.089 [b) 0.015 -V ----- [a) [b) 0.001 p - 0.015 0.011 0.017 Sn 0.008 _[b) --0.005 Al 0.030 --[b) -' 0.007 [af--All elements not listed are less than 0.010 weight-percent. [b]-Not measured. Heat Treatment Heat Treatment

Temperature Time (hr)

Coolant (° F) Material Intermediate I 550° /1650° F 4 Water quench~d Shell Plate B 4712-2 1225° F +/- 25° F 4 Air cooled 1150° F +/- 25° F 40 Furnace cooled Weldment 1150° F + 25° F 40 Furnace cooled 4-3 I

(3.17)X CAPSULE---. (TYP) (1.02)W-- (1.02)V--

======--~~===-=-====- -------- ---- 270° goo THERMAL SHIELD CORE BARREL Z(1.02) 5(1.02) Figur~ 4-1.

_Arrangement of Survemance Capsules in the Salem Unit '.'? Reactor Vessel (updated lead factors for the capsules shown in parentheses) 4-4

TENSILE n I I I I I I I I I I I 11 L: WOL Co Co!Cdl 590F MONITOR SPECIMEN NUMBERING CODE: JT - PLATE B 4712-2 (transverse orientation) JL - PLATE B 4712-2 (longitudinal orientation) JW - CORE REGION WELD METAL JH - HEAT-AFFECTED-ZONE METAL WOL WOL WOL TENSILE CHARPY I I 11 NI I I I I Co I 11 I I 'I u w n I I I

I CHAR PY CAPSULE T NP"'

u~ 005\\MEYE.R BLOCK [] NI CORE CHARPY CHARPY I I I I I I Ll u n 11 I '+-<t--Fo 11 11 CHARPV CHARPY NI Fo 'CAPSULE T ~ 45° CHARPV CHAR PY I I I I I I I I Co I I I I I 11 u u n I I I I I I I CHAR PY CHARPY CHAR PY 579°F MONITOR e Co Co(Cd) I

'---------------------------CENTER REGION OF VESSEL--------------------------------------~

YO TOP OF VESSEL Figure 4-2. Capsule T DiagrPin Showing Location of Specimens, Thennal Moniton;, c;;,~ Dosimeters 4-5 IL. ~.

I SECTION 5 TESTING OF SPECIMENS FROM CAPSULE T - i. 5-1'.- OVERVIEW -- The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclec.r Energy Systems personnel. -Testing was performed in accordance wi_th 1 OCFR50, Appendices G and _H, ASTM Specification E185-82 and Westinghouse_ Procedure MHL 7601, Revision 3 as modified-by AMF - Procedures 8102 and 8103. Upon receipt of the capsule atthe laborntory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8824. 111: No discrepancies were found. Examination of the two low-melting 304° C (~79° F) and 310° C (590° F) ~utectic alloys indicated no melting of either type ofth,:~rmal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed _was less than 304° C (579° F). _ The Charpy impact tests were performed per ASTM Specification E23-82 and AMF Procedure 8103 on a Tinius-Olsen Model. 74, 358J machine. The tup (striker) of the Charpy mEtchine is instrumented with an Effects Technology model 500 instrumenta-tion system. With this system, load-time and energy-time ~ignals can be recorded in - addition to the standard measurement of Charpy energy (E o ). From the load-time curve, the load* of general yielding (PG'.'), the time to general yielding (t GY ), the maximum load (PM), and the time to maximum load (t M) can _be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. - _The load at which fast fracture was initiated is identified asthefastfracture load (PF), and the load e.t which fast fracture terminated is identified a~ the arrest load (PA). 5-1

The.energy at maximum load (EM) was determined by comparing the energy-time record and the.load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (E p) is the difference between the total energy to fracture (E o) and the energy at maximum load. The yield stress (u y) is calculated from the three point bend formula. The flow stress is calculated from the average of the yield and maximum loads, also using the three point bef!d formula. Percent shear was determined from postfracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77. The lateral expan~. sion was measured using a dial gege rig similar to that sho~n in the same specifi- . cation. Tension tests were performed on a.2~,000-pound lnstron, split-console test machine (Model 1115) per ASTM Specificatons E8-81 and E21-79, and RMF Procedure 8102. All pull rods, grips, and pins were made of lnconel 718 hardened to R c 45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a con'stant crosshead speed of0.05 inch per minute through-out the test. Deflectiod measurements were mcde with a linear variable displacement transducer . * (l:.VDT) extensometer. The exten!;ometer knife edges were spring:.loaded to the. specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class 8-2 per ASTM E83-67. Elevated test temperatures were obtained With a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in *air: Because of the difficulty in remotely attaching a*thermocouple directly to the speci-men, the following procedure was used to monitor specimen temperature. Chromel-alum&i thermocouples were inserted in' shallow holes in the,center and each end of . the gage section of a dummy specimen and in each grip. In test configuration, with a

s'tight load on the specimen, a plot of specimen temperature versus upper and lower grip-~nd=c-6ntrolier temperatures was ~eveloped over the range room temperature to 550° F (2S8° C). The upper grip was used to control the furnace temperature. During 5-2

the actual testing the grip temperatures were used to obtain desired specimen temperatures. Experiments indicated that this method is accurate to plus or minus 2~F. The yield load, ultimate load, fracture load, total elongation, and uniform elongation. were determined directly from the load-extension curve. The yield strength, ultimate .strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) --~rid percent reduction in area was c,omputed using the final diameter measurement. 5;.,2..

-CHARPY V-NOl'CH IMPACT TEST RESULTS The re§ult§ of Charpy V-notch impact tests performed on the various materials con-tained fn Capsule T irradiated at 2.56x1018 n/cm2 are presented in Tables5-1 through 5~8 and Figures 5-1 through 5-4. A summary of the transition temperature increases and uppc.~~ shelf energy de ;reases for the Capsule T material is shown in Table 5-9.

lrradiatio-ri.:. of plate B 471~-2 material (transverse orientation) to 2.56 x 1018 n/cm2 (Figure 5-1) resulted in 30 and 50 ft lb transition temperature increases of 70° *and 65° F, respectively; and an upper shelf eriergy decrease of 8 ft lb. lrradiati_on of plate B 4712-~-~aterial (longitudinal orientation) to 2.56x1018 n/cm 2 (Figure 5-2) resulted in 30 and 50 ft lb transition temperature increases of 50 and 60° F; respectively, and an

upper shelf energy* decrease of 7 ft lb, Weld met~L irradiated to 2.56 x 1018 n/cm 2 (Figure 5-3) resulted in 30 and 50 ft lb transitiontemperature increases of 155 and 180° F, respectively, and an upper shelf energy decrease of 32 ft lb.

Weld HAZ~ metal irradiated to 2.56 x 1018 n/cm2 (Figure 5-4) res!Jlted in 30 and 50 ft lb transition temperature increases of 115 and 125° F, respectively, and an upper shelf energy decrease of 32 ft lb. The fraetare appearance of each irradiated Charpy specimen from the various nia'.'"' terials is 0shown in Figures 5-5 through 5-8 and show an increasing ductile or tougher appearance with increasing test temperature. 5-3 . I

Figur~ 5-_~ show~ a comparison of the 30 ft lb transition temperature increases for the. various Salem Unit 2 surveillance materials with predicted increases using the methods 'c:)f NRC Regulatory Guide 1.99, Revision 1. 131 This comparison shows that the transition increases resulting from irradiation to 2.56x10 18 n/cm 2 are greaterthan predicted by the Guide for plate B 4712-2, transverse and longitudinal orientations. The weld metal transition temperature increase resulting from 2.56 x 1018 n/cm 2 is also greater than predicted. Since the shifts in transition temperature are greaterthan predicted by the Guide, and the *chemical contents for the intermediate and lower shell vertical seams welds are estimated, future plant operating limits, as discussed in Appendix A, are being based on the Lipper limit of the Regulatory Guide l.'~9 prediction curve.

5-3.

TENSION TEST RESULTS.. The results of tension tests performed on plate B 47i2-2 and weld metal irradiated to 2.56 x 1018 n/cm2 are shown in Table 5-10 and Figures 5-10 and 5-11, respectively. These results show that irradiation produced an increase iri 0.2 percent yield strength of 11 ksi for plate B 4712~2 and* approximate~y 21 ksi for the weld metal. Fractured

. tension specimens for each of the materials ;He shown in Figures 5-12 and 5-13. A typical stress-strain curve for the tension specimens is shown in Figure 5-14.

5-4.. _ 0 WEDGE OPENING LOADING TE~*,T RESULTS The wedge opening loading (WOL) specimens that were contained in Capsule Twill be reported at a later time.* -~. ~l I,, ' 5-4

~*-- -

TABLE 5-1 CHAR PY* V~NOTCH IMPACT DATA FOR THE SALEM UNIT 2 INTERMEDIATE SHELL PLATE B 4712-2 (transverse) IRRADIATED AT 550°F, FLUENCE 2.56x1018 n/cm2 (E>1 MeV) Sample . Temperature lmpa-:t Energy Lateral Expansion. No. oc (oF) Jouks (ft-lbs) MM (mils) JT54 -18 (

0) 8.0

( 6.0) 0.14 ( 5.5).

JT52*

-4 ( 25) JT59 10 ( 50) 24.S (18.0) 0.47 (18.5) JT53 21 ( 70). 35.5 (26.0) 0.67 (26.5) JT57 24 ( 75) 34.0 (25.0) 0.62 (24.5) JT49. 38 (100) 49.0 (36.0) 0.81 (32.0) JT60' 66 (150) 76.0 (56.0) 1.30 (51.0) JT56 79 (175) 92.0 (68.0) 1.52 (60.0) JT50 93 (200) 115.0 (85.0) 1.77 (69.5) JT51 107 (225) 119.5 (88.0) 1.78 (70.0). JT55 121 (250) 122,0 (90.0).* 1.92 (75.5) JT58 149 (300) 127.5 (94.0) 2.06 (81.0).

Test Malfunction 5-5 Shear

(%) 6 2.1 27 25 82 66 70 100 . 100. 100 100

I ---~

----~-

- - ----- ________.,., - *-*'-*-~~--__..,__ ___.. _._ ---*******------------~-~'"--**------C_~--* TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE. SALEM UNIT 2 INTERMEDIATE SHELL PLATE 8 4712-2 (longitudinal) IRRADIATED AT 550° F, FLUENCE 2,56 x 1018 n/cm 2 (E > 1 MeV)

- Sample Temperature Impact Energy Lateral Expansion
Shear No~

oc (° F) Joules "(ft-lbs) MM (mils) (%) JL40 -18 (

0) 8.0 { 6.0)

C1.23 ( 9.0) 7 JL36 10 ( 50) 30.0 ( 22.0) 0.51 (22;5) 19 JL33 38 (100) 47.5 ( 35.0) Cl.57 (22.5). 38 JL34 66 (150) 92.0 {. 68.0) .. 45 (57.0) 54 JL39 93 (200) 139.5 (103.0)

?.02

{79.5) 94 JL38 107 (225) 150.5 (111.0)

2.06 (81.0) 100 JL35 121 (250) 153;0. (113.0)

'!.97 (77.5) 100 . JL37 149 (300) 162.5 {120.0} '.!.29 (90.0) 100. 5-6.

TABLE 5-3

CHARPY V-NOTCH IMPACT DATA FOR THE SALEM UNIT 2 PRESSURE VESSEL. WELD METAL IRRADIATED AT 550° F, FLUENCE 2.56 x 10 18 n/cm 2 (E > 1 MeV)
-Sample=*

Te.mperature Impact Energy Lateral* Expansion No. oc (oF) Joules (ft-lbs) MM. (mils) JW51 -32 (-25) 17.5 (13.0) 0.24 ( 9.5). JW54 . -18 (

0) 26.0 (19.0) 0.38. (15.0)

JW60 10 ( 50) 27.0. (20.0) 0.41 (16.0) JW50 21 ( 70) 38.0 (28.0) 0.60 (23.5) JW52-38 (100) 44.5 (33.0) 0.71. (28.0) Jw59.:.. 52 (125) 32.5 (24.0) 0.60 (23.5) JW53* 66 (150) 42.0. (31.0) 0.79 (31.0) JW57 66 (150) 30.0 (22.0) 0.60 (23.5) JW58

93 (200) 77.5 (57.0) 1.27. (50:0)..

JW49 107 (225) 95.0' (70.0) 1.55 (6l~O)" JW56* 121 (250) 104*.5 (77.0) 1.68 (66.0) JW55* 149. (300) 108.5 (80.0) 1.70 (67.0) 5-7 Shear (%) 6 13 27 29 38 33 55 53 83 95 100 100 --*-::.~

-1

--~-----*---~. -~~~--~-- . e TABLE 5-4 CHARPY V-NOTCH IMPACT DATA FOR THE SALEM UNIT 2 PRESSURE VESSEL WELD HEAT AFFECTED ZONE METAL IRRADIATED AT 5~0° F, FLUENCE 2:56 x 1018 n/cm2 (E > 1 MeV) Sample Temperature No~ oc (° F) JH52: -46 (-50) JH54 -46 (-50) JH58 -18 (

0)

JH49* -4 ( 25) JH56. -4 ( 25) JH51. 10 ( 50) . JH55 38 (100) JHSO 52. (125) JH59-66. (150).

JH57'-

93 (200) . JH60* 121 (250) . JH53 149 (300) Impact Energy

Joules (ft-lbs) 44.5 (33.0) 34.0 (25.0) 44.5 (33.0) 32.5 (24.0) 82.5 (61.0) 114.0.(84.0)*

99.0 (73.0) 112.5 (83.0) 1.16.5 (86.0). 124.5 (92.0)

)._

Test Malfunc~ion...

5-8 Lateral Expanskm *- MM (mils) 0.58 (23.0) 0.57 (22.5). 0.57 (22.5) a.so* (23.5) 1.26 (49.5) 1.44 (56.5) . -1.37 (54.0) 1:73 (68.0) 1.73 (68.0) 1.99 (78.5) Shear (%) 45 21 19. 22 65 100 79 100. 100 100

01 I co Sample No. JT54 JT52* JT59 JT53 JT57 JT49 JT60 JT56 JJ50 JT51 JT55 JT58 Test Temp. (oC) -18 -4 10 21 24 38 66 79 93 107 121 149 I Charpy Energy (Joules) 8.0 24.5 35.5 34.0 49.0 76.0 92.0 115.0 119.5 122.0 127.5 . TABLE 5-5 i I i I

I

. i I -1 i ' . INSTRUMENTED CHARPY IMPACT TEST: RESULTS FOR l l THE SALEM UNIT 2 INTERMEDIAT;E SHELl:- PLATE B 4712-2 (transverse) Normalized Energies I Charpy Maximum Prop Yield Time Maximum Time to Ed/A Em/A Ep/A Load to Yield Load Maximum (kJ/m2) (kJ/m2) (kJ/m2) (N) (µSec)* (N) (µSec) 102 63 39 14500 100 16300 110 305 133 172 14100 120 17700 205 441 ?89 151 14700 110 18200 345 424 305 119 14500 100 18200 355 610 ~99. 211 14200' 95 18500 445 949 445 504** 13600 95 18200 505 . 1152 . 450

703 *12900 110 18200 525 1441 440

'1000 13200 100 . 18q00 510 1491 520 971 13200 95 18300 585 I 1525 508 1017 12100 95 179'00 585 1593 512 1081 12400 100. 176p0 595

Test Malfunction Fracture Arrest Yield Load Load Stress (N)

(N) (MPa) 16300 0 746 17500 1900 726 18200 3500 755 18200 3200 748 18400 5900 730 17700 10000 699 16500 10300 666 682 679 625 638 Flow Stress (MPa) 791 818 .846 844 841 818 801 811 811. 773 771 I \\ '.i ,f,., ,';i 1;! d \\* I ei ']

CJ1 I ~ 0 Sample No. .JL40 JL36 JL33 JL34 JL39 JL38 JL35 JL37 Test Temp (oC) -18 10 / 38 66 93 107 121 149 Charpy Energy (Joules) 8.0 30.0. 47.5. 92.0* 139.5. 150.5 153.0 162.5. TABLE 5.;5 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE SALEM UNIT 2 INTERMEDIATE SHELL PLATE B 4712-2 (longitudinal) Normalized Energies Chai'py Ma~lmum Prop. Yield Time MaxlmL1m Time to Ed/A Em/A Ep/A. Load to Yield Load Maximum (kJ/m2) * (kJ/m2) (kJ/m2) (N) (µSec) (N) (µSec) 102 68 33 17100 120 373 227 146. 15300 120 17900 295 593 374 219 14400 115 18900 435 1152 535. 617 13700 105 18800 595 174G 542 1204 12900 . 85 18500 595 1881 559 1322 11300 65 18500 610 1915 581 1334 11600 80 18200 650. 2034 593 1441 11100 90 18100 675 'i Fracture Arrest Yield Flow Load Load Stress Stress (N) (N) (MPa) (MPa) 17100 800 18100 900 785 853 18500 3700 740 856 17900 8700 704 836 11900 8400 662 808 580 767 597 768 572 752

01 I...... Sample No. JW51 JW54 JW60 JW50 JW52 JW59

.JW57 JW53 JW58 JW49 JW56 JW55 Test Charpy Temp Energy

(°C) (Joules) -32 17;5 -18 26.0 10 27.0 21 38.0. 38 44.5 52 32.5 66 30.0 66 42.0 93 77.5 107 95.0 121 104.5 149 108.5

1 I.

TABLE 5-7 INSTRUMENTE[) CHARPY IMPACT TEST RE~ULTS SALEM UNIT 2 WELD METAL Normalized Energies * .Charpy Maximum Prop Yield Time Maximum Time to Ed/A Em/A Ep/A Load to Yield Load Maximum (kJ/m2) (kJ/m2) (kJ/m2) (N) (µSec) (N) (µSec) 220 167* 53 17700 100 19800 195 322 262 60 17400 100 19900 280 339 247. 9~. 16300. 115 18300. 295. 475 344 130. 16000 100 20500 360 559 422 137 16000 100 19900 435 407 304 102. 15300 100* 18100 350 373 135 237 14100 90 16100 185 525 ~98 227 13100 115 17400 370 966 455 511 13800 95 '18200 510 1186 443 743. 13500 95 17600 . 505

1305 441

' 864" 13400 90 17800 505 1356 514 841 12700 115 17500 605 Fracture Arrest Load Load (N) (N) 19300 0 19900 500 17900 2600 19900 1200 19900 300 . 18000 1200 16000 7000 17400 7500 17300 12700 Yield Stress (MPa) 909 896 839 824 823 787 727 675 710 692 688 651 Flow Stress (MPa) 963 961 890 940 923 859 779 .785 824 798 803 777 I i i I, f I __ C

01 I ~ I\\) Sample

No.

J°H54 JH52 JH58 JH56 JH49* JH51 JH55 JH50 JH59 JH57 JH60* i.!H53 Test Temp (oC) -46 -46 -:18 -4 -4 10 38 52 66 93 121 149 Charpy Energy (JQules) 34.0 44.5 44,5. 32.5. 82.5 114.0 99.0 112.5 116.5 124.5 TABLE 5-8 INSTRUMENTED CHARPY IMPACT TEST RESU.LTS FOR THE SALEM UNIT 2 WE~D HEAT AFFECTED ZONE METAL Normalized Energies Charpy Maximum Prop Yield Time Maximum Time to Ed/A Em/A Ep/A Load to Yield Load

Maximum (kJ/m2)

(kJ/m2) (kJ/m2) (N) (µSec) (N) (µSec) 424 259 165 17800 110 19600 285 559 349 210 17600 100 20600 350 559 508 51 17200 125

20700 530 407 320

.. 87.16400 110 19200 360 1034 . 573 461 16000 115 19900 600 1424 6q1 112* 15100 95 19800 '670 1237 637 600 14700 110 19600 680 1407 457 .950 14600 100 18600 510 1458 560 898 13900 90 18500 610 1559 497 1063 11700 85 17200 580

Test Malfunction i

I ' Fracture Arrest Yleld Load Load Stress (N) (N) (MPa) 19600 2600 917 20300 7800 904 20500. 0 885 18800 3300 844 18300 9100 822 16000 7000 777 17400 12000 758 751 714 604 Flow Stress (MPa) 964 983 976 915 924 898 884 855 832 743 i.. ~. .. t. ' **1 r

r.

I I ., ',I ..~ ' \\.,,,

TABLE5-9 THE EFFECT OF 550° F IRRADIATION AT 2.56 x 1018 (E > 1 MeV) ON THE NOTCH TOUGHNESS PROPERTIES OF THE SA~EM UNIT 2 REACTOR VESSEL MATERIALS. Average Ave~age 35 mil Average 30 ft lb Temp (° F) Lateral Expansion Temp (° F) 50 ft lb Temp (° F) Material

Unlrradlated Jrradlated AT Unlrradlated Irradiated AT Unlrradlated Irradiated AT Plate 10 80 70 40 105.

65 60 125 65 B 4712-2 (transverse) Plate 30 80 50 45 90 45 55 115 60 B 4712-2 (longitudinal) HAZ Metal -125. -10 115 -65 45 110 -95 30 125 Weld Metal* -30 125 155 -10 140 150 0 180 180 1 Average Energy Absorption at F~ll Sf'lear (ft lb) Unlrracllated Irradiated A (ft lb) 97 89 8 122 115 7 120 88 32

e.

111 79 32

TABLE 5-10 TE~SILE PROPERTIES FOR SALEM UNIT 2 'REACTOR VESSEL MATERIAL IRRADIATED TO 2.56 x 1018 n/cm2. Test .2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp Strength Strength Load Stress Strength Elongation Elongation in Area No. Material (o F) (ks!) (ksi) . (kip) (ksi) (ksi) (%) (%) (%) Plate .)T-9 8 4712-2 225 74.2 92.2 3.23 175.7 65.8 9.3 19.3 63 (transverse) Plate JT-10 8 4712-2 550 67.2 92.2 3.50 150.6 71.3 8.5 17.7 53 (transverse) JW-9 Weld 250 . 81,0 95.7 3,55 183.4 72.3 9.8 19.1 61 Metal JW'-10 Weld 550 76.4 93..7 3.52 175.1

71.. 7..

9.0 18.0 59 Metal

-~-----*-

100 80 ~ 0 - a: 60 w I 40 ti) 20 2 0 100 en "E 80 z 0 60 z < ~ 40 w ...I < a: 20 w ~ ...I 0 LEGEND o Unirradiated

Irradiated at 2.56 x 1018 n/cm 2 140 120 100 UNIAAADIATED\\

.c "j' 80

=..

IRRADIATED (550° F) CJ AT 2.56 x 10 18 n/cm 2 a: 60 w z w 40 20 2 0 -100 0 100 200 300 400 TEMPERATURE (° F) Figure 5-1. Irradiated Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Intermediate Shell Plate B 4712-2

(transverse orientation) 5-15

-~:.. -- - --------*-- - ---------**--**.:=-~---:-::::- ___==::~--:-=-====-::-:-----:--__,..,..., ___ ".""_~'"'."'-_:__""".'*-....,<,,.,.~_:___,=-....,..--~----:._,,---~-'-_--_--: _:____=---....,_:~-~:~--~.,..,. __ __,..,.._ -.:::-=--=----=-----_-_-__ ---___ -. -:__--__ --__ - 100 80 ,e Cl - a: < 60 U.I -~ (/) 40 20 0 - 100 .!!! *e - 80 z 0 (/) 60 z ~ )( U.I 40 ..I < a: U.I 20 !;( ..I 0 140 120 100 .ci 80 c.:J a: U.I 60 z U.I 40 20 0 Figure 5-2. .:.100 3 LEGEND O Unirradiated -

Irradiated at 2.56 x* 10 18 n.cm 2

3 0 100 - 200 TEMPERATURE {° F) IRRADIATED (550° F) AT 2.56 x 10 18 n/cm 2 300 400 Irradiated Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Vessel Intermediate Shell Plate B 4712-2 (longitudinal orientation) 5-16

~ Cl - a: -ct !.!I J: en . en e - z 0 en z ~ LU ...J < a:: LU ~- ...J -.ci "'j CJ c: LU z LU 100 80 60 40 20 o 100 80 60 40 20 2 0 140 120 100 80 Figure 5-3.

- ---~- -------

2 2 LEGEND O Unirradiated

Irradiated at 2.56 x 1018 n/cm 2 IRRADIATED (550° F)

.....:.::.::...:_ __ '4+---1'-.,_...----..... ---- 2.56 x 1018 n/c m 2 100 200 300 400 TEMPERATURE (° F) Irradiated Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Pressure Vessel Weld Metal S-17 i - I I

100 80 ~ ,50 a: < w

c 40

(/) 20 0 100 .!! e - 80 z (J ~ 60 '.?:

- (

>>b. .( 40 r.u ..I

4:

20 w ..I 0 180 160 140 120 .a 9:-- 100 0 a: 80 w z w 60 - 40 20 0 Figure 5-4.

****-*---* -------~---*- -

0 110°F LEGEND o Unirradiated e Irradiated at 2.56 x 10 18 n/cm 2 0 0 -100 0 100 200 300 400 TEMPERATURE (° F) Irradiated Charpy V-Notch Impact Properties for Salem Unit 2 Reactor Pressure Vessel Weld Heat-Affected Zone Metal

-* ~J. -""..... ~~-.::-=--~*--**--* -- I I I 1* I I 1* I JT 54 JT 57 JT 50 Figure 5-5. JT 52 . JT 59 JT 53 JT 49 JT 60 JT 56 JT 51 JT 55 JT 58 Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Pressure Vessel Intermediate Shell Plate B 4712-2 (transverse orientation) 5-19

-=====--------:-: - -~- -. ~.- -1 e e 1 JL 40 JL 39 Figure 5-6. JL 36 JL 33 JL 34 JL 38 JL 35 JL 37 Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Pressure Vessel Intermediate Shell Plate B 4712-2 (longitudinal orientation). 5-20 I I., i

JW 51 JW 52 JW 58 Figure 5-7. . JW 54 JW 60 JW 50 JW 59 JW 57 J\\V 53 JW 49 JW 56 JW E-5 Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Weld Metal 5-21

JH 54 JH 56 JH 59 Figure 5-8. JH 52 JH 58 JH 49 JH E:1 JH 55 JH 50 JH 57 JH-60 JH-53 Charpy Impact Specimen Fracture Surfaces for Salem Unit 2 Weld Heat Affected Zone Metal 5-22

01 I I\\) (,) 500r-~~~~~~~~~~~~~~~..,-~~~~~~~~.,..-~~~~~~~~~~~~ 400 300 200 100 90 80 70 60 50 1011 2 0 PLATE B 4712-2 (transverse) Q PLATE B 4712-2 (longitudinal) Q WELD METAL 3 4 5 6 7 8 9 1!::' 4 n... C" 6 789 Figure 5-9. FLUENCE (n/cm 2 ) Comparison of Actual versus Predicted 30 ft lb Transition Temperatu.re Increases For The Salem Unit 2 Reactor Vessel Material Based On The Prediction Methods of Regulatory Guide 1.99, Revision 1 I. A ~ r I* 1*

ULTIMATE 100 TENSILE STRENGTH~ Vi ~ s-..__._ "-2 en °<::2 en 80 w 0.2% YIELD a: ~ENGTHS.,, en 0- -e 60 40..... ~~~..._~~~"--~~~-~~~~ ......... ~~~--~~~- o 100 200 300 400 500 600 LEGEND O Unirradiated

lrradiat.~d at 2.56 x 10 18 n/cm 2 80 70 e-v-s 60 REDLJCTION,_

,,,e 50 IN AREA Q - g

J 40 j::

(.) 30 c 20 ~ 8 10 UNIFORM L_

0 ELONGATION 0

100 200 JOO 400 500 600 TEMPERATURE (° F) Figur~ 5-10. Tensile Properties for Salem Unit 2 Reactor Vessel Intermediate Shell Plate B 4712-2 (transverse orientation) 5-24

100 2 U) / U) 80 LU a: U) 0.2%YIELD < °"""= STRENGTH 60 e -9 40 0 100 200 300 400 500 600 LEGEND 0 Unirradiated

Irradiated at 2.56 x 10 18 n/cm 2

80 ~ 8 70 ~EDUCTION IN;.. --~o 60 AREA b 50 .,,e 0 - 40 ..J 0 30 2

> c g..

~ TOTAL ELONGATION<!: 1 0 20 8 2_,.,,o-. 2, ~ 10

z;.

() UNIFORM ELONGATION 00 100 . 200 300 400 500 600 TEMPERATURE (° F) Figure 5-11. Tensile Properties for Salem Unit 2 Reactor Vessel Weld Metal 5-25

I I I 1 I Figure 5-12. Fractured Tensile Specimens of TENSILE SPECIMEN JT 9 Tested at 225° F TENSILE SPECIMEN JT 10 Tested at 550° F Salem Unit 2 Reactor Vessel Intermediate Shell Plate B 4712-:i (transverse orientation) 5-26 1

- -"-'---~ TENSILE SPECIMEN JW 9 Tested at 250° F TENSILE SPECIMEN JW 10 Tested at 550° F Figure 5-13. Fracture Tensile Specimens of Salem Unit 2 Reactor Vessel Weld Metal 5-27

Ol I I\\) CX> UJ UJ w ii a: Q. UJ 120000 108000 96000 84000 72000 60000 48000 36000 24000 12000 0 " " .03 .09 SPECIMEN JT-9 ---~- -

-~- ------___..__ _____ __._ __ ~

.12 .15 in/in STRAIN .18 .21 .24 .27 .3 Figure 5-14. Typical Stress-Strain Curve For Tensile Speclmel'}s

--~ ~*- -~ *- -~ -_.,.-_-. - - ~*-..=...~:.::.------- ,_:_,,:::-~, -----=---------~---------*--------** ---.. ---~-,- -*--.;-~-~-..:---~----- ------ --- ----- - ----~---" SECTION 6 RADIATION.ANALYSIS AND NEUTRON DOSIMETRY 6-1. INTRODUCTION-Knowledge of the neutron environment within the pressure vessel - surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons. First, in the interpretation of radiation induced properties changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second, in relating . the changes observed in the test specimens to the present and future condition of the - reactor pressure vessel, a relationship between the environment at various positions within the.reactor vessel and that experienc*3d by the test specimens must be estabiished. The former requirement is normall:1 m*et by employing a combination of rigorous analytical techniques and measuremerits ob~ained with passive neutron flux monitors contained in each of the surveillance: capsules. The latter information is derived solely from analysis. This section describes a discrete ordinates Sn transport analysis performed for the Salem Unit 2 reactor to determine the fast neutron (E > 1.0 MeV) flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules and, in turn, to develop lead factors for use in relating the neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum averaged reaction cross sections derived from this calculation,* the analysis of the neutron dosimetry contained in Capsule T is discussed and comparisons with analytical - predictions are presented. 6-2. DISCRETE ORDINATES ANALYSIS A plan view of the Salem Unit 2 reactor geometry at the core mid plane is shown in *

Figure 6-1. $ince the reactor exhibits 1/8th core symmetry, only a Oto 45° sector is depicted. Eight irradiation capsules attached to the thermal shi~ld are included in the design to constitute the reactor vessel surveillance program. Four capsules are
located.symmetrically at.4 and 40° from the cardinal axes as shown in Figure 6-1.

6-1

-~-----**-....... ~~_::_..:_*_. ____ ~~------ -.:.. *- - --~----- - - oo 40. { CAPSULES} s,v,w,z

--------~-----~~---~--...... **-* --------- -----~--~--- ---*----- - -------------- -

-'~--~--'\\ { CAPSULES} T,U,X, Y 45° / Figure 6~1. I Salem Unit 2 Reactor Geometry 6-2

.. ---**--"--* ~~- A plan view of a single surveillance capsule attached to the thermal shield is shown in Figure 6-2. The stainless steel specimen container is 1-inch square and approximately 38 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 3 feet of the

12-foot-high reactor core..

. From a neutronic standpoint, the surveillance capsule structures are significant. In. fact, as is shown later, they have a marked impact on the distributions of neutron flux and energy spectra in the water annufds between the thermal shield and the reactor vessel. Thus, in order to properly ascertain the neutron environment at the test specimenclocations, the capsules themselves '"must be included in the analytical model: Use of at least a two-dimensional computation is therefore mandatory. In the:analysis of the neutron environment within the Salem Unit 2 reactor geometry, predictions of neutron flux magnitude and energy spectra were made with the DOTl4l two-dimensional *discrete ordinates code. The radial and azimuthal distributions were obtained from an R, 9 computation wherein the geometry shown in Figures 6-1 and 6-2 was described in the analytical model. In addition to the R, e computation, a

second-- calculation in R, Z geome;try was also carried out to obtain relative axial variations of neutron flux th rough out the geometry of interest In the R, Z analysis, the reactor core was treated as an equivalent volume cylinder and, *of course, the surveillance capsules were not included ih the model.

-Both _the-R, e and R, Z analyses employed 47 neutron energ15 groups and a P 3 expansion of the scattering cross sections. The cross sections used in the analyses were obtained from the SAILOR cross section library!51 which was developed specifically for light water reactor applications. The neutron energy group structure used iffthe analysis is listed in Table 6-1. A key input parameter in the analysis of the integrated fast neutron exposure of the reactor vessel is the core power distribution. For this analysis, power distributions representative of time-averaged conditions derived from statistical studies of long'- term operation of Westinghouse 4-loop plants were employed.. These input distribu-tions include* rod-by-rod spatial variations for all peripheral fuel assemblies. It should-be noted that this generic dssign basi~ power distribution is intended to provide acvehicle for long-term (end-of-life) projection *Of vessel exposure. Since . 6-3

(3° OR 39°) CHARPV SPECIMEN THERMAL SHIELD -~.... ;. Figure 6-2. *. Plan View of a Reactor Vessel Surveillance CapsJ1e 6-4

1 TABLE 6-1 47GROUPENERGYSTRUCTURE Lower Energy Lower Energy Group (MeV)

Group. (MeV) 1 14.19[a) 25 0.183 .2 12.21 26 0.111 . :3. 10.00 27 0.0674 A* 8.61 28 0.0409

- 5 7.41 29 0.0318 6

6.07. 30 0.0261 7 4.97 31 0.0242 -. --_8 3.68 32 0.0219 9 3.01 33 0.0150

10-2.73 34 7.10 x 10-3 11 2.47 35

. 3.36 x 10-3 12 2.37 36 1.59 £10-3 _1_3_ 2.35 37 4.54 x 10-4 _1_4_ 2.23 38 2.14x10-4 15 1.92 39 1.01 x 10-4 16 1.65

40.

3.73 x 10-5 17 . 1.35 41 1.07 x 10-5 5.04 x :10-s tS; 1.00 42* 1a 0.821 43 1.86 x 10-6 20: 0.743 44 8.76 x 10-7 21 0.608 45 4.14x10-7 22 0.498 46 1.00 x 10-7 23 0.369 47 0.00 x 24 0.298 [a] The Ll;Jper energy of group 1 is 17.33 MeV 6-5

I I plant s~iecific power distributions reflect only past operation,Jl:leir use for projection into the*-future may not be justified; the use of generic data which reflects long-term

*operation of similar reactor cores may provide a more suitable approach.

Benchmark testing of these generic power distributions and the SAILOR cross sections-a:gainst surveillance capsule data obtained from 2-loop and 4-loop Westing-house pla-lits indicates that this analytical approach yields conservative results with calculations exceedin,g measurements from 10 to 25%. [SJ One further point of interest regarding these analyses is that the design basis assumes an out-in fuel loading pattern (fresh fuel on the periphery). Future commitment to low leakage loading patterns* could significantly reduce the

calculatea=-:-neutron flux levels presented in Paragraph 6-4. *IQ _addition, capsule lead

- factors *col.ild be changed, thus impacting the withdrawal schedule of the remaining surveillance capsules.

Having the results of the R, e and A, Z calculations, three-dimensional variations of neutron frux may be approximated by assuming that the following relation holds for the applicable regions of the reactor. 4>(R,Z,SEg) = Q>(R,0Eg) F(Z,Eg) (6-1) where: 4>(R,Z,0E g) = neutron flux at point R,Z,e within energy group g 4>(R,e,E 9 ~ = neutron flux at point R,e within ene.rgy group g obtained from the R,e calculation F(Z,E g) = relative axial distribution of* neutron flux within energy group g obtained from the R,Z calcula~ion 6-3. NEUTRON DOSIMETRY The passive neutron flux monitors included in Capsule T of Salem Unit 2 are listed in Table 6-2>The first five reactions in Table 6-2 are used as fast neutron monitors to 6-6

-~*..:;__"'-':...::.:.:_.:....:..~*. -----:.-~.._.._....._.,_ ___ . -** -~--*---- -~---*-**---~---*-~*-* "'--'---- ~ relate neutron fluence (E > 1.0 MeV) to measured materials properties changes. To properly account for burnout of the product isotope generated by fast neutron . reactions, it is necessary to also determine the magnitude of ttie thermal neutron flux at the monitor location. Therefore, bare and cadmium-covered cobalt-aluminum monitors were also included. -~. TABLE 6-2 NUCLEAR CONSTANTS FOR NEUTRON FLUX MONITORS CONTAINED IN THE SALEM UNIT 2 SURVEILLANCE CAPSULES Target Weight Product Monitor llilaterial Reaction of Interest Fraction Half-life Copper Cu63 (n,a) Co60 0.6917 5.27 years Iron

Fe54 (n,p) Mns4 0.0585 314 days

. Nickel Niss (n,p) Cose 0.6777 71.4 days a]. u23a (n,f) Cs 137 30~2 years Uranium238' 1.0

13?£a]

Np231 (n,f} Cs 137 1.0 30.2 years . Neptunium* Cobalt-aluminum [aJ Cos9 (n,y) Co60 0.0015 5.27 years. Cobalt-aluminum Cos9 (n,y) Co60 0.0015 5.27 years [a] Denotes that monitor is cadmium shielded. Fission Yield (%) 6.3 6.5 The relative locations of the various monitors within the* surveillance capsule are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the center of the capsule. The use of pasf-ive monitors su~h as those listed in Table 6-2 does* not yield a direct measure of the Gnergy-dependent flux level at the point of interest. Rather, the activa-tion or fission process is a measure of the integrated effect that the time-and energy- .6-7

-~-* --~~* . dependerif neutron flux has on the target material over the course of the irradiation period: An-accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest: The operating history of the reactor The energy response of the monitor The neutron energy spectrum at the monitor location The physical characteristics of the monitor The analysis of the passive monitors and subsequent derivation of the average. neutron flux requires completion of two procedures. First, the disintegration rate of

product isotope per unit mass of monitor must be determined. Second, in order to define a suitable spectrum averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated.

The specific activity of each of the monitors is determined using establish.ad ASTM I procedures. [7.s.9*10*111 FolJowing sample preparation, the activity of ~ach monitor is determined by means of a lithium-drift~d germanium Ge(Li);. gamm~ spectrometer. * .. The overall standard deviation of the measured data is a function of the precision of .. sample weighing, the uncertainty in counting, and -the acceptable error in detector ~ calibration~ For the samples removed from Salem Unit 2, the overall 2 u deviation in the mea_sured data is* determined to be plus or minus 10 percent. The neutron energy

. spectra are determined analytically* using the method descri.bed in Paragraph 6-1.

Having thEf.measured activity of the monitors and the neutron energy spectra at the. locations -of interest, the calculation of the neutron flux proceeds as follows. The reaction product activity in the monitor is expressed as 'Fi= : 0 f; vJ. u(E)lfJ(E)dE+/- PPj . E. i=", max -i\\t. -i\\td (1-e J)e . 1, (6-2) 6-8

-~-::-_.~:- *,_. - .,,.__....:._.,,..;..:.,;-i.::*=-~*~.abi.:o*~~l'1:~c.. ;."!:..;:!;:i;H_,_.:..o;o;U.... *i-'*:A.__ ___ ~..,.-**------*--*--*-** ~--... --~*----* __,__........,,. _ _.....:__-_-~-**-..*.* ;..-:...-...~ ...... -~.:.*-*

where:

R = induced_product activity N 0 = Avogadro's number A =. atomic weight of the target isotope . f i = weight fraction of the target isotope in the target material Y = number of product atoms produced per reaction ( u(E) = energy-dependent reaction cross section

- : - ¢'(E) = energy-dependent neutron flux at the monitor location with the reactor at full power P j = avernge core power level during irradiatio*n* period j Pmax = maximum or reference core power level t* J decay constant of the product isotope length of irradiation period j.

td decay time following irradiation period j Becaus~ _peutron f.lux distributions are calculated using multigroup transport methods.~nd, further, because the prime interest is. in the fast neutron flux.above 1.0 MeV, spe¢trum-averaged* reaction cross sections are defined such that the integral term in ~quation (6-2) is replaced by the following relation: Ju(E) ¢(E)dE = a ¢(E > 1.0 MeV) E 6-9

1------~~-~* ~--'--~~-----. e where: N f""' a(E) </J (E)dE L ag </Jg 0 g=1 a = = (""' </J (E)dE J1.0Mev N L <Pg g=g: 1.0 MeV Thus, equation (6-2) is rewritten N No. . L Pj -i\\t* -i\\td R *= -- fj.Y *a </J (E > 1.0 MeY)... . (1-e J) e. A P max* j=1 or, solving for the neutron flux, </J (E > 1.0 Me~/)' = _______ R _______ N 0 N P* -i\\t* -i\\t f i ya~_.,(l-e J) e d (6-3) A L-J P_max * . j=l The total fluence above 1.0 MeV.is. then given by where: N ~.(E > 1.0 MeV) = </J (E > 1.0 MeV) 2:. l t.. Pmax 1 1=1 (6~) N 'LP* -.. _J_ ti'= Pmax total effective full power seconds of reactor operation up. to the time of capsule removal 1=1 6-10

An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered Co59 (n,y) Co~ data by means of cadmium ratios and the use of 37-barn, 2,200 m/sec cross section. Thus, 4'Th = ___

_R_b_a_re_.~~~

0 _0_-_ 1 _J ___ _ (6-5) N No 2: Pj -iH. -At --f. ya

-- (1-e J) e d

A Pmax j=1 where: - 1Yis defined as Abare Red covered . 6-4. TRANSPORT ANALYSIS RES UL TS Results of the Sn transport calculations for the Salem Unit 2 reactor are summarized in Figures 6-3 through 6-8 and in Tabies 6-3 through 6-5*. In Figur~ 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressur_e vessel _inner radius, 1/4 thickness location, and 3/4 thickness location are presented as a function of azimuthal angle. The influence of the surveillance capsules on the fast neutron flux distribution is clearly evident. In Figure 6-4, the radial distribution of maximum fast neutron flux (E > 1.0 MeV) through the thickness of *the reactor pressure vessel is shown. The relative axial variation of neutron flux within the vessel _ is given j~ ~igure 6-5. Absolute axial variations of fast.neutron flux may be obtained by multiRIY.ing the levels given in Figure 6-3 or 6-4 by the appropriate values from Figure 6~~:- In Figure 6-6, the radial variations of fast neutron flux within surveillance capsules at4 and 40° are presented. These data, in conjunction with the maximum vessel flux, are used to develop lead factors for each of the capsules. Here the lead factor is defined as the ratio of the fast neutron flux (E > 1.0 MeV) at the dosimeter block location (capsule center) to the maximum fast neutron flux at the pressure vessel inner radius. Updated lead factors for all of the Salem Unit 2 surveillance capsules are listed in Table 6-3. 6-11

-~~-,,__-"'".........__~---------*. -*-. ***-* *----

~~-====

.. *...... ---------------*-----* 1011 SURVEILLANCE 8 CAPSULES R = 211.41 cm 6 4 PRESSURE VESSELIR u 2 GJ In*- I., -1/4 T LOCATION E u c --

~ 1010

...J

u.

.Z 8 0 a: I-6 w

z.

4 . 3/4 T LOCATION .2 109...... ~~~_..~~~~ ...... ~~~---~~~~..-~~~-----~~~ 0

Figure 6-3.

10 20 30 40 50 60 AZIMUTHAL ANGLE (deg.) Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel - Surveillance Capsule Geometry 6-12

-~... ;..... _....... _......:- -~::* *:*'*-~ -:.-*

~

~*._-

*::.<::::,::~-~-::*.:~=~~-----~~-~~--'"---.--~----------------- -

. - ---- -------~ u Cll Ill I 6 4 E 2 u ....... c - -I LI. 1010 z 0 cc... LLI z 8 6 4 10s ____._ __.___..____..____..____. __ __.. __ ~~....a.~__._-.-__._ __ _._ __ -'-_o_R~ 214 - 216 218 220 222 224 226* 228 230 232 234 236 238 240 242 Figure 6-4. RADIUS (cm) Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel 6-13

I - -~---------'-=--=--*...o.o.--;;-===-=o=-""7=--.o..-_- 1 I ~-----~--*--~----~--~-----. ***10° r'.'.""'-------:::::i-------~-------------..., 8 6 4 2 10-1 8 ..I LL. 6 z 0 a: w z 4 ~ 2 ~- LL! cc 10-2 8 6 4 2 CORE MIDPLANE TO VESSEL CLOSURE HEAD 10-3 ~......_ __ --..lL...-----lL...-----1----L---"-~---....1 -300 -200 -100 0 100 200 300 400 DISTANCE FORM CORE MIDPLANE (cm) Figure 6-~. Relative Axial Variation of Fast Neutron Flux (E > 1.0 MeV) Within the* Pressure Vessel 6-14

u GI. in NI E u ...... c - ..J IL z* 0 a: w z 1012 r-----------------------------------------...;_---------. 8 6 2 1011 8 6 4 2 1010 8 6 4 2 THERMAL SHIELD. 40° CAPSULES 4° CAPSULES CAPSULE CENTER TEST SPECIMEN 10 9 ______ __.__.._ __ _._. ____ -L ______ __. ______ _._ ____ ....,~ ..__~ 207 208* 210* 211 212 213 214 209 RADIUS (cm) Figure-6-6. Calculated Radial Distribution of Maximum Fast* Neutron Flux (E > 1.0 MeV) Within the Surveillance Capsule 6-15

10 9 r-~~~~~~~~~___;~~~~~~~~~~~~~~~~-- 8 6 4 2 10s 8 6 T~ en ..!! Cl "C ~ 2 j:: CJ 107 c 8 w I-ct a: 6 ~ 4 en 2 106 - 8 6 4 2 10s* 207 ~ ! THERMAL SHIELD 208 - Figure 6-7. Niss (n,p) Coss -- Np237 (n.f) Cs 137 _. r 211.41 u~3s (n~ cs137 Fe;4 (n,p) Mns4 CAPSULE CENTER Cu;;3 (n,a) Co 60 -- TEST SPECIMENS 209 210 211 212 213 214 RADIUS (cm) Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules Located at 40 Degrees . 6-16

I =

c

---~ I.

-~**=-~- -

~ -~--~ :-..::...~.. -_....,".i..;::-..:;.,:.?;i.;.:-!*.i:z.t..:.:.~~*:.-f::!..:.S~,,._-.:.:...~2.:.:..:.*._,.., __...... :..*.:..:....-.~-*-----*-

_ ---~. _,... -;; ___ .___...,.~.:-::..:.*, *. J~~~.J.:.J..;f,j...::..*..:.....~**:;..::.~::::..~---:-- -**~-,,..~~;,:.-.:....-{.,!~~~-.. ~o.~.::.:...:..~-,..!..~.-.I:.~*~~~-~~~~.s:_s.<,;~ i e I I r 10 8.-~~~~~~~~~~~~~~~~~~~~~~~~~--- 8 6 4 2 107 8 6 2 105 8 6 4 2 THERMAL* SHIELD Ni53 (n,p) Ca58 Np231 (n,-fa Cs 131 _ -.......... __ 211.41 Fe 54 (n,p) Mn 54 I u23B (n;f7 Cs 137 - ........._:::::::~===~~-- I. CAPSULE CENTER Cu 63 (n,d) Ca60 _._..._...__~ 1 TEST SPECIMENS 10 4 --~~~---_;.r,~--i~~__.:;ii....i........ ~~-'-~~~ .... ~--'~1.----L-.......I 207 Figure 6-8. 208 209 210 211 212 213 214 RADIUS (cm) Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules Located at 4 Degrees F:-17

i

*-*-. -*~-----*---*------ -------------------------- --

e TABLE 6-3 CALCULATED FAST NEUTRO~ FLUX (E > 1.0 MeV)

AND LEAD FACTORS FOR SALEM UNIT 2 SURVEILLANCE CAPSULES-Capsule Azimuthal

1.0 MeV) Lead Identification Location (deg) (n/cm2-sec) Factor

s.

4.

3.04 x 1010 1.02 v 4 3.04 x 1010 1.02 w 4 3.04 x 1ofo 1.02 z 4 3.04 x 1010 1.02 T 40 9.44 x 1010 3.17 9.44 x 1010 u 40 3.17 x 40 9.44 x 1010 3.17 y 40 9.44 x*.1010 3.17 Since the'neutron flux monitors contaired within the surveillance capsules are not all located at the same radial location, the :neasured disintegration rates are analytically adjusted for the gradients that exist witt;in the capsules so th~tflux and fluence levels may be derived on a common basis at Ci common location. This point cf comparison. . was* chosen to be the capsule center. Analytically determined r13acti9n rate. gradients

- fo"r"usefrlthe adjustment procedures ar.~ shown in Figures 6-IaQd.6-8forcapsules at 4 and 40°. All of the applicable fast-neutron.reactions are included.

In. order to derive n_eutron flux and fluence levels from the measured disintegration ~ates, su,ifable spectrum-averaged reaction cross sections are required. The neutron energy spectrum calculated to exist at the center of each of the Salem surveillance. capsules* fs listed in Table 6-4; The associated spectrum-averaged cross sections for each of the fast neutron reactions are given in Table 6-5. 6-5.. DOSIMETRY RE.SULTS The irradiation history of the Salem Unit 2 reactor-up to-the tirQ.~ 9f removal of Capsufe T'is listed in Table 6-6. Comparisons of measured a:ri.9--calculated saturated activity-of the 11ux monitors contained in Capsule T based on the irradiation history shown* in Table 6-6 are given in Table 6-7. The data are presented as measured at the 6-18 ___ _,____* _.:.; ___ ~----* --*--- TABLE 6-4 CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF THE SALEM UNIT 2 SURVEILLANCE. CAPSULES Group ¢ (n/cm2-sec) Group ¢ (n/cm2-sec) No. 4° Capsules 40° Capsules No. 4° Capsules 40° Capsules t 1.35 x 107 2.08 x 107 25 8.59 x 109 3~42 x 1010 2 4.85 x 107 7.65 x 107 26 8.10 x 109 3.29 x 1010 3 1.56 x 108 2.67 x 108 27 6*.50 x.109 2;67 x 1010 4 2.74 x 108 4.89 x 108 28 4.80 x 109 1.99 x 1010 5 4.32 x 108 8.20 x 108 29 1.68 x 109 6.92 x 109 6 9.33 x 108 1.85 x 109 30 1.04 x-109 i 4.27 x 109 7 1.18 x 109. 2.57 x 109 31 1.11 x ro9 7.15 x 109 8 2.07 x 109 5.17 x 109 32 1.05 x 109 4.41 x 10.9 9 1.62 x 109 4:54 x 109 33 2.52 x 109 1.05x1010 10 1.27 x 109 3.71 x 109 34 4.21 x 109 1.75 x 1010 11 1.46 x 109 4.38X*109 35 5.!56x 109 2.35 x 1010 12 7.19 x 108 2.19 x 109 36 5.16x109 2.17 x 1010 13 2.12 x 108.

6.52*x 108 37 7.79 x 109 3.31 x 1010 14 1.04 x 109 3.22 x 109 38 4.42 x 109 **

1.88 x 1010

15 2.67 x 109 8.37 x 109 39 4~68 x 109 2.0:1 x 10 1 ~

16 . 3.21x109 1.05x1010 40 9 6.27 x 10 2.Tt x 1010 17 4.67x 109 1.57 x 1010 41 . 7.59 x 109 3.31 x* 1010 18 8.45 x 109 2~98 x 1010 42 4.33 x 109 l.90 x 1010 19 5~73 x 109 2.09 x 1010. 43 5~24 x 109 2.31 x 1010 20 2.83 x 109. 1.04*x 1010 44 3.46 x 109 l.53 x 1.010 2l 8.14 x 109 3.14 x 1010 45 2.'93 x 10£> l.29 x 1010 22 6.21 x 109 2.45 x 1010 46 5.59 x 109 2.41 x 1010 23 7.46 x 109 2.93 x 1orn 47 1.41*x1010 5.66 x 1010 24 6.51 x109 2.59 x 1010 6-19 .*. ~ TABLE6-5 SPECTRUM-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF SALEM UNIT 2 SURVEILLANCE CAPSULES Reaction Fe54 (n,p) Mn54 Cu63 (n,a) Co60 Niss (n,p) Co58 Np231 (n,f) Cs 131 u23e (n;j.) Cs 131 I. u {barns)" Capsules at 4° 0.0980 0.00112. 0.127 2.62 0.385 u = \\:u(E)</l(E)ctE. ~ 00 ¢( E)d E.. 1.0 MeV Capsules at 40° 0.0735 0.000659 0.0993 2.83 ll.385 actual monitor locations as well as adjusted to the capsule center. All gradient adjLJst-nients to the ~apsule center were based on the data presented in Figure 6-7. The*.fast neutron (E >. 1,0 MeV) flux and fluence :evels derived for* Capsule T are *. presented iq table 6-8. The thermal neutron flux obtained.from the cobalt-aluminum monitors is summarized in Table 6-9. Due to the relatively low thermal neutron flux at

the capsule location, no burnup correction* was made to any of the measured activities. The maximum error introduced by this assumption is estimated to be less than 1" percentfolthe i\\liss (n,p) Coss reaetion and even less significant for all of the other fast neutron reactions.
An examination of Table S-8 shows that t_he fast neutron flux (E > 1.0 MeV) derived from the five threshold reactions ranges from 6.91 x **,010 to 7.93 x 1010 n/cm2-sec, a total span of less than 15%. It may also be noted thatthe calculated flux value of 9.44x 1010 n/cm2-sec exceeds all of the. measured values with calculation to experimental ratios ranging from 1.19 to 1.37. This behavior is consistent with prior benchmarking studies.

6-20 TABLE 6-6 IRRADIATION HISTORY OF SALEM UNIT 2 SURVEILLANCE CAPSULE T Irradiation Pj Pmax Pj/ Time* Decay Time

Month-~ Year

. (MW) (MW) Pmax (day) (day) 6 1981

54.

3565* .015 28 930 T

1981 2985.

3555* .. 837 31 899 8 1981

2439.

3565 .684 31 868 9 1981 1699. 3565 .477 30 838

~

10 1981 1725.

3565 .484 31 807 -11 1981 2465. 3565 .691 30 777 12* 1981 2771. 3565 .777 31 746 1 1982 3012. 3565 .845 31 715 2 1982 3311. 3565 .929 28 687 3 1982 3388. 3565' .950 31 656 4

1982 2973.

3565 .834-30 626 5 1982 3367. 3565 .945 31 595 6 1982 .*3415. 3565 .958 30 565 7 1982 2402. 3565 .674 31 534 8 1982 . 2672. 3565 .749 31 503 9 1982 2170. 3565 . 609 30 473. 10 .1982 2632. 3565 .738 31 442 11* 1982 2565.

'3565*

. *.719 30 . 412 12 1982 2422. 3565 .679 31 381 1 1983 1469. 3565 - .412 . 22 359 EFPS =.3.66E+ 07 SEC EFPY = 1.16 NOTE: (1) Decay time is referenced to 1/16/84. (2) Total irradiation time equal to 3.66 x 107 EFPS. 6-21 . ':* i '*1 I .------------~---

i.

I i ' l TABLE 6-7 i COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX I MONITOR SATURATED ACTIVITIES FOR CAPSULE T I i i I l1' . I' I I I I. Saturate~ Activity Adjusted ~at~rated Actlvl.ty I l*.. 111 1, I Ir* 'I I i I I Ji '

I I r di /s I I,,,,,

I I!'

Ill di /S 1, 1 I
I Iii Reaction Rad la I g

g and Location Capsule T Calculated Capsule T Axial Position (cm)* Calculated Fe54 (n,p) Mn54. Top 211.68 3.23 x 106

3.40 x* 1'06 Top-Mb'dle 211.68 3.23 x 106 3.40 x 106 Middle 211.68

. 3.09 x 106 3.26 x 106 Bottom-Middle 211.68 3.18 x 106 3.35 x 106 Bottom 211.68 3.18 x 106 3.35 x 106. Average 3.18 x 106 4.30 x 106 3.35 x 106 4.53 x 106 , Cu63 (n,a) Co 60 I Top-Middle 211.18 3.33 x 105 3.17 x 105 Middle 211.18 3A4 x 105 3.27 x 105 Bottom-Middle 211.18 3.36 x 105 3.19 x 105 Average 3.38 x 105 4.3~ x 105 3.21 x 105 4.11 x 105 Ni58 (n,p) Co58

Top-Middle 212.18 4.57 x 107 5.25 x 107 Middle 212.18 4.54 x 107 5.21 x 107.

Bottom-Middle 212.18 4.59 x 107 5.27 x 107 Average 4.57 x 107 5.75 x 107 5.24 x 107 6.59 x 107 Np231 (n,f) Cs 131 Middle 211.41 3.71 x 107 4.41 x 107 3.71 x 107 4.41 x 107 u23B (n,f) Cs 137 Middle 211.41 4.82 x 106 5.31 x 106 4.82 x 106 5.31 x 106 '. i

I l

=- ---**

;---,,..~--=*.

Comparisons of measured and calculated current fast neutron exposures for Capsule T as well as for the inner radius of the pressure vessel are presented in Table ~10. Measured values are given based on the Fe54 (n,p) Mn54 reaction alone as well as for the average of all five threshold reactions. Based on the data given in Table 6-10,

the best estimate exposure of Capsule T is:

~T = 2.56 x 10 18 n/cm 2 (E>1.0 Me'1__: Since the calculated fluence levels were based on conservative representations of

~'?re power distributions derived fdr long-term operation while the Capsule T data are representative only of cycle 1 operation, it is recommended thaf projections of vessel toughness into the future be based on design basis calc!Jlated fluence levels.

Withdrawal of future surveillance capsules should further sub~tantiate the adequacy of this approach. -:1 6-23 - i

m I I\\) ~ I TABLE 6-8 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULE T Adjusted Saturated Activity dis/s </J (E > 1.0 MeV) </J (E > 1.0 MeV) g (n/cm2-sec) (n/cm2) Reaction Measured Calculated Measured Calculated Measured Calculated Fe54 (n,p) Mn54 3.35 x 106 4.53 x 106. 6.99 x 1010 9.44 )( 1010 2.56 x 1018 3.46 x 1018 Cu6:3 (n,a) Co6*:i 3.21 x 105 4.11 x 105 7.37 x 1010 9.44 x 1010 2.69 x.1018 3.46 x 1018 Ni58 (n,p) Co58 5.24 x 107 6.59 x 107 7.50 x 1010 9.44 x 1010 2.74 x 1018 3.46 x 1018 Np231 (n,f) Cs 131 3.71 x 107 . 4.41 x 107 7.93 x 1010 9.44 x 1010 2.90 x 1018 3.46 x 1018 u238 (n,f) Cs 131 4.24 x 106 5.31 x 106 6.9*1 x 1010 . 9.44 x 1010 2.53 x 1018 3.46 x 1018 U238 adjusted saturated activity has been multiplied by 0.88 to correct tor 350 ppm U235 impurity. I[ i

--*~:2~=~:~~:;;_,: :*~~_::=-~:~:*:~~-.::::,..__~=*.*~*>'-*-**--**.. **----*-*-:.-::':.~:~~-- -*-*-* *---=---.-~ ---~~~~~* --~._.:..:::.~"~~* ~-. ~-~-*~->'A<>C"- :* -. -~~-~~~jl e e TABLE.6-9 . RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULE T Saturated Activity dis/s g . :_.::Axial ¢th =tocation Bare Cd - Covered (n/cm2-sec) .. -!!JP 6.69 x 107 2.70 x 107 7.04 x-1010 Bottom 6.55 x 107 2.55 x 107 7.09 x 1010 Average 6.62 x 107 2.63 x 1*07 7.07 x 1010 6-25

-~*---~~~ -~~~~~=~ -*c------~~~~--~---------~~~---~-1 TABLE 6-10

SUMMARY

OF NEUTRON DOSIMETRY RESULTS FOR CAPSULE T Current ¢ (E > 1.0 MeV) EOL ¢* (E > 1.0 MeV) . (n/cm2). (n/cm2) Location Measured Calculated Measured Calculated -Capsule T 2.56 x 1018 3.46 x 1018 Vessel IR 8.08 x 1017 1.09 iC1018 _., 2.23 x 1019 3.01 x 1019 Vessel 1/4T 4.48 x 1017 6.05 x 1017. 1.23x1019 1.67 x 1019 Vessel 3/4T 9.22 x 1016 1.24x*1017

1.87x1018 3.43 x 1018 NOTE: EOL fluences are based on operation *at 3565 MWt for 32 effective-full-power years.

6-26.

SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule is recommended for future capsules to be removed from the Salem Unit 2 reactor vessel. This removal schedule meets the requirements of ASTM E185-82. Capsule Identification T u x y s v w z TABLE 7m1 SALEM UNIT 2 SURVEILLANCE CAPSULE REMOVAL SCHEDULE Orientation Removal of Lead Time c;_1psules [a]

Factor[bJ (EFPY) 40° 3.17 1.16 (Removed) 140° 3.17.

3 220° 3.1_7 6 320° 3.17 10 40 1.02 32 176° 1.02 Standby 184° 1.02 Standby 356° 1.02 Standby [a] Reference Irradiation Capsule Arrangement Drawing, Figure 4-1. Expected Capsule Fluence {n/cm2 ) 2.56 x 1018 6.24 x 1018 1.25 X 1019[c) 2.08 x 1019 [d] 2.14 x 1019 [b] The factor by which the capsule fluence leads the vessels maximum inner wall fluence. [c] Approximate Fluence at 1/4 wall thickness at End-of-Life. [d] Approximate Fluence at vessel inner wall at End-of-Life. 7-1

SECTION 8 REFERENCES

1. Davidson, J. A., Smith, R. A. "Public Service Electric and Gas Company Salem Unit No: 2 Reactor Vessel. Radiation Surveillance Program," WCAP 8824 January 1977.

2. ASTM Standard E185-73, "Recommended Practice for Surveillance Tests For Nucle-~r Reactor Vessels" in ASTM Standards, Part 10 (1973), American Society for Testing and Materials, Philadelphia, Pa. 1973.

. 3. Reguratory Guide 1.99, Revision 1, "Effects of Residual.Elements on Predicted Radiation Damage to Reacto*r Vessel Materials," U.S.-Nuclear Regulatory Commission, April 1977. ... ______ 4.. SoltesZ. R. G., Disney, R. K., Jedruch, J., and Zeigler, S. L., "Nuclear Rocket Shielding Methods, Modification, Updating and Input Data-Preparation. Vol. 5 - Two-Dimensional Discrete. Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.

5. SAILOR RSIC Data Library Collection DLC-76, "Coupled, Self-shielded, 47 Neutron, 20 Gamma-ray, P 3, Cross Section* Library for Light Water Reactors."

6. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology.

to* be published.

7. ASTM Designation E261-77, Standard Practice for Measuring Neutron Flt.ix, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards (1981), Part 45, Nuclear Standards,~- 915-926, American Society for Testing and Materials, Philadelphia, Pa., 1981.

8-1

--~:-. - --~--***---~---'-* --'-~~ **--'--- *-*--* -- -*-* - - - .*----.....'...._*--~---*--__:.... *---* -*-* -***.:......-* --'-----' I*

8. ASTM_ Designation E262-77, "Standard Method for Measuring Thermal Neutron

-Flux-by Radioactivation Techniques," in ASTM Standards (1981 ), Part 45, Nuclear Standards, pp. 927-935, American Society for Testing and Materials, Philadelphia, Pa., 1981.

9. ASTM Designation E263-77, "Standard Method for Measuring Fast-Neutron Flux_

by Radioactivation of Iron," in ASTM Standards (1981 ), Part 45, Nuclear Standards, pp. 936-941, American Society for Testing and Materials, Phila-delphia, Pa., 1981.

10.

ASTM Designation E481-78, "Standard Method of Measuring Neutron-Flux
Density by Radioactivation of Cobalt and Silver," in ASTM Standards (1981), Part
- 45; Nuclear Standards, pp. 1063-1070, American Society for Testi'ng and Materials, Philadelphia, Pa., 1981.

11. ASTM Desig~ation E264-77, "Standard Method for Measuring Fast-Neutron Flux bx Radioactivation of Nickel," in ASTM Standards (1981), Part 45, Nuclear Standards, pp. 942-945, American Society for Testing and Materials, Phila-

. delphia, Pa., 1981.

~

'i.

APPENDIX A SALEM UNIT 2 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A-1. INTRODUCTION Heatup *and cooldown limit curves are calculated using the most limiting value of ~-- 'I RTNDT.. (reference nil-ductility temperature). The most limiting RT NOT of the

material in the core region of the reactor vessel is determined by using the preservice *.

reactor vessel material properties and estimating the radiation-induced.6.RT NOT*:* RT NOT i_s designated as the higher of either the drop weight nil-duC:tility transition. temperature (NOTT) or the temperature at which the material exhibits at least 50ft lb: of impact *energy and 35-mil lateral expansion (normal to the major working direction) minus 60° F.

.RJ" NOT.Increases as the material is exposed to fast-neutron radiation~ Thus, to find the*mosflimiting RT NOT at any time period in the reactor's life,.6.RT NOT due to the
radiation exposure associated with that time period must be added to the original un'"'

irradiated_RTNOT The extent ofthe shift in RT NOT is enhanced by certain ch*emica(

elements (such as copper and phosphorus) present in reactor vessel steels.. Design curves which show the effect of fluence and copper and phosphorus contents on..
ART NOT for reactor vessel steels are shown in Figure A-1.

Given the copper and phosphorus contents of the most.limiting material; the radiation-induced.6.RT NOT can be estimated from Figure A-1. Fci.st-neutron fluence-* \\E >1 MeV) at the 1/4T (wall thickness) and 3/4T (wall thickness)vessel locations are. given as a 'function of full.:.power service life in Figure A-2. The data for all otherferritic materials in the reactor cool'1Jlt pressure b~undary are examined to insure that no other component will be limiting with respect to RT NOT- -A-1

.:._.::;~ ~** ______ _____.:._- --..--~*---.............. ____ . __ _,_.~ *-* --- A-2" FRACTURi: TOUGHNESS PROPERTIES -The preirradiation fracture-toughness properties* of the Salem Unit 2 reactor vessel -materials are presented in Table A-1. The fracture-toughness-properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan. [11 The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the Salem Unit 2 Vessel Material Surveillance Program.

A-3.

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME:approach for calculating the allowable limit curves for various heatup and . cooldowi'f rates specifies that the total stress intensity factor,_-K i-, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K IR, for the metal temperature at that time. f<*1R is obtained fr-om the reference fracture toughness curve, defined in Appendix G to the ASME Code. r21 The K IR curve is given by the equation: K IR.= 26.78 + 1.223 exp [0.0145 (T-RTNOT + 160)] (A-1) where ~IA is the reference stress intensity factor as a function of the metal -temperature-.T and the metal reference nil-ductility temperature RT NOT. Thus, the governin~f equation of the heatup-cooldown analysis is defined in Appendix G of the ASME Coder21 as follows:*_ (A-2)

1. "Fracture Toughness Requirements," Branch Technical Position MTEB No. 5-2, Chapter 5.3.2 in Standard Review Plan, NUREG-0800, 1981.

2. ASME Boiler and Pressure Vessel Code, Section Ill, Division 1 -

Appendices, "Rules for Construction of Nuclear Vessels," Appendix G, "Protection Against Nondu.ctile Failure," pp. 559-569, 1983 Edition. American Society of Mechanical Engineers, New York, 1983. A-2

~~---------**---- where KIM is the stress intensity factor caused by membrane (pressure) *_stress K It. is the stress intensity factor caused by the thermal gradients K IR is a function of temperature relative to the RT NOT of the mq.terial C = 2.0 for Level A and Level B service limits C = 1.5, for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, K IR is determined by the metal temperat~re at the tip of the postulated flaw; the appropriate value for RT NOT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are* calculated a11d _then the corresponding (thermal) stress intensity factors, K It, for the reference fla*w are computed~ From Equation (A-2), the pressure stress intensity factors are obtained. and, from these, the allowable pressures.are calculated..

For the: e~lculation of the allowable pressure-versus-coolant temperature during

- cooldqwn,-;:.the Code reference flaw is assumed to exist at the insi_de of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase* with increasing cooldown rates. Allowable pressure-temperature relations are generated-for both steady-:-state and finite cooldown rate situations~ From these relations, composite limit curves are const.ructed for each cooldown rate of interest. The use ot the composite in the cooldown analysis is necessary because control of the cooldown procedure is based* on measurement of reactor coolant temperature, . w'hereas the limiting pressure is actually dependent on the n;aterial temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state A-3

. *---------*--* --:..:..c.~-- ~-----*. ---*~---* __ _:-:-***----* --*---~*-..:.--------*-* -----~~:_.:: __.J situation. 0lt follows that, at any given reactor coolant temperature, the ~ T developed during cooldown results in a higher value of K IR at the 1/4T location for finite cooldown-rates than for steady-state operation. Furthermore; if conditions exist such that the increase in K1R exceeds K It, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above' procedures are needed because there is no directcontrol on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of-cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the, system for the entire cooldown period. Thi*ee separate calculations are required to determine the limit curves forfinite heatup ratr.s. As rs-done in the cooldown analysis, allowable pressure-temperature relation-ships are developed for steady-state conditions as well as finite heatup rate conditions assuming-the presence of a V4T defect~at the inside of the vessel wall. The thermal gr~dients during heatup produce compressive stresses at the inside of the wall that allfNiate tne tensile stresses produced by internal pressure. The metal temperature at th~* crack trp lags the coolant temperature; therefore, the K IR for the 1/4T crack during heatup is Tower than the K IR for the 1/4T crack during steady-state conditions at t_he same coolant temperature. During heatup, especially at the end of the transient, co11ditions may exist such that the effects of cdmpressive thermal stresses and lower . KL={ 's do not offset each other, and the pressure-temperature curve based ori steady- . state concHtions-no longer represents a lower bound of all _$Jrnilar curves for finite heatup* rales when. the 1/4T flaw is considered; Therefore, -both cases have to be analyzed'-iri order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady:-state and finite heatup rates is obtained. The second_ portion of the heatup analysis concerns* the calculation of pressure'"" temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed.- -Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in _ nature and. thus tend to r~inforce any pressure stresses present. These thermal stresses:=-a.re dependent on both the rate of heatup and the time (or coolant temperatiire) along the *heatup ramp. Since the thermal stresses at the outside are tensile. ana increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. A-4

-..--<......., _ _..-..::=--~*-::.* : *.....:,.* ~--~-~-.

=;-_--.--= -

-*-*--;::.*.-*.*: -~.--- _._..__~~_:.-**---: _. ~~.,,;. __ ~:....;,~,:.!.'~.:0~~~:...:..;;1.!.'!.{..._L-;_*.:,_*'-::~.-'""-.,;;,~*J'.....:.:.* _, __.. _.. _,_; __,4.. _..._ ** -**-*** -**-*.* -...:'... -~--------------"~~.--.;.,~~~:!'",~;_-,;::....._.~~-* --*** -......:..;.*,....,:!.-..;.~.:.:-....-_;.:._:_*~*~*

~

....... ~~~ Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given te_mperature, the allowable pressure is taken to be the lesser of the three values taken 1rom the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup . ramp, the controlling condition switches from* the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Then, composite curves for the heatup rate data and the cooldown rate data are adjusted for*

possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

'HEATUP AND COOLDOWN LIMIT CURVES. Limit curves for normal he.atup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in Section A-3. The derivation of the limit curves is presented in the NRC Regul~tory Standard Re~iew Plan. l1l Transition temperature shifts occurring in. the pressure vessel materials due to radiation exposure have been obtained directly fr_om the reactor pressure vessel

surveillance pro'gram. Charpy test specimen~:: from Capsule.T indicate _that the representative core region weld metal and.. the limiting core region plate (8 4712-2) exhibited shifts in RT NOT of ~55° and 70° F, ri:;!spectively.Comparison of these shifts

with predicted increases using the methods of NRC Regulatory Guide.199 [21 shows that the RT NOT for the weld and plate material shifted nwre than predicted (see
Figure A-1 ). Since the shifts were greater thari predicted and t.r~ intermediate and lower shell vertical weld seams chemistries were estimated", the heatup arid cool down limits for the vessel are based on the upper bound of the Regulatory Guide 1.99 trend curves in Figure A-1. Heatup and cool down limit curves for normal operation of the re~ctor vessel are presented in Figures A-3 and A4 and represent an operational time *

. period of 7 effective full power years. *.

1. "Pressure-Temperature Li mi.ts; Chapter 5.3.2 in Standard Review Plan, NUREG-0800, 1981.

2. Regulatory Guide 1.99, Revision 1, "Effects of Residual Elements*on Predicted Radiation Damage.to Reactor Vessel Ma~*

terials", U.S. Nuclear Regulatory Commission, April 1977.. A-5

1
.,,1

_..~-=-----*-

=--._... *-~-:..

..:-.:._.:._.. _~.:;:.....:.....:.......:..~-'---~*---~----- ____.. ___ ~. Allowable *combinations of tef'Tlperature arid pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown -curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure A-3. This is in addition to other criteria which must be met before the reactor is made critical. The leak test limit curve shown in Figure A-3 represents minimum temperature re-quirements at the leak test pressure specified by applicable codes. The leak test limit curve was: determined by methods of References 1 and 2. Figures A-3 and A-4 define limits tor insuring prevention *of nonductile failure.

1. "Pressure-Temperature Limits", Chapter 5.3.2 in Standard Review Plan, NUREG-0800, 1981.

2. ASME Boiler and Pressure Vessel Code, Section Ill, Division 1 ~ Appendices. "Rules for Construction of Nuclear Vessels", Appendix G, "Protection Against Nonductile Failure", pp. 559-569, 1983 Edition, American Society of Mechanical Engineers; New York, 1983.

A-6

)> I....... Component Closure Head Dome Closure Head Pe~I Closure Head Peel Closure Head Peel Closure Head Flange . Vessel Fl~nge Inlet Nozzle Inlet Nozzle Inlet Nozzle Inlet Nozzle Outlet Nozzle Outlet Nozzle Outlet Nozzle Outlet Nozzle Upper Shell Upper Shell Upper Shell Inter. Shell Inter. Shell Inter. Shell Lower Shell Lower Shell Lower Shell Bottom Head Peel Bottom Head Peel Bottom Head Peel Bottom Head Inter. and Lower Shell Vert. Welds Inter. to Lower Shell Girth Weld TABLE.A-1 SALEM UNIT 2 (PN,.J) REACTOR VESSEL TOUGHNESS DATA Cu p TNDT Plate No. Grade 1%) (%) (Of) 8470B A5338CL1 0.11 0.014 -40 85007-3 A5338CL1 . 0.12 0.012 -20 84707-1 A5338CL1. 0.10 0.010 0 84707-3 A5338CL1 0.13 0.011 0 84702-1 A50BCL2 0.010 2B* 8E001 A50BCL2 0.010 12* 84703-1 A50BCL2 0.010 60* 84703-2 ASOBCL2 0.010 60*' 84703-3 A50BCL2 O.Q10 60* 84703-4 A50BCL2 . 0.011 60* . 84704-1 A508CL2 0.006. 60* B4704-2 A50BCL2 0.006 60* 84704-3* A50BCL2 0.007 2B* B4704-4 A50BCL2 0.007 60* B4711-1 'A533BCL1 . 0.11 0.010 o* B4711-2 A533BCL1 0.14 0.009 -10 B4711-3 A533BCL1 0.12 0.009 -10 B4712-1 A533BCL1 0.13 O.D12 0 B4712-2 A533BCL1 0.14 O.Q11 -20 B4712-3 A533BCL1 0.11 O.D10 -50 B4713-1 A533BCL1 0.12 0.010 -10 B4713-2 A533BCL1 0.12 0.010 -20 B4713-3 A533BCL1 0.12 O.D12 -10 84709-1 A533BCL1 0.12 0.010 -30 B4709-2 A533BCL1 0.12 0.011 -20 B4709-3 A533BCL1 0.11 0.009 -20 B4710 A533BCL 1 0.12 0.009 -30 0.30*** o* 0.17 .022 o*

Estimated pc:r NRG Standard Review Plan Section 5.3.2
100% Shear not reached

.... Estimated 50 ft-lb 35 mll Temp RTNDT (Df) (D F) 45* -15* 15* -20* 51* o* 66* 6* 39* . 2B* 4* 12* 62* 60* 25* 60* 32* 60* 40* 60* B* . 60* 20* 60*, B* 2B* 40* 60* 50* o* 60* o* BB* 28* <60 0 72 12 70 10 68 B 6B B 70 10 54* -6* 42* -18* 71* 11* 60* o* o* o*

~

Average Upper Shelf Energy Normal to Prlncl1,>al Principal Working Working Direction Direction (ft-lb) (ft-lb) B2.5* 127 97* 149 B4* 129 B4* 129.5 104* 160 107* 164 > 72* >111 ** > 61* > 94** > 71* >109** BO* 123.5 B2* 126 75* 116 B2* 126.. 77* 119 B7* 134 79* 122 69* 107 105 13B 97

127.5 107 116 9B 127 103 135.5 122 135.5 90*

139 B9* 137.5 93* 143 77* 118

LL 400 0 0: ~ 300 UJ 0 z ~ 200 w LL w a: LL 0... z ~ 100 )> (/) I ()) .., c c( c w 0 50 c 0.35 w a: D. a z . a: <l 2x1017 aRT NOT= (40 + 01 1000 (0h Cu~ 0.08) + 5000 (% p - 0.008)) (f/10 1?] ~.~ 2 ~ 0.30 0,25

0.20% Cu

.A. WELD METAL e SHELL PLATE B 4712-2 4 6 8 ' 1018 0.15% Cu 0.10% Cu %P = 0.012 4 6 FLUENCE, n/cm2 (E > 1 MeV) 8 1019 2 Figure A-1. . Effect of fluence and Copper ar1d Phosphorus Contents on ~RT NDT for Reactor Vessel Steels LOWER LIMIT % Cu= 0.08 % p = 0.008 4 6

"' E 0 ........s Iii )> u z I UJ c.o ...I L1. z 0 a: I-UJ z 'i I * \\ t I ' " \\ ~ I I *. I I ' I I " I ' I '. ' 1019 1018 1017 t 1,,, HI t ' I ! I . i 1/4 T 3/4 T 1016._~....... ~~...._~_..~~...l-~--IL--~...L.~~...L...~--IL--~-L~~.L-~-.l.~~---~--L-~~~~--~--'._. 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 Figure A-2. YEARS Fast Neutron Fluence (E > 1 MeV) as a Function of Full Power Service life (EFPY) I Ii 1: I* \\\\,, f ' t /'* 1*. i*

-- ------- *::....--'---* ~: ~-'---- --- -------* -* . --*-----.c....::::====::::i MATERIAL PROPERTY BASIS: CONTROLLING MATERIAL RTNDT INITIAL RT NOT AFTER 7 EFPY WELD METAL (UPPER BOUND OF REGULATORY GUIDE TREND CURVE) 0°F 1/4 T, 167° F 3/4 T, 76°F A CURVES-APPLICABLE FOR HEATUP RATES UPTO 60°F/HR FOR THE SERVICE PERIOD UPTO 7 EFPY AND CONTAINS MARGINS OF 10° F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000.0 j LEAK TEST LIMIT -........ ~~ ' I

*---*--o-----.

I j j . - - ------. ~'. 2000'.0 w cc en cn w a::. ~ Q w ~ ~ Q z .. 1000~0 0.0 0.0 Figure A-3. J j, I j j HEATUP RATES UP ro.. ""'~ 60°F/HR ~ j j l-4 i.. ~ i.ir"" l'I CRITICALITY LIMIT -. ~ BASED ON INSERVICE

1' HYDROSTATIC TEST

~' TEMPERATURE (312° F) - - FOR THE SERVICE PERIOD OF UP TO 7 ~~ EFPY 100.0 200.0. 300.0 400.0' 500.0 INDICATED TEMPcRATURE (DEG. F) Salem Unit 2 ReactC'r Coolant System Heatup Limitations Applicable Up to 7 EFPY A-10

~ ~ MATERIAL PROPERTY BASIS: CONTROLLING MATERIAL WELD METAL (UPPER BOUND OF REGULATORY GUIDE T.REND CURVE) RT NOT INITIAL 0°F 1/4T,167°F RT NOT AFTER 7 EFPY 3/4 T, 76°F -CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100° F/HR FOR THE SERVICE PERIOD UPTO 7 EFPY AND CONTAINS MARGINS OF 10° F AND 60 PSIG FOR POSSIBLE iNSTRUMENT ERRORS 3000.0 _.,~l""T""l""T""l""T""'l""'I"'.,..,.....................,..,.............,...,...,...,...,..........................................._............... __............. t+-t-+-+-+-++++++++++++++-++++++-H-++-H-H-H-H-H-H-H-+-~'4--~ . j, I - 2000.CL CJ c;;

11.

w a:. 1.: tn tn w a::.

11.

c w ~ ~ - I (J c z 1000.0:- - ,--~ 0.0 100.0 200.0. 300.0 400.0 500.0 INDICATED TEMPERATURE (DEG. F) Figure A-4. Salem Unit 2 R~actor Coolant System Cooldown Limitations Applicable Up to 7 EFPY A-l1}}

ML18092A231

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