ML18353A867

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Letter Proposed Technical Specifications Change on Heatup and Cooldown Curves for Cycle 3 Operation
ML18353A867
Person / Time
Site: Palisades Entergy icon.png
Issue date: 02/22/1978
From: Hoffman D
Consumers Power Co
To: Schwencer A
Office of Nuclear Reactor Regulation
References
Download: ML18353A867 (69)
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consumers Power company

. General Offices: Z12 liv~st Mlc;:h;g~n Avenue, Ja~kson, Michigan 49201 *.Area.Coda 517 788*0.560 February 22, 1978

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Director of Nuclear Reactor Regulation Att: Mi- Albert Schwencer, Chief Operating Reactors Branch No 1 .

US Nuclear Regulatory Commission

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DOCKET 50-255- LICENSE DPR PALISADES PLANT - PROPOSED TECHNICAL

sPECI:F'ICATIONS CHANGE HEATUP AHD COiJLDGvlN CURVES Amendment* No 27 dated MaJr 5, 1977 to the Palisades Plant Provisional Operating License limit~d the effective operating period for the plant hee.tup and cool-down curves (Technical _Specification 3.1.2). This letter submits proposed Technical Specifications changes to provide heatup. and cooldown curves for Cycle 3 operation. These existing curves will allow approxiL18.tely one month of additional operation during Cycle 3 and, therefore, these proposed changes will be needed by May 1, i978. '
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David P Hoffman l Assistant Nuclear Licensing Administrator t

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CONSUMERS POWER COMPANY Docket 50-255 Request for Change t~ the Technical Specifications License DPR-20 For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in Provisional Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on October 16, 1972 for the Palisades Plant be changed as described in Section I below:

I. Change Replace Technical Specifications 3.1.2 and 3.1.3 with the attached new

  • Technical Specifications 3.1.2 and 3.1.3.

II. Discussion The Safety Evaluation by the Office of Nuclear Reactor Regulation supporting Amendment No 27 dated May 5, 1977 to the Palisades Plant Provisional License recognized that additional testing of unirradiated reactor vessel specimens would be completed in 1977. The effective period for the heatup and cooldown 6

operating curves was, therefore, reduced from a requested 3 x 10 MWdt to 2.2 x 10 6 MWdt. This request includes the* Battelle report entitled, "Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Proper-ties," dated August 25, 1977, related correspondence between Consumers Power, EPRI and Battelle dated 9/16/77, 9/20/77 and 11/22/77 and an increase of the effective operating period back to 3 x 106 MWdt. This operation will allow us to complete Cycle 3 and the evaluation of first irradiated capsule.

III. Conclusion Based on the foregoing, both the Palisades Plant Review Committee and the Safety and Audit Review Board have reviewed the proposed changes and recom-CONSUMERS n mend their approval.

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  • * ; * .Jackson County, Michigan My co~ission expires April 13, 1980.

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3.1 PRIMARY COOLANT SYSTEM (Contd) 3.1.2 Heatup and Cooldown Rates The primary coolant pressure and the system heatup and cooldown rates shall be limited in accordance with Figure 3-*l, Figure 3-2 and as follows:

a. Allowable combinations of pressure and temperature for any heatup rate shall be below and to the right of the limit lines as shown on Figure 3-1. The average heatup rate shall not exceed l00°F/h in any one-hour time period.
b. Allowable combinations of pressure and temperature for any cooldown rate shall be below and to the right of the limit lines as shown on Figure 3-2. The average cooldown rate shall not exceed l00°F/h in any one-hour time period.
c. Allowable combinations of pressure and temperature for inservice
  • testing from heatup are as shown in Figure 3-3. Those curves in-elude allowances for the temperature change ra,tes noted above.

Interpolation between limit lines for other than the noted temper-ature change rates is permitted in 3.l.2a, b or c.

d. The average heatup and cooldown rates for the pressurizer shall not exceed 200°F/h in any one-hour time period.
e. Before the radiation exposure of the reactor vessel exceeds the ex-posure for which the figures apply, Figures 3-1, 3-2, and 3-3 shall be updated in accorda.'1.ce with t.he following criteria and procedure:

(1) US Nuclear Regulatory Commission Regulatory Guide 1.99 has been used to predict the increase in transition temperature based on integrated fast neutron flux.

If measurements on the irradiated specimens show increase above

  • this curve, a new curve shall be constructed such that it is above and to the left of all applicable data points.

3-4

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    • - 3.1 3.1.2 PRIMARY COOLANT SYSTEM (Contd)

Heatup and Cooldown Rates (Contd)

{ 2) Before the end of the integrated power period for which Figures 3-1, 3-2 and 3-3 apply, the limit lines on the figures shall be updated.for a new integrated power period. The total integrated reactor thermal power from start-up to the end of the new power period shall.be converted to an equivalent integrated fast neutron exposure (E .::_ 1 MeV). Such a conversion shall be made consistent with the dosimetry evaluation of the initial surveillance program capsule to be removed before the beginning of the Cycle 3. For purposes of determining fluence at the reactor vessel beltline for the present fuel cycle, the following basis was established:

3.64 x 1019 nvt calculated at the reactor vessel.beltline for

  • ( 3) 2540 MWt for 40 years at a 80% load factor.

resulted in a correlation of 1.23 *x 10 12 This conversion has nvt per 1 MWdt.

The limit lines in Figures 3-1 through 3-3 shall be moved parallel to the temperature axis in the direction of increasing temperature a distance associated witµ the RTNDT increase during the period since the curves were last constructed. The RTNDT increase will be based upon surveillance program testing of the specimens in the ini t.ial surveillance capsule.

Basis All components in the primary coolant system are designed to withstand the effects of cyclic loads due to primary system temperature and pres-sure changes.

( l) These cyclic loads are introduced by normal unit load transients, reactor trips and start-up and shutdown operation .

  • 3-5

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  • 3.. 1 3.1.2 PRIMARY COOLANT SYSTEM (Contd)

Heatup and Cooldown Rates (Contd)

During unit start-up and shutdown, the rates of temperature and pressure changes are limited. A maximum plant heatup and cooldown rate of 100°F per hour is consistent with the design number of cycles and satisfies

. *t s f or eye l"ic opera t*ion. ( 2 )

s t ress 1 imi The reactor vessel plate and material opposite the core has been purchased to a specified Charpy V-notch test result of 30 ft-lb or greater at an NDTT of+l0°F or less. The testing of base line specimens associated with the reactor surveillance program indicates that the vessel plate has the highest RTNDT of plate, weld and HAZ specimens. The RTNDT has been determined to be 0°F. ( 3 ) An RTNDT of 0°F is used as an unirradiated value to which irradiation effects are added. In addition, this plate has been

  • 100% volumetrically inspected by ultrasonic test using.both longitudinal and shear wave methods. The remaining material in the reactor vessel, and other primary coolant system components, meets the appropriate design code requirements and specific component function and has a maximum NDTT of + 4o°F. ( 4 )

As a result of fast neutron irradiation in the region of the core, there will be an increase in the NDTT with operation. The techniques used to predict the integrated fast neutron (E > 1 MeV) fluxes of the reactor vessel are described in Section 3.3.2.6 of the FSAR and also in Amendment 13,Section II, to the FSAR.

Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition

  • 3-6

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  • 3.1 PRIMARY COOLANT SYSTEM (Contd) 3.1.2 Heat up and Cooldown Rates. (Contd}

shift for a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated ~lux magnitude. The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calculated azimuthal neutron flux variation. The maximum integrated fast neutron (E > 1 MeV) exposure of the reactor vessel is computed to be

.. 19 (5) 3.64 x 10 nvt for 40 years' operation at 2540 MWt and 80% load factor.

The predicted RTNDT shift for a given fluence and copper-phosphorus weight percent has bee~ made from a correlation for that purpose.

(6) The actual shift in RTNDT will be established periodically during plant opera-tion by testing of reactor vessel material samples which are irradiated

  • --

cumulatively by securing them near the inside wall of the reactor vessel as described in Section 4.5.3 and Figure 4-11 of the FSAR. To compensate for any increase in the NDTT caused by irradiation, limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup and cooldown.

Reference 7 provides a procedure for obtaining the allowable loadings for ferritic pressure-retaining materials in Class 1 components. This procedure is based on the principles of linear elastic fracture mechanics and involves a stress intensity factor prediction which is a lower bound of static, dynamic and crack arrest critical values. The stress intensity factor computed( 7 ) is a function of RTNDT' operating temperature, and vessel wall temperature gradients.

  • Pressure-temperature limit calculational procedures for the reactor coolant pressure boundary are defined in Reference 8 based upon Reference 7.

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  • 3.1 3.1.2 PRIMA.11.Y

. COOIJ\NT SYSTEM (Contd)

Heatup and Cooldown Rates (Contd)

The limit lines of Figures 3-1 through 3-3. consider a 54 psi *pressure allowance to account for the fact that pressure is measured in the pres-surizer rather than at the vessel belt line. In addition, for calcula-tional purposes, 5°F and 30 psi were taken as measurement error allowances for temperatur~ and pressure, respectively. By Reference 7, reactor vessel wall locations at 1/4 and 3/4 thickness are limiting. It is at these locations that the crack propagation associated with the hypothetical flaw must be arrested. At these locations, fluence attenuation and thermal gradients have bee.n evaluated. During cooldown, the 1/4 thickness location is always more limiting in that the RTNDT is higher than that at the 3/4 thickness location and thermal gradient stresses are tensile there. During

  • heatup, either the 1/4 thickness or 3/4 thickness location may be limiting depending upon heatup rate.

Figures 3-1 and 3-2 define stress limitations only from a fracture mechanic's point of view.

Other considerations may be more restrictive with respect to pressure-temperature limits. For normal operation, other inherent plant charac-teristics may limit the heatup and cooldown rates which can be achieved.

Pump parameters and pressurizer heating capacity tends to restrict both normal heatup and cooldown ~ates to less than 60°F per hour.

The revised pressure-temperature limits are applicable to reactor vessel inner wall fluences of up to 3.7 x 10 18 nvt or approximately 3.0 x io 6 MWd of thermal reactor power. The application of appropriate fluence attenua-

  • tion factors at the 1/4 and 3/4 thickness locations results in fluences 3-8
  • 3.1 3.1.2 PRIMARY COOLANT SYSTEM (Contd)

Heatup and Cooldown Rates (Contd) of 2.26 x 10 18 nvt. and 5.2 x 1017 nvt, respectively. From Reference 6, thepe are consistent with RTNDT shifts of 114°F and 55°F, respectively.

The criticality condition which defines a temperature below which the cor1= cannot be made critical (strictly based upon fracture mechanics' considerations) is RTNDT + 132°F. The most limiting wall location is at 1/4 thickness. The minimum criticality temperature (246°F) is the minimum permissible temperature for the inservice system hydrostatic pressure test. That temperature is calculated based upon 2100 psig operat'ion pressure.

The restriction of heatup and cooldown rates to l00°F/h and the mainte-nance of a pressure-temperature relationship to the right of the heatup,

  • cooldown and inservice test curves of Figures 3-1., 3-2, and 3-3, respec-tively, ensures that the requirements of References 6, 7, 8 and 9 are met.

The-core operational.limit applies only when the reactor is critical.

The criticality temperature is determined per Reference 8 and the core operational curves adhere to the requirements of Reference 9. The inser-vice test curves incorporate allowances for the thermal gradients associa-ted with the heatup curve used to attain inservice test pressure. These curves differ from heatup curves only with respect to margin for primary membrane stress. ( 7 ) For heatup rates less than 60°F/h, the hypothetical 0°F/h (isothermal heatup) at the 1/4 T location is controlling and heatup curves converge. Cooldown curves cross for various cooldown rates, thus a composite curve is drawn. Due to the shifts in RTNDT' NDTT requirements

  • associated with nonreactor vessel materials are, for all practical purposes, no longer limiting.

3-9

  • 3.1 PRIMARY COOLANT SYSTEM (Contd) 3.1. 2 Heatup and Cooldown Rates (Contd)

References (1) FSAR, Section 4.2.2

( 2) ASME Boiler and Press.ure Vessel Code,Section III, N-415 (3) Battelle Columbus Laboratories Report, "Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Properties,"

August 25, 1977 (4) FSAR, Section 4.2.4 (5) FSAR, Amendment 15

( 6) US Nuclear Regulatory Commission, Regulatory Guide 1. 99, . "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," July 1975

(7) ASME Boiler and Pressure Vessel Code,Section III, Appendix G, "Protection Against Non-Ductile Failure," 1974 Edition (8) US Atomic Energy Commission Standard Review Plan, Directorate of Licensing, Section 5.3.2, "Pressure~Temperature Limits."

(9) 10 CF Part 50, Appendix G, "Fracture Toughness Requirements,"

August 31, 1973.

3.1. 3 Minimum Conditions for Criticality

a. Except during low-power physics test, the reactor shall not be made critical if the primary coolant temperature is below 525°F.
b. In no case shall the reactor be made critical if the primary coolant temperature is below RTNDT + 132°F.
c. When the primary coolant temperature is below the minimum temperature specified in "a" above, the reactor shall be subcritical by an a.mount
  • equal to or greater than the potential reactivity insertion due to depressurization.

3-10

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  • PALISADES PLANT TECH SPEC PRIMARY COOLANT PRESSURE, PSIG PRESSURE-TEMPERATURE RELATIONSHIP DURING HEAT UP-TO 3.0X10 6 MWDT FIGURE 3-1 DATE:

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3-13

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  • 3-14

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  • 3.1 PRIMARY COOLANT SYSTEM (Contd)

Minimum Conditions for Criticality_ (Contd)

d. No more than one control rod at a time shall be exercised or withdrawn until after a steam bubble and normal water level are established in the pressurizer.
e. Primary coolant boron concentration shall not be redµced until after a steam bubble and normal water level are established in the pres-surizer.

Basis At the beginning of life of the initial fuel cycle, the moderator tempera-ture coefficient is expected to be slightly negative at operating tempera-tures with all control rods withdrawn. (l) However, the uncertainty of the calculation is such that it is possible that a slightly positive

  • coefficient could exist .

The moderator coefficient at lower temperatures will be less negative or

. (1 2) more positive than at operating temperature. ' It is, therefore, prudent to restrict the operation of the reactor when primary coolant temperatures are less than normal operating temperature (~ 525°F).

Assuming the most pessimistic rods out moderator coefficient, the maximum potential reactivity insertion that could result from depres-surizing the coolant from 2100 psia to saturation pressure at 525°F is 0.1% 6p.

During physics tests, special operating precautions will be taken. In addition, the strong negative Doppler coefficient( 3 ) and the small integrated 6p would limit the magnitude of a power excursion resulting

  • from a ~eduction of moderator density .

3-15

  • 3.1 3.1.3 PRIMARY COOLANT SYSTEM (Contd)

Minimum Conditions for Criticality (Contd)

The requirement that the reactor is not to be made critical below RTNDT +

132°F provides increased assurance that the proper relationship between primary coolant pressure and temperature will be maintained relative to the NDTT of the primary coolant system. Heatup to this temperature will be accomplished by operating the primary coolant pumps.

If the shutdown margin required by Specification 3.10.1 is maintained, there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.

Normal water level is established in the pressurizer prior to the with-drawal of control rods or the dilution of boron so as to preclude the

  • possible overpressurization of a solid primary coolant system .

References (1)

(2)

FSAR, Table 3-2 FSAR, Table 3-6 (3) FSAR, Table 3-3

  • 3-16
  • FINAL REPORT
    • , on PALISADES PRESSURE VESSEL IRRADIATION CAPSULE PROGRAM: UNIRRADIATED MECHANICAL PROPERTIES

. to

  • CONSUMERS POWER August 25, 1.977 by J. S. Perrin and E. 0. Fromm
  • BATT ELLE Columbus Laboratories 505 King Avenue Columbus, Ohio 43201

SUMNARY . . .

TABLE OF CONTENTS Page

.1

  • INTRODUCTION. . *l SPECIMEN PREPARATION. *2 EXPERIMENTAL PROCEDURES . . * *3 Drop-Weight Properties * *7 Charpy Impact Properties * . .7 Tensile Properties . ........ . 10 RESULTS AND DISCUSSION . * . . . . 12 Drop-Weight Properties 12 Charpy Impact Properties . . 18 Tensile Properties 32 CONCLUSIONS .
  • 44 REFERENCES . * .
  • 45 APPENDIX A COMPOSITIONAL ANALYSIS OF SURVEILLANCE TEST MATERIALS
  • LIST OF FIGURES Page FIGURE 1. SKETCH DF P-3 TYPE DROP HEIGHT SPECIMEN. .4 FIGURE 2. CHARPY V-NOTCH IMPACT SPECIMEN * *5 FIGURE 3. TENSILE SPECIMEN . . . . . . . .6 FIGURE 4. SKETCH OF BCL DROP WEIGHT HA.CHINE . . . . . .. . 8 FIGURE 5. LOAD TRAIN USED FOR DETERMINATION. OF TENSILE PROPERTIES . . . . . . . . 11 FIGURE 6. PHOTOGRAPHS OF DROP WEIGHT SPECIMENS TESTED AT -30 F . 14 FIGURE 7. PHOTOGRAPHS OF DROP WEIGHT SPECIMENS TESTED AT -20 F . 15 FIGURE 8. PHOTOGRAPHS OF DROP WEIGHT SPECIMENS TESTED AT -10 F 16 FIGURE 9. PHOTOGRAPHS OF DROP WEIGHT SPECIMENS TESTED AT 0 F . . 17 FIGURE 10. CHARPY IMPACT PROPERTIES FOR BASE METAL, PLATE NO. D3803-l, LONGITUDINAL ORIENTATION . . . . 23

-. FIGURE 11. CHARPY IMPACT PROPERTIES FOR BASE METAL, PLATE NO. D3803-l, TRANSVERSE ORIENTATION .

  • FIGURE 12. CHARPY IMPACT PROPERTIES FOR WELD METAL . .
  • 24

. . . . 25 FIGURE 13. CHARPY IMPACT PROPERTIES FOR HEAT AFFECTED ZONE METAL. 26 FIGURE 14. CHARPY IMPACT SPECIMEN FRACTURE SURFACES.FOR BASE METAL, PLATE NO. D3803-l, LONGITUDINAL ORIENTATION 28 FIGURE 15. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR BASE METAL, PLATE NO. D3803-l, TRANSVERSE ORIENTATION 29 FIGURE 16. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR WELD METAL. . . . . . . . . . . . . . . . . . . . . . 30 FIGURE 17. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR RAZ METAL 31 FIGURE 18. TYPICAL STRESS-STRAIN CURVE . . . . . . . 37 FIGURE 19. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 70 F, BASE METAL, LONGITUDINAL ORIENTATION. 38 FIGURE 20. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 70 F, BASE METAL, TRANSVERSE* ORIENTATION . . 38 FIGURE 21. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 70 F, WELD METAL. . . . . . * . . . . . 39 FIGURE 22. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 70 F, HAZ METAL . . * . . . . . . . . . . 39 FIGURE 23. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED

  • AT 535 F, BASE METAL, LONGITUDINAL ORIENTATION 40 FIGURE 24. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 535.F, BASE METAL, TRANSVERSE ORIENTATION . 40

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LIST OF FIGURES (Continued)

FIGURE 25. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 535 F, WELD METAL . . * . . . * * . . . * . 41

  • FIGURE 26. POSTTEST PHOTOGRAPHS OF TENSILE ~PECIMENS TESTED AT 535 F, HAZ METAL. . . . . . . . . . . . . * . 41 FIGURE 27. POSTTEST PHOTOGRAPHS OF TENSILE SPECL.'1ENS TESTED r* AT 565 F, BASE METAL, LONGITUDINAL ORIENTATION 42 FIGURE 28. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED.

AT 565. F, BASE METAL, TRANSVERSE ORIENTATION . . * . 42 FIGURE 29. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 565 F, WELD METAL . . . . . . . . * . . 43 FIGURE 30. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 565 F, HAZ METAL. . . . . . . . . * . . * . . 43

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  • LIST OF TABLES TABLE 1. CALIBRATION DATA FOR BCL HOT LABORATORY CHARPY IMPACT MACHINE *
  • 9 TABLE 2. DROP WEIGHT RESULTS.. . . ...... . .13 TABLE 3. CHARPY V-NOTCH IMPACT RESULTS FOR BASE METAL PLATE NO. D3803-l, LONGITUDINAL ORIENTATION_. .. .19 TABLE 4. CHARPY. V-NOTCH IMPACT RESULTS FOR BASE METAL PLATE NO. D3803-l, TRANSVERSE ORIENTATION . *
  • 20 TABLE 5. CHARPY V-NOTCH IMPACT RESULTS FOR WELD METAL. .21 TABLE 6. CHARPY V-NOTCH IMPACT RESULTS FOR HAZ METAL . . . . 22 TABLE 7. UPPER SHELF ENERGY AND TRAi.~SITION TE~fPERATURE FOR PALISADES . . . . . . . . - . .27 TABLE 8. PREIRRADIATION TENSILE PROPERTIES OF BASE METAL, PLATE NO. D3803~1, LONGITUDINAL ORIENTATION . . . . 33 TABLE 9. PREIRRADIATION TENSILE PROPERTIES OF BASE METAL,
  • PLATE NO. D3803-l, TRANSVERSE ORIENTATION . . . . . 34 TABLE 10. PREIRRADIATION TENSILE PROPERTIES OF WELD METAL . .35 TABLE 11. PREIRRADIATION TENSILE PROPERTIES OF HAZ METAL. .36

..

FINAL REPORT on PALISADES PRESSURE VESSEL IRRADIATION CAPSULE PROGR.A!-1: UNIRRADIATED

  • MECHANICAL PROPERTIES to CONSUMERS POWER from BATTELLE Columbus Laboratories August 25, 1977

SUMMARY

A pressure vessel surveillance prog(am is being conducted for the

  • Palisades reactor by Battelle's Columbus Laboratories. This report surmnarizes the preirradiation Charpy v.:..notch impact and.tensile properties of base metal, weld metal, and heat affected zone metal specimens for the surveillance program.

In addition, the drop weight properties of base metal specimens were also determined~ The results will be used in the future as baseline data in determining the shift in impact and tensile propert.ies for samples being irradiated in capsules in the reactor.

  • INTRODUCTION Irradiation of materials such as the pressure vessel steels used in reactors causes changes in the mechanical properties, including tensile, impact, and fracture toughness. These effects have been well documented in the technical 11.terature.Cl-l) T ensi*1 e proper t"ies s h ow a d ecrease o f b o th uni"f orm e 1 onga t"ion an d reduction in area accompanied by an increase in yield strength and ultimate tensile strength with increasing neutron exposure. The impact properties as determined by the Charpy V-notch impact test show a shift of the complete Charpy energy-temperature curve reflecting the brittle-failure temperature range extending to higher temperatures. In addition, the upper shelf of the Charpy curve shows a drop in energy level.

(1) References at end of text.

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I 2

  • Commercial nuclear power reactors are put into operation with reactor pressure vessel surveillance programs. The purpose of the surveillance program associated with a reactor is to monitor the changes in mechanical properties as a function of neutron exposure. The surveillance program includes a determination of both the preirradiation base line mechanical properties and periodic determi-nations of the irradiated mechanical properties. The materials included in the present surveillance program are base metal, weld metal, and heat~affected-zone (HAZ) metal from the actual components used in fabricating the vessel.

The irradiated mechanical properties are determined periodically by testing specimens from surveillance capsules. These capsules typically contain neutron flux monitors, Charpy impact specimens, and tensile specimens. Capsules are located between the inner wall of the pressure vessel and the reactor core, so the specimens receive an accelerated neutron exposure. Capsules are periodically removed, and sent to a hot laboratory for disassembly and specimen evaluation.

The Palisades reactor pressure vessel surveillance program is described in. a report issued by Combustion Engineering (S), and' is, based on ASTM El85-73,

  • "Surveillance Tests on Structural Materials in Nuclear Reactors". C9 ) At the time of initial operation of the reactor, the pressure-temperature operating curves were based on the NDT temperature of the limiting materials. During the life of the reactor, the operating curves are to be revised to account for the shift in mechanical properties determined by Charpy impact tests.

The present report describes the preirradiation base line tensile and Charpy impact properties of the three materials being used in the surveillance capsule program. The mechanical properties of the Charpy and tensile specimens were determined following the general recommendations of ASTM El85-73. In addition, the NDT temperature for the base metal was established from drop weight specimens tested in accordance with ASTH E208-69, "Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels". (lO)

SPECIMEN PREPARATION The base metal of the reactor pressure vessel is SA-302 Grade B.

Mechanical property specimens were prepared from actual vessel plate in accordance with CE specifications and provided to Consumers Power. (S) All specimens were

  • made from flat slabs taken parallel to the plate surfaces and at the 1/4 plate thickness.*

......

3 Longitudinal base metal Charpy, tensile, and drop weight specimens were oriented with the major axis of the specimen parallel to the principal

  • rolling direction of the plate and parallel to the su~face of the plate. Trans-verse base metal Charpy and tensile specimens were orientated with the major axis of the specimen perpendicular to the principal rolling direction and parallel to the surface of the plate. Longitudinal weld metal Charpy and tensile specimens were oriented with the major axis of the specimen parallel to the direction of the weld and parallel to the surface of the weld. Transverse weld metal and heat-affected-zone specimens were oriented with the major axis of the specimen perpendicular to the direction of the weld and parallel to the surface of the weld.

The axis of the notch of all base metal and weld metal Charpy impact specimens was perpendicular to the surface of the plate or weld and the axis of the notch of *all heat-affected-zone Charpy impact specimens was parallel to the surface of the plate.

Compositional analyses of the materials used in fabrication of the specimens are tabulated in Appendix A.

The drop weight specimen used for the program is shown in Figure 1.

.

It is based upon the design recommended in ASTM E208-69 (10) for the P-3 type

  • specimen. The Charpy impact specimen is shown in Figure 2 and is the standard specimen recolllillended in ASTM E23-72. ( 9 ) The tensile specimen design is shown in Figure 3. It has a nominal 0.250-in. gage diameter and a nominal 1.00-in. gage length.

EXPERIMENTAL PROCEDURES This section describes the experimental procedures used in the determination of the drop weight, Charpy impact and tensile .properties. All testing was conducted at Battelle's Columbus Laboratories according to applicable ASTM procedures. The data for the program are recorded in BCL Laboratory Record Book No. 32899.

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  • FIGURE 1. SKETCH OF P-3 TYPE DROP WEIGHT SPECIMEN

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OF .REDUCED SECTION FIGURE 3. TENSILE SPECIMEN

7 Drop.,..Weight Properties The drop-weight tests were conducted using the P-3 type specimen shown

  • in Figure 1. The specimens were fabricated from base longitudinal metal and were tested in the BCL Drop Weight Machine according to procedures based upon ASTM Method E208-69. (lO) The verification of the machine was performed following the requirements of ASTM Method E208.:.69.
    • .The drop weight machine is shown schematically in *Figure 4. It consists of a vertically guided, free falling weighted tup and a rigidly supportedanvil which provides for the loading of a rectangular plate specimen as a simple beam under the falling weight. The rails are held in a vertical position in a fixed relationship to the base. The rails guide the weighted tup so that it strikes the specimen at the proper location. The impact energy of the weighted tup was 300 ft-lb in all tests.

The specimen was iII1II1ersed in an agitated low temperature b.ath consisting

  • of methyl alcohol and dry ice for a minimum of 45 minutes prior to testing. The temperature of the bath was held to +/-2 F during the 45-minute soak time. The specimen was then removed from the bath, placed on the anvil, and tested within 20 seconds. The resultant "break or no break" performance was then noted.

Charpy Impact Properties The impact properties were determined using a standard 240 ft-lb Wiedema.nn Baldwin impact machine in accordance with the recommendations of ASTM Method E-23-72. (ll) The machine was verified according to the applicable sections of this standard. In addition, the proof test of the machine was performed using standard Charpy V-notch specimens purchased from the U.S. Army Materials Research Agency. The results of the standard specimens are given in Table 1.

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 0 ft-lb; (b) without resetting the pointer, again releasing the pendulum from the 240 ft-lb upright position and permitting i t to swing 11 half cycles. After the pendulum starts its 11th half 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 0.4 percent (0.96 ft-lb) of the 240 ft-lb capacity.

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  • TUP
  • FIGURE 4. SKETCH OF BCL DROP WEIGHT MACHINE

9

  • TABLE 1. CALIBRATION DATA FOR BCL HOT LABORATORY CHARPY Il1PACT KA.CHINE Average Standarta)

BCL Energy, Energy, Variation Group ft-lb ft-lb Actual Allowed Low Energy 12.2 12 . 4 -0.2 ft-lb +/-1 ft-lb Medium Energy 50.6 52.9 -4.3% +/-5 %

High Energy 69.4 71.6 -3.1% +/-5%

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

......-_ .J -

  • 10 The ASTH recommendations for specimen temperature control were followed. The low temperature bath consisted of agitated methyl alcohol cooled with additions of liquid nitrogen. The container was a Dewar flask which con-tained a grid to keep the specimens at least 1 in. from the bottom with a minimum of 1 in. of liquid over the specimens. The Charpy specimens were held at tempera-ture for a minimum of at least the ASTM recommended time. The tests above room temperature were conducted in a similar manner except that a temperature controlled oil bath was used.

The specimens were transferred from the temperature bath to the anvil of the impact machine by means of tongs that had also been brought to temperature in the bath. The specimens were removed from the bath and impacted in less than 5 sec . The energy required to break each specimen 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 comparison of the specimen fracture surface to an ASTM fracture appearance chart. (ll)

  • Tensile Properties The design of the tensile specimens is shown in Figure 3. The tensile tests were conducted on a screw-driven Instron testing machine having a 20,000 lb capacity. Crosshead speeds of 0.005 and 0.05 in. per min were used. A strain gage extensometer was attached to the specimen gage length. The strain gage unit senses the differential movement of two extensometer extension 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 5 shows the extensometer extension arms and strain gage assembly used for tensile 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 .

PULL ROD---------:1 PULL RQ11--------'"!'t EXTENSOMETER EXTENSION ARM-~

  • CUP GAGE-----;

4752 FIGURE 5. LOAD TRAIN USED FOR DETERMINATION OF TENS ILE PROPERTIES Specimen is located under extensometer knife edges in center of photograph. Clip-on

  • strain gage attached to extensometer anns is shown in lower region of photograph.
  • 12 The specimens were pulled in a load-elongation mode at a crosshead speed of 0.005 in. per min until the vicinity of maximum load. The run s were then finished in a load-time mode at a crosshead speed of 0.05 in. per min.

The tensile specimens were tested at room t emperature, 535 F, and 565 F.

The elevated temperature tensile tests were conducted using a split test furnace.

The specimens were held at temperature for 20 minutes befo re testing to stabi.l ize the temperature. Temperature was monitored using two

  • Chromel-Alumel -thermocouples directly attached within the gage section of the specimen. Temperature was controlled within +/-5 F of the test temperature throughout the test period.

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 of the necked down area using a blade micrometer.

RESULTS AND DISCUSSION

  • Drop Weight Properties The results of the drop weight tests for the Pa lisades specimens are listed in Table 2 . In the drop weight test, the NDT temperature is defined as the highest temperature at which a specimen breaks, with a pair of specimens exhibiting "no break" behavior at a temperature 10 F higher. As indicated in the table, duplicate drop weight tests were conducted at 10 F intervals from -3 0 F to 0 F.

Based on the "break, no break" behavior, the NDT for the base longitudinal metal was

-10 F. This is 20 F higher than that report ed by the reactor vendor for this material. (S) The table of drop weight data from the present pro gram shows that at -20 F, one specimen exhibited "break" and one "no break" performance. If both had exhibited "no break" performance, then the NDT tempera ture would have been

-30 F as report ed by the vendor , and tests at -10 F and 0 F would not have been performed.

Post-test photographs of the drop weight specime ns are given in Figures 6 through 9 . The "break" or "no break" performance is specified for each specimen .

  • 13
  • TABLE 2. DROP WEIGHT RESULTS Specimen Test Resul{s of Identification Temperature, F Test a) lCl -30 Break 1C2 -30 Break lCS -20 No Break 1C7 -20 Break lCA -10 Break lCB -10 No Break 1C3 0 No Break lCC 0 No Break

(a) Break - Fracture to one or both edges of tension surface .

No Break - Visible crack in crack starter weld bead but not propagated to either edge of tension sur face.

..

  • lCl

"Br e ak" 1C2

" Br e ak" FIGURE 6 . PHOTOGRAPHS OF DROP WE I GHT SPECIMENS TESTED AT - 30F

  • 1C5 "No Break"
  • 1C7 "Break" FIGURE 7. PHOTOGRAPHS OF DROP WEIGHT SPEC I MEN S TESTED AT - 20 F

,

  • lCA "Break"
  • lCB "No Break" FIGURE 8. PHOTOGRAPHS OF DROP WEIGHT SPECIMENS TESTED AT - lOF

'

  • 1C3 "No Break"
  • lCC "No Break" FIGURE 9. PHOTOGRAPHS OF DROP WEIGHT SPEC lliENS TES TED AT OF
  • 18 Charpy Impact Properties The impact properties determined as a function of temperature are listed in Tables 3 through 6. 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 of the deformation produced by the striking edge of the impact machine hammer when 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 3 through 6 are graphically shown in Figures 10 through 13. These figures show the change in impact properties as a function of temperature. Of particular interest 'is the temperature corresponding to the impact energies of 30 and 50 ft-lbs. The energy level of the upper shelf is also of interest. If the upper shelf energy is relatively low (e.g. 50 ft-lb

  • or lower), the poss{bility of failure by low energy ductile tearing is greater.

In terms of. fracture mechanics, a lower upper shelf is accompanied by low values of Kic' the plane strain fracture toughness.

Table 7 summarizes the 30 and 50 ft-lb transition temperatures and the upper shelf energies for the reactor. The 50 ft-lb transition temperature ranges from ~so F (weld metal) to 55 F (base metal-transverse). The upper shelf energy level levels are all above 100 ft-lb.

Figures 14 through 17 show the fracture surfaces of the Charpy specimens.

Figure 14, as an example, shows how the fracture surface changes as the test temperature is increased for the base metal-longitudinal specimens. The -100 F specimen (147) shows an* almost flat fracture surface with essentially 0 percent shear fracture appearance. This specimen absorbed only 3.0 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 quite small, and was measured as being only 4.5 mils. As the test temperature is increased, specimens show an increasing amount of shear fracture appearance. The +150 F specimen (142) fracture surface is typical of the type

  • seen at the higher temperature end of the Charpy transition curve.

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19

  • TABLE 3. CHARPY V-NOTCH IllPACT RESULTS FOR BASE METAL PLATE NO.* D3803-l, LONGITUDINAL ORIENTATION i,. Impact Lateral Fracture Specimen Temperature, Energy, Expansion, Appearance, No. OF ft-lb mils Percent Shear 14E -150 2.0 2.0 0 147 -100 3.0 4.5 0 145 -33 10.0 12.0 1 14D -15 15.5 19.0 5 14C +5 34.0 33.5 10 143 +21 49.0 45.0 15 171 +49 77 .5 63.0 25 172 +49. 67.0 54.5 20
  • 173 +49 90.0 67.0 25 146 +50 69.0 56.0 25 141 +72 94.0 75.0 50 14A +110 129.5 88.0 80 142 +150 135.5 93.5 100 14B +225 162.5 85.5 100 144 +294 143.0 94.0 100 14J +360 178.0 78.0 100
  • e "f.:_. "'

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--* 20 TABLE 4. CHARPY V-NOTCH IMPACT RESULTS FOR. BASE METAL PLATE NO. D3803-l, 'IRANSVERSE ORIENTATION

[.

Impact Lateral Fracture Specimen Temperature, Energy, Expansion, Appearance, No. OF ft-lb mils Percent Shear 23P -150 2.0 2.5 0 21E -100 4.0 4.0 0 21B -33 11.0 13.5 2 23M -15 12.0 16.0 5 23L +s 41.5 38.5 10 217 +21 2 7 .o 29.0 15 21D +so 46.0 44.0 . 20 215 +72 60.0 53.5 30

  • 23T * +90 68.5 58.0 50 23J +110 94.0 75.0 80 216 +150 114.0 79.0 100 23K +225 107 .o 79.0 100 21A +295 92.0 75.5 100 252 +296 102.0 81.5 100 23U +360 93.0 78.0 100

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e e 21

  • TABLE 5. CHARPY V-NO'ICH Il!PACT RESULTS FOR WELD METAL Impact Lateral Fracture Specimen Temperature, Energy, Expansion,. Appearance, No. OF ft-lb mils Percent Shear 36T -170 4.0 5.0 0 36J -150 7.0 7.0 0 36P -135 14.5 14.5 2 367 -100 8.5 11.0 1 36H -100 31.0 28.0 5 36L -85 35.0 33.5 10 36E . -75 47.5 41.0 15
  • 36D -50 41.0 39.0 20 365 .,.33 56.0 52.0 30 36C -5 87.0 75.0 60 363 +20' 86.0 77 .o 80 366 +50 92.0 79.0 80 361 +72 96 .o 85.0 90 36A +110 117 .5 94.0 100 362 +150 112 .o 88.5 100 36B +225 127~5 92~0 100 364 +296 111.0 87.0 100 35K +360 120.5. 91.5 100

~-*

22 TABLE 6. CHARPY V-NOTCH INPACT RESULTS FOR HAZ :METAL Impact . Lateral Fracture Specimen Temperature, Energy, Expansicn, . Appearance,.

No. oy ft-lb mils *Percent, Shear 47Y -170 5.0 4.0 0 471 -150 5.0 3.0 0 43A -145 6.0 5.0 0 475 -137 6.0 3.5 2 474 -135 13.0 10.0 1 473 -120 15.0 11.0 5 .

476 -109 . 20.0 15.0 5 477 -101 25.0 18.5 10.

467 -100 64.0 43.0 is

  • 41A -75 13 .0 13 .5 5 465 -34 76.5 56.0 40 463 +20 77 .o 45.5 so 466 +50 94.0 71.0 85 461 +72 112 .0 76.5 80 46A +110 90.0 74.0 100 .

462 +151 114.0 81.0 100 46C +225 120.0 86.0 100 464 +296 138 .5 87.0 100 472 +360 115.0 75.0 100

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  • FIGURE 11. CHARPY IMPACT PROPERTIES FOR BASE METAL, PLATE NO. D3803-l, TRANSVERSE ORIENTATION
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  • 27
  • TABLE 7. UPPER SHELF ENERGY AND 'IRANSITION TEMPERATURE FOR PALISADES 30 ft- lb 50 ft-lb .Upper Shelf I , Transition Transition Energy, Material Temperature, F Temperature, F ft-lb Base (Longitudinal) 0 +20 165 Base (Transverse) +25. +55 105 Weld Metal -85 .-50 120 HAZ Metal -90 -65 125
  • Specimen No. 14E 147 145 14D 14C 143

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Test Temperature, op -150 -100 -33 -15 +5 +21 Specimen No. 171 172 173 146 141

  • Test Temperature, °F +49 +49 +49 +so +72 Specimen No . 14A 142 14B 144 14J Test Tempera tu re, °F +llO +150 +225 +294 +360 FIGURE 14. CHARPY Il1PACT SPECil1EN FRACTURE SURFACES FOR BASE METAL, PLATE NO. D3803 - l , LONGITUDINAL ORIENTATION
  • Specimen No. 23P 21E 21B 23M 23L

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Test Temperature, °F - 150 -100 -33 - 15 +5 Specimen No . 217 21D 215 23T 23J Te s t

  • Temperature, °F +21 +50 +72 +90 +llO Specimen No*. 216 23K 21A 252 23U Test Temperature, °F +150 +225 +295 +296 +360
  • FIGURE 15. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR BASE METAL, PLATE NO. D3803 - l, TRANSVERSE ORIENTATION

~- '

  • Specimen No . 36T 36J 36P 367 36M 36L Test Temperature , °F - 170 - 150 - 135 - 100 -100 -85 Specimen No. 36E 36D 365 36C 363 366
  • Test Temperature, °F - 75 - 50 -33 -5 +20 +50 Specimen No. 361 36A 362 36B 364 35K Test Temperature, °F +72 +llO +150 +225 +296 +360
  • FIGURE 16. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR WELD METAL
  • Specimen No. 47Y 471 43A 475 474 473 476 Test Temperature, °F -170 - 150 -145 -137 -135 -120 -109 Specimen No. 477 467 41A 465 463 466 Test
  • Temperature, °F -101 -100 -75 - 34 +20 +so Specimen No . 461 46A 462 46C 464 472 Test Temperature, °F +72 +110 +151 +225 +296 +360 FIGURE 17. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR HAZ METAL

.. .

-* 32 The fracture surface shows large shear lips with a 100% shear fracture appearance.

The specimen absorbed a relatively large amount of energy, 135 ft-lb during impact.

The substantial amount of plastic deformation occurring during this test is reflected in the large value of 93.5 mils lateral expansion.

The NDT temperature as found from the drop weight test for the base longitudinal material was -:-10 F. The reference temperature, RTNDT' was established (13) according to paragraph NB-2331 of the ASME Code. .The RTNDT is established as

~--

follows: at a temperature not greater than NDT +60 F, each of three Charpy V-notch specim~ns tested must exhibit at least 35 mils lateral expansion and not less than 50 ft-lb of absorbed energy. When these requirements are met the NDT is the reference temperature.

Triplicate base longitudinal specimens were tested at 49 F to establish the RTNDT" As seen from Table 3, all three* specimens (171, 172, and 173) met the above criteria. Consequently, the NDT temperature of -10 F is also the reference

. temperature RTNDT"

  • Tensile Properties The preirradiation tensile* properties determined as a function of temperature are listed in Tables 8 through 11. The tables list the test temperature, 0.2 percent offset yield strength, ultimate tensile strength, uniform elongation, total elongation, and reduction in area. A typical tensile test curve is shown in Figure 18; the particular curve shown is for base metal specimen lJC tested at 565 F.

Posttest photographs of the tensile specimens are shown in Figures 19 thr?ugh 39.

Tensile tests were run at room temperature, 535 and 565 F. The higher ten1perature tests exhibited a decrease in the 0.2 percent offset yield strength and a decrease in the ultimate tensile strength for each material. In general, ductility values (as determined by total elongation and reduction in area) were lower at 535 and 565 F than at 75 F for each material.

The three RAZ specimens tested at 75 F all fractured near one end of the gage length. All other specimens including the high temperature RAZ specimens, fractured at or close to the center of the gage length~ There is no obvious explanation for the behavior observed for the three room temperature HAZ specimens .

.*

  • 1 **'

'\ ...:r) 33

  • TABLE 8. PREIRRADIA TION TENS ILE PROPER TIES OF BASE METAL, PLATE NO. D3803-l, LONGIWDINAL, ORIENTAT!ON 0.2 Percent Offset Ultimate Reduction Yield Tensile in Specimen Temp, Strength, Strength, Elongation 2 Percent Area,

.i No. F psi psi Uniform Total Percent lEl . 70 64,270 85,700 15.4 30.7 72 .3 1E2 70 63,920 85 ,200 15.6 31.2 72. 7

  • 1E3 70 63,170 85, 720 16.3 30.9 71.3 lIM 535* 57,170 82, 720 12 .3 23.3 64.5 IDP 535 57,580 82 ,470 11.8 23.1 64.8 IDI 535 57,490 82'180 11.8 23.1 64.6 IJA 565 58,840 84,920 13.0 24.3 62.6 IJB 565 57,860 84;350 13. 7 25.2 68.6 IJC 565 58,340 84,510 14.2 26.2 67.7
  • I .I *

'\.- .,

34 TABLE 9. PREIRRADIATION TENSILE PROPERTIES OF BASE HETAL, PLATE NO. D3803-l, TRANSVERSE ORIENTATION .

0.2 Percent Offset Ultimate Reduction Yield Tensile in Specimen Temp,. Strength, Strength, Elongation 1 Percent Area, No. F psi psi Unifonn Total Percent 2Dl 70 65,180 86,640 16.0 29.6 68.2 2D2 70 65,350 86,990 15.3 29.0 68.2

  • 2D3 2DE 2DJ 2DK 2DA
  • 70 535 535 535 565 65,040 57,740 57 ,630 57 ,630 57 ,220 86,990 81,840 82 ,280 82 ,380 84,210 14.7 12.2 11.8 11.6 14.7 28.5 25.8 16.3 17.8 25.7 68.4 63.8 33.9 42 .4 63.1 2DB 565 56,920 83,800 13.9 21.8 42.7 2DC 565 58,360 83,780 13. 9 23.9 57.5
  • 35
  • TABLE 10. PREIR.RADIATION TENSILE PROPERTIES OF HELD METAL 0.2 Percent Offset Ultimate Reduction Yield Tensile in Specimen Temp, Strength, Strength, Elonga tion 2 Percent Area, No. F ESi ESi Uniform Total Percent 3El 70 64,980 82,190 17.2 32.0 68.9 3E2 70 63,490 81,540 17.2 32.3 70.3 3E3 70 64,330 82,010 17.7 32.6 71.0 3E5 3E6 3E7 3EA 3EB 535 535 535 565 64,630 63' 100 63,320 59,640 86,470 83,900 83,420 82,760 13.5 12.8 11.9
  • 13.7 21.6 21.9 21.3 24.7 56.4 57.0 53.8 63.8
  • 565 61,060 84,220. 13.7 21.9 49.3 3EC 565 60,800 84' 960 13.2 22.7 54.8
  • - \ *' .
  • 36 TABLE 11. PREIRRADIATION TENSILE PROPERTIES OF HAZ HET..A.L 0.2 Percent Offset Ultimate Reduction Yield Tensile in Specimen Temp, Strength, Strength, Elonga tion 2 Percent Area, No. F psi psi Uniform Total Percent -

4El 70 63,180 84,380 15.3 28.5 66.2 4E2 . 70 63,860 84,480 15.7 30.2 69.5 4E3 70 63,95~ 84,450 15.4 30.1 71.3

  • 4E4 535 59,060 81,590 12.5 22.9 61.8 4E5 535 58, 110 81,970 12.5 23.0 66.6 4E6 535 57,730 82,180 12.4 21.3 64.7 4Jl 565 56,790 82,900 14.4 24.9 63.7 4J2 565 55,260 81,930 13.9 25.4 65.3 4J3 565 55,090 82,280 14.2 25.5 6.6 .2

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  • 0 5 10 15 20 25 30 PERCENT ELONGATION FIGURE 18. TYPICAL STRESS-STRAIN CURVE Curve shown is for base metal specimen lJC, longitudinal orientation, tested at 565 F.

...... '

  • Specimen lEl Specimen 1E3 FIGURE 19. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 70 F, BASE METAL, LONGI~DINAL ORIENTATION
  • Specime n 2Dl Specimen 2D2 Specimen 2D3 FIGURE 20 . POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 70 F, BASE METAL, TRANSVERSE ORIENTATION

..

. '

  • Specimen 3El Specimen 3E2 Specimen 3E3 FIGURE 21 . POSTTEST PHOTOGRAPHS OF TENSILE SPECil1ENS TESTED AT 70 F , WELD METAL
  • Specimen 4El Specimen 4E2 Specimen 4E3 FIGURE 22 . POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 70 F, HAZ
  • METAL

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..- '

  • Specimen llli Specimen lDP Specimen lDT FIGURE 23. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 535 F, BASE METAL, LONGITUDINAL ORIENTATION
  • Specimen 2DE Specimen 2DJ Specimen 2DK FIGURE 24 . POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 535 F, BASE METAL, 1RANSVERSE ORIENTATION

..

' .

Specimen 3E5

  • Specimen 3E6 Specimen 3E7 FIGURE 25. POSTTEST PHOTOGRAPHS OF TENSILE SPECil1ENS TESTED AT 535 F, WELD METAL Specimen 4E4
  • Specimen 4E5
  • 'I' * * - * - - '

"'--*,..,...,... iJit:I-'.,,,...*. "!"',,..............

~'; *'""!" i -~~, '*

. ""~~ ~~-,~4..,..::,. ,s,;;\;....~ _.,,.....,

.,,~-~*t ~-:~.,~Y'lo~~~-.....,...., __ ".

, '*- , ~ * ' . * - ~ ' -~ ' - e * *

  • Specimen 4E6 FIGURE 26. POSTTEST PHOTOGRAPHS OF TENSILE SPECil1ENS TESTED AT 535 F, HAZ METAL
  • Specimen lJA Specimen lJB Specimen lJC FIGURE 27. POSTTEST PHOTOC1\APHS OF TENSILE SPECIMENS TESTED AT 565 F, BASE METAL, LONGITUDINAL ORIENTATION
  • Specimen 2DA Specimen 2DB Specimen 2DC FIGURE 28. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 565 F, BASE METAL, TRANSVERSE ORIENTATION

. .,,

'

. .

,

.

Specimen 3EA

  • Specimen 3EB

-*~." .r. ~~ '*.. '*:* *, .~ : . * ,.~- *": .-,: ... *-:*' ,*:.*~

~~-=f~*;~3~~'."-: Specimen 3EC FIGURE 29. POSTTEST PHOTOGRAPHS OF TENSILE SPECIMENS TESTED AT 565 F, WELD METAL Specimen 4Jl

  • Specimen 4J2

- ~ .... - ' . .

- ~.~-::.:~;;c:: * ""I Specimen 4J3

. '\. . ~ *~

FIGURE 3 0. POS TTES T PHO TO GRAPHS OF TENS ILE SPECIMENS TESTED AT 565 F, HAZ METAL

  • 44 CONCLUSIONS The drop weight properties of base metal longitudinal orientation and the Charpy impact and tensile properties of base metal longitudinal and transverse orientations have been determined.. The data generated fell in the general range of values to be expected for these materials.

The NDT temperature for the base m~tal (longitudinal orientation) as i **** established from the drop weight test was -10 F. The corresponding reference temperature RTNDT for the same material as determined by Charpy tests was also

-10 F. The upper shelf energy levels range from 105 ft-lb for the base metal transverse orientation to 165 ft-lb for the base metal longitudinal orientation material *

  • rj, .,

~J ........

45

  • REFERENCES (1) Reuther, T. G., and Zwilsky, K. M., 11 The Effects of Neutron Irradiation on the Toughness and Ductility of Steels", in Proceedings of _Toward Improved Ductility and Toughness Symposium, published by Iron and $teel Institute of Japan (October 1971), pp 239-319.

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

(3) Berggren, R. G., "Critical Factors in the Interpretation of Radiation- Effects on the Mechanical Properties of Structural Metals", Welding Research Council Bulletin, 87, 1 (1963).

(4) Witt, F. J., "Heavy-Section Steel Technology Program Semiannual Progress Report for Period Ending February 29, 1972", ORNL Report No. 4816*

(October 1972).

(5) Hawthorne, .J. R., "Radiation Effec;:ts Information Generated on the ASTM Reference Correlation-Monitor Steels", American Society for Testing and Materials Data Series Publication DS54 (1974).

(6) Steele, L. E., and Serpan, C. z., "Neutron Embrittlem.ent of Pressure Vessel Steels - A Brief Review, Analysis of Reactor Vessel Radiation Effects Surveillance Programs, American .Society for Testing and Materials Special Technical Publication 481 (1969), pp 47-102.

(7) Integrity of Reactor Vessels for Light-Water Power Reactors, Report by the USAEC Advisory Cormnittee on Reactor Safeguar~s (January 1974).

(8) Groeschel, R. C., Sunnnary Report -on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Palisades Reactor Vessel Materials, CE Report No. P-NLl.~-019, (April 1, 1971).

(9) ASlli Designation El85-73, 11 Surveillance Tests on Structural Materials in Nuclear REactors", Book of AS'Il1 Standards, Part 10 (1976), pp 314-320.

(10) AS'JM Designation E208-69, 11Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels", Book of ASTM Standards, Part 10 (1976), pp 330-349.

(11) AS'Il1 Designation E23- 72, Notched Bar Impact Tes ting of Metallic Materials",

Book of ASlli Standards, Part 10 (1976), pp 197-213.

(12) AS'Il1 Designation A370-75, "Mechanical Testing of Steel Products", Book of ASTM Standards, Part 10 (1976), pp 28-79.

(13) ASME Boiler and Pressure Vessel Code,Section III, Division, Rules for Construction of Nuclear Power Plant Components, Subsection NB, Class 1 Components, 1974 Edition.

  • r.

!i APPENDIX A

  • COMPOSITIONAL ANALYSIS OF SURVEILLANCE TEST MATERIALS
  • A-1 APPENDIX A
  • COMPOSITIONAL ANALYSIS OF SURVEILLA.t~CE TEST HATERIALS The sample chemical analyses of the surveillance test materials for the three plates and two welds that make up the surveillance program as reported by Combustion Engineering(B) ~re given in Table A-1.

The base metal test material was fabricated from Plate No. D3803~1.

The weld metal test material was fabricated by welding together intermediate shell Plate Nos. 03803-1 and 03803-2. The heat-affected-zone test material was fabricated by welding together interinediate shell Plate Nos 03803-2 and 03803-3.

  • **
  • A-2 TABLE A-1. SAMPLE CHEMICAL ANALYSIS OF SURVEILLANCE TEST MATERIALS

'

..

D-3803-3/ . D-3803-2/

D-3803-l(a) D-3803-2 D-3803-3 D-3803-2(b) D-3803-1 (c)

Elements Plate Plate Plate Weld @ 2 in. Weld @ 2 in.

Root Face Root Face Si .23 .32 .24 .24 .25 .25 .22 s .019 .021 .020 .009 .010 .010 .010 p .011 .012 .010 .Oll .012 .011 .011

  • Mn 1.55 1.43 1.56 1.08 1.03 1.01 1.02 c .22 .23 .21 .098 .080 .088 .086 Cr .13 .42 .13 .05 .04 .OS .03 Ni .53 .SS .S3 .43 1.28 .63 1.27 Mo .58 .58 .S9 .S4 .S3 .SS .S2 Al .037 .022 .037 Nil Nil Nil Nil v .003 .003 .003 Nil Nil Nil Nil Cu .25 .25 .2S .2S .20 .26 .22 (a) Used to fabricate base metal specimens.

(b) Used to fabricate HAZ metal specimens.

1 * (c) Used to fabricate weld metal specimens .

  • ,}
  • *

,

  • Area Code 517 788*0550 September 16, 1977 Mr Ted Marston EPRI.

Nuclear Syste::J.s and Materials Dept 3lil2 Hillview Ave

?O :Box 10412 Palo Alto, CA 94303 I am enclosing the two docuc:ents I spoke of in ou~ conver.sation of last Thursday. These documents are the Eattellc re:?ort on our unirradiatcd.

baseline sur1eillance specimens and the US AEC Regulatory Standard. Review Plan 5.3.2. The Standard Review Plan is supplemented by Branch TecnnL~al Position ?*ITEB 5-2 which prescribes certain fracture toughness requirements for older plants~

Based upon 3attelle test data, the i'TDT temperature for the longi tudir.ally

. t e d .oase me t fl.L- specimens orien . . -1.

is . 0°.....

r. ,.,,.

ine re.1.erence 'terr::perature is ... h e

.C' * . ' * ...

. same based trpon Charpy V-ifotch. results. - Eowever, the Consumers P~wer Co baseihie speci::-:en :pactage from CE. does not contain tro.nsverse drc:p weie;:ht specimen~. This i.r:rplies that the transverse Charpy V-iiotch data a1*e controlline and serve to define the base metal reference temperature. Fron the MTEi3 5-2 section A.l.1(1), the.reference terr:perature for 0ur s:peci!!:ens would e.ppear to default to the trcnsverse Charpy V-Hotch 30 ft lb valve of 25°F.

It is our judc;:ent that such a value is unreasonable. We would lil-:e to be advised of any tcchr.ical basis which mi[ht exist for adopting a less restrictive reference temperature.

Any information which you could provide would be very much appreciated.

Sr Enr..ineer

  • Operatine Services Dept JENY. 39-77
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