Text

Analyses of Capsule TE1-B,Toledo Edison Co,Davis-Besse Nuclear Power Station,Unit 1 Reactor Vessel Matl Surveillance Program
	ML20094J664
Person / Time
Site:	Davis Besse
Issue date:	05/31/1984
From:	Collins L, Ewing J, Lowe A; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML20094J656	List: ML20094J654; ML20094J664;
References
	BAW-1834, TAC-56041, NUDOCS 8408140413
	Download: ML20094J664 (75)
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BAW-1834 May 1984 s

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l ANALYSES OF CAPSULE TEl-B THE TOLEDO EDIS0N COMPANY DAVIS-BESSE NUCLEAR POWER STATION UNIT 1

-- Reactor Vessel Material Surveillance Program --

Babcock &Wilcox ls8is8sPJs88jjg a ucoemote company

BAW-1834 May 1984 i

ANALYSES OF CAPSULE TEl-B THE TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATION UNIT 1

-- Reactor Vessel Material Surveillance Program --

by A. L. Lowe, Jr., PE L. L. Collins J. W. Ewing W. A. Pavinich W. L. Redd l

B&W Contract No. 582-7168 BABCOCK & WILCOX Utility Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Batacock&WUHoom a McDermott company

SUMMARY

This report describes the results of the examination of the second capsule of the Toledo Edison Company!s Davis-Besse Nuclear Power Station Unit I re-actor vessel surveillance program.

The capsule was removed and examined at the end of the third fuel cycle.

The objective of the program is to moni-tor the effects of neutron irradiation on the tensile and fracture tough-ness properties of the reactor vessel materials by the testing and evalua-tion of tension, Charpy impact, and compact fracture toughness specimens.

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The program was designed in accordance with the requirements of Appendix H to 10 CFR 50 and ASTM specification E185-73.

18 n/cm2 (E > 1.0 The capsule received an average fast fluence of 5.92 x 10 MeV) and the predicted fast fluence for the reactor vessel T/4 location at the end of the third cycle is 8.4 x 1017 n/cm2 (E > MeV). Based on the cal-culated fast flux at the vessel wall, an 80% load factor, and the planned fuel management, the projected fast fluence that the Davis-Besse Unit i re-actor pressure vessel will receive in 40 calendar years' of operation is 1.6 x 1019 n/cm2 (E > 1 MeV).

The results of the tension tests indicated that the materials exhibited nor-mal behavior relative to neutron fluence exposure.

The Charpy impact data results exhibited the characteristic behavior of shift to higher tempera-ture for both the 30 and 50 ft-lb transition temperatures as a result of neutron fluence damage and a decrease in upper shelf energy. These results demonstrated that the current technignes used for predicting the change in both the increase in the RTNDT and the decrease in upper shelf properties due to irradiation are conservative.

The compact fracture toughness speci-mens were not tested at this time because no approved testing procedure was available.

The results of these tests are the subject of a separate re-port.

BabcockaWHcom

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The recommended operating period was extended to 15 effective full power years as a result of the second capsule evaluation.

These new operating limitations are in accordance with the requirements of Appendix G of 10 CFR 50.

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1 1

l CONTENTS Page 1.

INTRODUCTION..........................

1-1 1

2.

BACKGROUND...........................

2-1 3.

SURVEILLANCE PROGRAM DESCRIPTION................

3-1 4.

PREIRRADIATION TESTS......................

4-1 4-1 4.1.

Tension Tests......................

4.2.

Impact Tests.......................

4-1 4.3.

Compact Fracture Tests..................

4-2 5-1 5.

POSTIRRADIATION TESTS.....................

5.1.

Th e rmal Mo n i to rs.....................

5-1 5.2.

Tension Test Results...................

5-1 i

5.3.

Charpy V-Notch Impact Test Results............

5-2 1

6-1 6.

NEUTRON DOSIMETRY.......................

6.1.

Background........................

6-1 6.2.

Vess el Fl uence......................

6-2 6-2 6.3.

Capsule Fluence 6-3 6.4.

Uncertainty 7-1 7.

DISCUSSION OF CAPSULE RESULTS.................

7.1.

Freirradiation Property Data...............

7-1 7.2.

Irradiated Property Data.................

7-1 7.2.1.

Tensile Properties................

7-1 7-2 7.2.2.

Impact Properties 8.

DETERMINATION OF RCPB PRESSURE-TEMPERATURE LIMITS 8-1 l

9.

SUMMARY

OF RESULTS.......................

9-1 10-1

10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE 11-1

11. CERTIFICATION.........................

4 N A M ess

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i CONTENTS (Cont!d)

Page 4

APPENDIXES A.

Reactor Vessel Surveillance Program -- Background Da ta a nd I n fo rma t i o n..................

A-1 B.

Preirradiation Tensile Data B-1 C.

Preirradiation Charpy Impact Data C-1 D.

Fl uence Analysis Procedures D-1 E.

Capsule Dosimetry Data.................

E-1 F.

Re fe rence s.......................

F-1 4

4 List of Tables Table 3-1.

Specimens in Surveillance Capsule TEl-B 3-2 3-2.

Chemical Composition and Heat Treatment of 4

Surveillance Material s....................

3-3 5-1.

Tensile Properties of Capsule TEl-B Irradiated Base Me tal a nd Wel d Me tal.....................

5-2 5-2.

Charpy Impact Data for Capgule TEl-B Base Metal Irradiated to 5.92E18'n/ca' 5-3 5-3.

Charpy Impact Data for Capsule TEl-B Weat-Affected Zone Metal Irradiated to 5.92E18 n/ca'............

5-3 5-4. ' Charpy Impact Data for Capgule TEl-B Weld Metal Irradiated to 5.92E18 n/ca' 5-4 6-1.

Surveillance Capsule Detectors................

6-3 6-2.

Reactor Yessel Flux 6-4 6-3.

Reactor Vessel Fluence Gradient 6-5

'6-4.

Surveillance Capsule Fluence ;................

6-5 7-1.

Comparison of Tensile Test Results..............

7-5 7-2.

Summary of Davis-Besse Reactor Vessel Surveillance Capsules Tensile Test Results 7-6 7-3.

Observed Vs Predicted Changes in -Irradiated

~ Charpy Impact Properties...................

7-7 i

7-4.

Summary of Davis-Besse Reactor Vessel Surveillance Capsules ~Charpy Impact Test Results 7-8

1.

Data for Preparation of Pressure-Temperature Limit Curves for Davis-Besse -- Applicable Through 15 EFPY 8-4 A-1.

Unirradiated Impact Properties and Residual Element i

Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials.--

Davi s-Bes se 'hi t 1...................... -

A-3 A-2.

Test Specimens for Detennining Material Baseline 1

Properties..........................

A-4 A-3.

Specimens in Upper Surveillance Capsules...........

A-5 A-4.

Specimens in Lower Surveillance Capsules...........

A-5 NEh

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Tables (Cont'd)

Page Table B-1.

Preirradiation Tensile Properties for Base B-2 Metal Heat No. BCC241 B-2.

Preirradiation Tensile Properties for Weld B-2 Metal, WF-182-1 C-1.

Preirradiation Charpy Impact Data for Shell Forging C-2 Material -- Transverse Orientation, Heat BCC241 C-2.

Preirradiation Charpy Impact Data for Shell Forging C-3 Material -- Heat-Affected Zone, Heat BCC241 C-3.

Preirradiation Charpy Impact Data for Weld Metal, C-4 WF-182-1...........................

D-4 0-1.

Extrapolation of Reactor Vessel Fluence E-2 E-1.

Detector Composition and" Shielding..............

E-3 E-2.

Dosimeter Specific Activities E-5 E-3.

Dosimeter Activation Cross Sections, b/ atom List of Figures Figure 3-1.

Reactor Vessel Cross Section Showing Location of Davis-Besse Unit 1 Capsule TEl-B in Davis-Besse 3-4 Unit 1............................

3-5 3-2.

Loading Diagram for Test Specimens in TEl-B 5-1.

Impact Data for Irradiated Shell Forging Material, 5-5 Heat BCC241.........................

5-2.

Impact Data for Irradiated Shell Forging Material, 5-6 Heat-Affected Zone, Heat BCC241 5-7 5-3.

Impact Data for Irradiated Weld Metal, WF-182-1 6-6 6-1.

Reactor Vessel Flux / Fluence Gradient.............

6-2.

Azimuthal Flux / Fluence Gradient Inside Surface of 6-7 Reacto r Ye s sel........................

8-1.

Predicted Fast Neutron Fluences at Various Locations Through Reactor Vessel Wall for First 15 EFPY -- Davis-Besse 1....

8-5 8-2.

Reactor Vessel Pressure-Temperature Limit Curves for Normal 8-6 Operation -- Heatup Applicable for Firs.t 15 EFPY,......

8-3.

Reactor Vessel Pressure-Temperature Limit Curve for Normal 8-7 Operation -- Cooldown Applicable for First 15 EFPY......

8-4.

Reactor Vessel Pressure-Temperature Limit Curve for Inservice Leak and Hydrostatic Tests, Applicable 8-8 for First 15 EFPY A-1.

Location and Identification of Materials Used in A-6 Fabrication of Reactor Pressure Vessel............

C-1.

Impact Data for Unirradiated Shell Forging Material, C-5 Heat BCC241.........................

C-2.

Impact Data for Unirradiated Shell Forging Material, C-6 Heat-Affected Zone, Heat BCC241 C-7 C-3.

Impact Data for Unirradiated Weld Metal, WF-182-1 Babcock &WHcom

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1.

INTRODUCTION I.

This report describes the results of the examination of the second capsule of the Toledo Edison Company's Davis-Besse Nuclear Power Station Unit i reactor vessel material surveillance program.

The capsule was removed and examined at the end of the third fuel cycle. The first capsule of the pro-

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gram was removed and examined after the first year of operation; the re-sults are reported in BAW-1701.1 The objective of _ the program is to monitor the effects of neutron irradia-tion on the tensile and impact properties of reactor pressure vessel mate-rial s under actual. operating conditions.

The surveillance program for

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. Davis-Besse Unit I was designed and furnished by Babcock & Wilcox (B&W) as described in BAW-10100A.2 The program, designed in accordance with the re-l quirements of Appendix H to 10 CFR 50* and ASTM Specification'E185-73 and I

was planned to monitor the effects of neutron irradiation on the reactor vessel material for the 40-year design 11fe of the reactor pressure vessel.

4 The future operating limitations established after the evaluation of the surveillance capsule are also in accordance with the requirements of 10 CFR'

{

50, Appendixes G and H.

The recommended operating period was extended to 15 effective full: power years (EFPY) as a result of the second capsule eval-uation. -

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2.

BACKGROUND The ability of the reactor pressure vessel to resist fracture is the pri-mary factor in ensuring the safety of the primary system in light water-cooled reactors.

The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation.

The general effects of fast neutron irradiation on the mechanical proper-

. ties of such low-alloy ferritic steels as SA508, Class 2, used in the fabri-

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cation of the Davis-Besse Unit I reactor vessel, are well characterized and documented in the literature.

The low-alloy ferritic steels used in the beltlinc region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation.

The most serious mechanical property change in reactor pres-

'sure vessel steels is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the Charpy upper shelf impact toughness.

Appendix G to 10 CFR 50, " Fracture Toughness Requirements," specifies mini-mum fracture toughness requirements for the ferritic materials of the pres-sure-retaining components of the reactor coolant pressure boundary (RCPB) j of water-cooled power reactors, and provides specific guidelines for deter-mining the pressure-temperature limitations on operation of the RCPB.

The toughness and operational requirements are specified to provide adequate

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safety margins during any condition of normal operation, including antici-pated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.

Although the l

requirements of Appendix G to 10 CFR 50 became effective on August 13, 1973, the requirements are applicable to all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in op-eration on the effective date.

Appendix H to 10 CFR 50, " Reactor Vessel Materials Surveillance Program Re-quirements," defines the material surveillance program required to monitor 1

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changes in the fracture toughness properties of ferritic materials in the re-actor vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the thermal environment.

Fracture toughness test data are obtained from material specimens withdrawn periodically from the reactor vessel.

These data will pemit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life.

A nethod for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code,Section III, " Nuclear Power Plant Components."

This method utilizes fracture mech-anics concepts and the reference nil-ductility temperature, RTNDT, which is defined as the greater of the drop weight nil-ductility transition tempera-ture (per ASTM E-208) or the temperature that is 60F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion.

The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve), which appears in Appendix G of ASME Section III.

The KIR curve is a lower bound of dynamic, static, and crack arrest fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature.

Allowable operating limits can then be determined using these allowable stress intensity factors.

The RTHDT and, in turn, the operating limits of a nuclear power plant, can be adjusted to account for the effects of radiation on the properties of the reactor vessel material s.

The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be mon-itored by a surveillance program in which a surveillance capsule containing prepared specimens of the reactor vessel materials is periodically removed from the operating nuclear reactor and the specimens are tested.

The in-crease in the Charpy V-notch 30 ft-lb temperature is added to the original RTNDT to adjust it for radiation embrittlement.

This adjusted RTNDT is used to index the material to the KIR curve which, in turn, is used to set operat-ing limits for the nuclear power plant.

These new limits take into account the effects of irradiation on the reactor vessel materials.

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SURVEILLANCE PROGRAM DESCRIPTION The surveillance ~ program comprises six surveillance capsules designed to nonitor the effects of neutron and thermal environment on the materials of i

the reactor pressure vessel core region.

The capsules, which were inserted into the reactor vessel before initial plant startup, were positioned inside the reactor vessel between the thermal shield and the vessel wall at the locations shown in Figure' 3-1.

The six capsules, originally designed to be placed two in each helder tube, are positioned near the peak axial and azi-

. muthal neutron flux.

However, with the use of Davis-Besse Unit 1 as one of the irradiation sites of the 177 fuel assembly integrated reactor vessel material surveillance program, the capsules are irradiated on a. schedule integrated with the capsules of the other participating reactors.

This integrated schedule is described in BAW-1543.3 BAW-10100A includes a full description of the capsule design.

Capsule TEl-B was removed during the third refueling shutdown of Davis-Besse Unit 1.

This capsule contained Charpy Y-notch impact and tension test speci-mens fabricated from base metal (SA508, Class 2) and weld metal, and weld metal compact fracture specimens.

The specimens contained in the capsule are described in Table 3-1, and the location of the individual specimens within the capsule are described in Figure 3-2.

The chemical composition and heat treatment of the surveillance material in capsule TEl-B are described in Table 3-2.

l All test specimens were machined from the 1/4-thickness (1/4T) location of the forging material.

Charpy V-notch and tension test specimens from the vessel material were oriented with their longitudinal axes perpendicular to the principal working direction of the forging.

Capsule TEl-B contained dosimeter wires, described as follows:

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Dosimeter wire Shielding U-Al alloy Cd-Ag alloy Np-Al alloy Cd-Ag alloy Nickel Cd-Ag alloy I

0.66 wt % Co-Al alloy Cd-Ag alloy 0.66 wt % Co-Al alloy None Fe None Thermal monitors of low-melting eutectic alloys were included in the cap-sule. The eutectic alloys and their melting points are as follows:

Melting Alloy point, F 90% Pb, 5% Ag, 5% Sn 558 97.5% Pb, 2.5% Ag 580 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Cadmium 610 Lead 621 Table 3-1.

Specimens in Surveillance Capsule TEl-B Number of test specimens 1/2T comp 9cy fractureta Material description Tension CVN impact Weld metal 2

12 8

Weld-HAZ 12 Heat SS, transverse Base metal Heat SS, transverse 2

12 i

Total per capsule 4

36 8

(a) Compact fracture toughness specimens precracked per ASTM l

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Table 3-2.

Chemical Composition and Heat Treatment of Surveillance Materials Chemical Analysis Heat (a)

Weld meta' )

WF-182-1 b Element BCC241 C

0.22 0.09 Mn 0.63 1.69 P

0.011 0.014 S

0.011 0.013 Si 0.27 0.41 Ni 0.81

.0.63 CR 0.32 0.15 Mo=

0.63 0.40 Cu 0.02 0.21 Heat Treatment

Time, Heat No.

Temp, F h

Cooling BCC241 1640t10 4

Water quenched 1570t10 4

Water quenched 1240t10 6

Air cooled 1125125 40 Furnace cooled WF-182-1 1100-1150 15 Furnace cooled (a)Per Certified Materials Test Report.

(b)Per Licensing Document BAW-1500.

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Figure 3-1.

Reactor Vessel Cross Section Showing Location of Davis-Besse Unit 1 Capsule TEl-B in Davis-Besse Unit 1 X

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PREIRRADIATION TESTS Unirradiated material was evaluated for two purposes:

(1) to establish a baseline of data to which irradiated properties data could be referenced, and (2) to determine those materials properties to the extent practical from available material, as required for compliance with Appendixes G and H to 10 CFR 50.

4.1.

Tension Tests Tension test specimens were fabricated from the reactor vessel shell course forging and weld metal.

The subsize specimens were 4.25 inches long with a reduced section 1.750 inches long by 0.357 inch in diameter.

They were tested on a 55,000-lb load capacity universal test machine at a crosshead speed of 0.050 inch per minute.

A 4-pole extension device with a strain gaged extensometer was used to determine the 0.2% yield point.

Test condi-tions were in accordance with the applicable requirements of ASTM A370-72.

For cach material type and/or condition, six specimens in groups of three were tested at both room temperature and 580F.

The tension-compression load cell used had a certified accuracy of better than 10.5% of full scale (25,000 lb).

All test data for the preirradiation tensile specimens are given in Appendix B.

4.2.

Impact Tests Charpy V-notch impact tests were conducted in accordance with the require-ments of ASTM Standard Methods A370-72 and E23-72 on an impact tester certi-fled to meet Watertown standards.

Test specimens were of the Charpy V-notch type, which were nominally 0.394 inch square and 2.165 inches long.

Prior to testing, specimens were temperature-controlled in liquid immersion baths, capable of covering the temperature range froli -85 to +550F.

Speci-mens were - removed from the baths and positioned i ?.

the test frame anvil Babeeck&WWNees 4-1 a uconmost camp ny

with tongs specifically designed for the purpose.

The pendulum was re-leased manually, allowing the specimens to be broken within 5 seconds from their removal from the temperature baths.

Impact test data for the unirradiated baseline reference materials are pre-sented in Appendix C.

Tables C-1 through C-3 contain the basis data that are plotted in Figures C-1 through C-3.

4.3.

Compact Fracture Tests The compact fracture specimens fabricated from the weld metal, which were a part of the capsule specimen inventory, were tested by an approved single specimen J-integral testing procedure.5 The results of the testing of these specimens is reported in a separate report, BAW-1835.4 4

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POSTIRRADIATION TESTS 5.1.

Thermal Monitors Capsule TEl-B contained three temperature monitor holder tubes, each contain-ing 'ive fusible alloy wires with melting points ranging from 558 to 621F.

All the thermal monitors at 558, 580, and 588F had melted while those at the 610F location showed partial melting or slumping; the monitor at the 621F lo-cation melted in all three holder tubes.

It is therefore assumed that the 610F and 621F monitors were placed in the wrong locations in the holder tubes.

From these observations, it was concluded that the capsule had been exposed to a peak temperature' in the range of 610 to 621F during the reactor operating period.

These peak temperatures are attributed to operating tran-sients that are of short durations and are judged to have insignificant ef-fect on irradiation damage.

Short duration operating transients cause the use of thermal monitor wires to be of limited value in. determining the maxi-mum steady state operating temperature of the surveillance capsules, however, it is judged that the maximum steady state operating temperature of specimens in the capsule was held within 125F of the 1/4T vessel thickness location temperature of 577F.

It is concluded that the capsule design temperature may have been exceeded during operating transients but not for times and temperatures that would make the capsule data unusable.

5.2.

Tension Test Results The results of the postirradiation tension tests are presented in Table 5-1.

Tests were performed on specimens at both room temperature and 580F using the same test procedures and techniquas used to test the unirradiated specimens (section 4.1).

In general, the ultimate strength and yield strength of the material increased with a corresponding slight decrease in ductility; both effects were the result of neutron radiation damage.

The type of behavior observed and the degree to which the material properties changed is within the range of changes to be expected for the radiation environment to which the specimens were exposed.

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The results of the preirradiation tension tests are presented in Appendix B.

5.3.

Charpy V-Notch Impact Test Results The test results from the irradiated Charpy V-notch specimens of the reac-tor vessel beltline material are presented in Tables 5-2 through 5-4 and Figures 5-1 through 5-3.

The test procedures and techniques were the same as those used to test the unirradf 6ted specimens (section 4.2).

The data show that the material exhibited a sensitivity to irradiation within the values predicted from its chemical composition and the fluence to which it was exposed.

The results of the preirradiation Charpy V-notch impact test are given in Appendix C.

Table 5-1.

Tensile Properties of Capsule TEl-B Irradiated Base Metal and Weld Metal Specimen Test temp.

- - - -Strength, psi Elongation, %

9 No.

F Yield Ult.

Unif Total Base Metal. Transverse 5.92 x 1018 n/cm2 (E > 1 Mev)

SS-621 76 70,100 91,100 11 26 65 SS-619 580 66,900 87,500 8

21 57 Weld Metal, 5.92 x 1018 n/cm2 (E > 1 Mev)

SS-014 76 85,500 100,900 10 16 54 SS-017 580 77,800 93,900 8

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Table 5-2.

Charpy Impact Data For Caplule TEl-B Base Metal Irradiated to 5.92E18 n/cM (E >l MeV)

Absorbed lateral Shear Specimen Test temp,

energy, expansion,

fracture, No.

F ft-lb 10-3 in.

SS-669

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0 SS-631

-2 29.0 20 0

55-692 40 50.0 43 10 55-626 75 62.0 49 30 55-632 115 93.0 68 80 55-601 153 08.0 76 70 SS-613 222 107.0 78 100 SS-653 290 112.0 87 100 55-648 326 117.0 87 100 S5-619 376 115.0 89 100 SS-640 456 118.0 94 100 SS-602 554 102.0 70 100 Table 5-3.

Charpy Inpact Data For Capsule TEl-B Heat-Affected Igne Metal Irradiated to 5.92E18 n/ W (E > 1 MeV)

Absorbed Lateral Shear Specimen Test temp, enery,

expaDsion,

fracture, No.

F ft-lb 10-J in.

SS-316

-47 26.0 13.0 20 SS-312

-15 52.0 30.0 10 SS-366

-2 66.0 45.0 80 SS-345 40 95.0 62.0 70 SS-383 75 82.0 59.0 70 SS-326 153 104.0 74.0 100 SS-324 222 100.0 78.0 100 SS-350 288 110.0 82.0 100 SS-372 326 117.0 84.0 100 SS-360 376 113.0 86.0 100 SS-318 456 106.0 79.0 100 SS-388 548 103.0 81.0 100 5-3 Estesser&WWIsons a AkDermott company L

Table 5-4.

Irradiated Charpy Impact Data for Capsule TEl-B Weld Metal Irradiated to 5.92E18 n/cm2 (E > 1 MeV)

Absorbed Lateral Shear Specimen Test temp,

energy, expagston, fracture.

No.

F ft-lb 10-3 in.

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Impact Data for Irradiated Shell Forging Material.

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=

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I 0-00 40 0

40 80 120 160 200 240 200 320

}60 400 fastinneenafvet,F 5-5 N&h a hecDefmott tempany

Figure 5-2.

Impact Data for Irradiated Shell Forging Material, Heat-Affected Zone, Heat BCC241 W

I I

I i

1 i

I-i-

i -

,n e

I-y2 25 -

I t

I I

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g DATA SulW RY 180-i not 12F Tcy (35 ns)

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=140 gi

-88F y

not d120 E

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t 60

~

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liest flo, SCC 291 I

I I

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00

-40 0

40 80 120, 160 200 24 200 320 360 400 Test Taw enatune, F N8h 5-6 a AAcDermott company L

Figure 5-3.

Impact Data for Irradiated Weld Metal, WF-182-1 W

I I

I 6

6 i

~

ee is..

.!yo - - - - - - - - - - -. - - - -

32s e

f f

1 I

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I I

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. tier lf4 g.02

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I I

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-- DATA SUPUF4Y N.A.

go-f,,,

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+teer f

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~

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30 - ~ ~ - - ~ ~ ~ - ~ ~ -

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

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4 6.

NEUTRON DOSIMETRY 6.1.

Background

Fluence analysis as a part of the reactor vessel surveillance program has three objectives:

(1) determination of maximum fluence at the pressure ves-sel as a function of reactor operation, (2) prediction of pressure vessel fluence in the future, and (3) determination of the test specimen fluence within the surveillance capsule.

Vessel fluence data are used to evaluate changes in the reference transition temperature and upper shelf energy levels, and to establish pressure-temperature operation curves.

Test spect-men fluence data are used to establish a correlation between changes in ma-terial properties and fluen'ce exposure.

Fluence data are obtained from cal-culations based on measured capsule dosimeter activities.

A significant aspect of the surveillance program is to provide a correla-tion between the neutron fluence above 1 MeV and the radiation-induced prop-erty changes noted in the surveillance specimens.

To pennit such a correla-tion, activation detectors with reaction thresholds in the energy range of interest were placed in each surveillance capsule.

The significant proper-ties of the detectors are given in Tables 6-1 and E-1.

Because of a long half-life (30 years) and effective threshold energies of 137 s production from fission reac-0.5 and 1.1 MeV, the measurements of C

238 ) are more directly appitcable to analytical deter-tions in 237Np (and U

minations of the fast neutron fluence (E > 1 MeV) for multiple fuel cycles than are other dosimeter reactions.

Other dosimeter reactions are useful as corroborating data for shorter time intervals and/or higher energy fluxes.

Short-lived isotope activities are representative of reactor condi-tions only over the latter portion of the irradiation period (fuel cycle),

whereas reaction: with a threshold energy > 2 or 3 MeV do not record a sig-nificant part of the total fast flux.

I 6-1 a McDermott con 9any L_

The energy-dependent neutron flux is not directly available for activation detectors because the dosimeters register only the integrated effect of neu-tron flux on the target material as a function of both irradiation time and neutron energy.

To obtain an accurate estimate of the average neutron flux incident upon the detector, several parameters must be known:

the operat-ing history of the reactor, the energy response of the given detector, and the neutron spectrum at the detector location.

Of these parameters, the definition of the neutron spectrum is the most difficult to obtain.

6.2.

Vessel Fluence The maximum fluence (E > 1.0 MeV) in the pressure vessel was determined to be 9.88 (+17) n/cm2 2

based on an averge neutron flux of 2.01 (+10) n/cm.3 for cycles 2 and 3 (Tables 6-2 and 6-3).

The location of maximum fluence is assumed to be at the same point at the cladding / vessel interface as for cycle 1:

at an elevation of about 110 cm above the lower active fuel bound-ary and at an azimuthal (peripheral) location of N11' from a major axis (across flats dia eter).

Fluence data have been extrapolated to 32 EFPY of operation based on the premise that excore flux is proportional to fast flux that escapes the reactor core (Appendix D).

Core escape flux values are available for fuel management analyses of future fuel cycles.

Relative fluence as a function of radial location in the pressure vessel is shown in Figure 6-1.

Reactor vessel lead factors (clad interface flux /in-vessel flux) for the T/4, T/2, 3T/4 locations are 1.8, 3.7, and 7.9, respec-tively.

Relative fluence as a function of azimuthal angle is shown in Fig-ure 6-2.

A peak occurs at s11' which roughly corresponds to a corner of the core and to three f,yenetric rapsule locations.

Two other capsule loca-tions (including capsulo TEl-D) correspond to the azimuthal minimum at

%27*.

However, it should be noted that the maximum to minimum flux ratio is only 1.5.

Fast neatron flux is increased by $1.25 in the capsule due to differences in scattering and absorption cross sections between steel and water.

6.3.

Capsule Fluence Fast fluence at the center of the surveillance capsules was calculated to be 5.92 (+18) n/cm2 (Table 6-4).

These data represent average valuus in the capsule.

Capsule TEl-B was located in a upper holder tube position at 6-2 h*8k M888 a wonmons wmoany

427' off axis and 202 cm from the core exis during cycles 1, 2, and 3 (943 EFPD).

6.4.

Uncertainty 1.

The estimated uncertainty of the capsule fast flux and fluence values is 15%.

2.

The estimated uncertainty of the calculated reactor vessel fluence values is 28%.

3.

The estimated uncertainty of the EOL reactor vessel fluence predictions is 30%.

Table 6-1.

Surveillance Capsule Detectors Effective lower energy limit, Isotope Detector reaction MeV half-life 54 e(n.p)S4 n 2.5 312.5 days F

M 58N1(n.p)58 o 2.3 70.85 days C

238 (n,f)l37 s 1.1 30.03 years U

C 237Np(n f)137 s 0.5 30.03 years C

i Babeeck4h 6-3

. w o.,

m

,.ny

Table-6-2.

Reactor Vessel Flux Fa'st Flux, n/cm2-s

. Fast flux, n/cm2-s -(E > 1 MeV)-

(E > 0.1 MeV)

Inside surface Inside surface (max location)

T/4 T/2 3T/4-(max location)

Cycle 1(a) 1.61 (+10) 8.9 (+9) 4.4 (+9) 2.0 (+9) 3.4(+10)

(374 EFPD)

Cycles 2 8.3(a) 2.01 (+10) 1.1 (+10) 5.4 (+9)-

2.5 (+9)

(569 EFPD)

Cycle 4(b) 2.12.(+10) 1.2 (+10) 5.7 (+9) 2.7 (+9)

Cycle 5(b) 1.51 (+10) 8.4 (+9) 4.1 (+9) 1.9 (+9)

(a) Calculated.

(b) Predicted.

2 1ao

Table 6-3.

Reactor Vessel Fluence Gradient Fast fluence, n/cm2 (E > 1.0 MeV)

Cumulative Inside surfcce -

irradiation time (max location)

T/4 T/2 3T/4 End of cycle 1 5.19 (+17) 2.9 (+17) 1.4 (+17) 6.6 (+16)

(374 EFPD) i End of cycle 3 1.51 (+18) 8.4 (+17) 4.1 i+17) 1.9 (+17)

(943 EFPD)

End of cycle 4 1.98 (+18) 1.1 (+18) 5.4 (+17) 2.5 (+17)

(3.3 EFPY)

End of cycle 5 2.5 (+18) 1.4 (+18) 6.7 (+17) 3.2 (+17)

(4.4 EFPY)

End of 8 EFPY 4.2 (+18) 2.3 (+18) 1.1 (+18) 5.4 (+17)

End of 15 EFPY 7.6 (+18) 4.2 (+18) 2.0 (+18) 9.6 (+17) i End of 21 EFPY 1.0 (+19) 5.6 (+18) 2.8 (+18) 1.3 (+18)

End of 32 EFPY 1.6 (+19) 8.9 (+18) 4.2 (+18) 2.0 (+18)

Table 6-4.

Surveillance Capsule Fluence E > 1.0 MeV Flux Fluenge z

z (n/cm -s)

(n/cm )

TEl-F cycle 1 6.08 (+10) 1.96 (+18)

(374 EFPD)

.TEl-B cycles 2 & 3 8.06 (+10) 3.96 (+18) l (569 EFPD)

TEl-B cycles 1, 2, 5.92 (+18) and 3 (943 EFPD) i Sa4 cock &WUHcom a McDermott company

,.. - ~.. -....,, -.

- ~ -..... - - -

l Figure 6-1.

Reactor Vessel Flux / Fluence Gradient 1.5

1. 0. ( Cl ad Interface)

Lead Factor I.D.

l.0 T/4 l

= 1. 81

.55

.8 T/2

= 3.69 217.32 CH

.27

.7 3T/4 1.

= 7. 9 4

.126

.6

.5 T/4

.g 222.68 CH 9

8 E

~

.3

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5 T/2

.5

2 228.04 CM a:

~

3T/4

.08 233.39 CM

.07 o,0.

.06

.05 g

i I

2 20 225 230.

235 238.75 CM f rom Core Center t-4,

6-6 WSWHcom

. ucoermott company

Figure 6-2.

Azimuthal Flux / Fluence Gradient Inside Surface of Reactor Vessel l.07 l

l.05 l

1.00 i

O.95 i

8 3

0.90 2

m i

0 y

0.85 b

i 2

E I

0.80 0.75 i

E i

0.70 8

j 0.67 I

I I

i i

l i

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fh O'

5 10 15 20 25 30 35 40 45

)

Deviation f rom Major Axi s, Degrees i

i

l 7.

DISCUSSION OF CAPSULE RESULTS 7.1.

Preirradiation Property Data A review of the unirradiated properties of the reactor vessel core belt re-gion indicated no significant deviation from expected properties except in l

the case of the upper shelf properties of the weld metal. Based on the pre-dicted end-of-service peak neutron fluence value at the 1/4T vessel wall lo-cation and the copper content of this weld, it was predicted that the end-of-service Charpy upper shelf energy (USE) will be below 50 f t-lb.

This weld was selected for inclusion in the surveillance program in accordance with the criteria in effect t.t the time the program was designed for Davis-Besse Unit 1.

The applicable selection criterion was based on the unirradiated properties only.

7.2.

Irradiated Property Data 7.2.1.

Tensile Properties Table 7-1 compares irradiated and unirradiated tensile properties.

At both room temperature and elevated temperature, the ultimate and yield strength changes in the base metal as a result of irradiation and the corresponding changes in ductility are negligible.

There appears to be some strengthen-l ing, as indicated by increases in ultimate and yield strength and similar decreases in ductility properties.

All changes observed in the base metal are, such as to be considered within acceptable limits.

The changes at both room temperature and 580F in the properties of the weld metal are greater than those observed-for the base metal, indicating a greater sensitivity of

- the weld. metal to. irradiation damage.

In _either case, the changes in ten-sile properties are insignificant relative to the analysis of the reactor vessel materials at this period in service life.

A comparison of the tensile data from the first capsule (capsule TEl-F) with the corresponding data from the capsule reported in this report is i

7-1 MNEN a McDermott company

.k

shown in-Table 7-2.

The currently reported data experienced a fluence that is three times greater than the previous capsule.

The general behavior of the tensile properties as a function of neutron ir-radiation is an increase in both ultimate and yield strength and a decrease in ductility as measured by both total elongation and reduction of area.

The most significant observation from these data is that the weld metal ex-hibited much greater sensitivity to neutron radiation than the base metal.

7.2.2.

Impact Properties The' behavior of the Charpy V-notch impact data is more significant to the calculation of the reactor system's operating limitations.

Table 7-3 com-pares the observed changes in irradiated Charpy impact properties with the predicted changes.

The 50 ft-lb transition temperature shift for the base metal is not in good agreement with the shift that would be predicted according to Regulatory Guide 1.99.

The less than ideal comparison may be attributed to the spread in the data of the unirradiated material combined with a minimum of data points to establish the irradiated curve.

Under these conditions, the com-parison indicates that the estimating curves in RG 1.99 for medium-copper materials and at mid-range fluence levels are conservative for predicting the 50 ft-lb transition temperature shifts.

The 30 ft-lb transition temperature shift for the base metal is not in good agreement with the value predicted according to Regulatory Guide 1.99, al-

~ though it would be expected that these values would exhibit better agree-ment when it is considered that a major portion of the data used to develop Regulatory Guide 1.99 was taken at the 30 ft-1b temperature.

The increase in the 35-mil lateral expansic, transition, temperature is.com-pared with the shift in RTNDT curve data in a manner similar to the compari-son made for the 50 ft-lb transition temperature shift.

These data show a behavior similar to that observed from the comparison of the observed and predicted 50 ft-lb transition data.

e The transition temperature measurements at 30 and 50 ft-lbs for the weld metal are in poor agreement with the predicted shift.

This may be attrib-uteti to the decrease in upper shelf which caused a larger shift at the 50 a

MW 7-2 a

a McDermott company

ft-lb level than at the 30 ft-lb level.

At the 30 ft-lb level the shift is much less than the predicted value which indicates that the estimating curves are conservative for predicting the 30 ft-lb transiticn temperature shifts.

The data for the decrease in Chtrpy USE with irradiation did not show good s

agreement with predicted values for both the base metal and the weld metal.

However, the poor comparison of the measured data with the predicted value is not unexpected in view of the lack of data for medium-to high-copper-content materials at low to medium fluence values that were used to develop the estimating curves.

A comparison of the Charpy impact data from the first capsule (capsule TEl-F) with the corresponding data from the capsule reported in this report is shown in Table 7-4.

The currently reported data experienced a fluence that is three times greater than the previous capsule.

The base metal exhibited shifts at the 30 ft-lb and 50 ft-lb levels for the latest capsule that were greater than those of the first capsule.

The cor-responding data for the weld metal showed about a 70% increase at the 50 ft-lb level and for all practical purposes no change at the 30 ft-lb level.

This was due to the fact that there was a large decrease in the upper shelf drop which caused a larger shift at the 50 ft-lb level as compared to the 30 ft-lb level.

Both the base metal and the weld metal exhibited further decrease in the upper shelf values.

These data confirm that the upper shelf drop for this weld metul does not reach saturation as observed in the results of capsules evaluated by others.

This lack of saturation of Charpy USE drop for this weld metal should not be cwsidered indicative of a similar lack of satura-tion of upper sitelf region fracture toughness properties.

This relation-ship must await the testing and evaluation of the data from compact frac-ture toughness test specimens.

Results from other surveillance capsules also indicate that RTNDT estimat-ing curves have greater inaccuracies than originally thought.

These inac-curacies are a function of a number of parameters related to the basic data available It the time the estimating curves were established.

Some of these include inaccurate f~iuence values, poor chemical composition values, 7-3 i

Babcock &WHcom a McDermott company

and variations in data interpretation.

The change in the regulations re-quiring the shift measurement to be based on the 30 ft-lb value will mini-mize errors that result from using the 30 ft-lb data base to predict the shift behavior at 50 ft-lbs.

The design curves for predicting the shif t will be modified as more data be-come available; until that time, the design curves fcr predicting the RTNDT shift as given in Regulatory Guide 1.99 are considered adequate for predict-ing the RTNDT shift of those materials for which data are not available and will continue to be used to establish the pressure-temperature operational limitations for the irradiated portions of the reactor vessel until the time that new prediction curves are developed and approved.

The lack -of good agreement of the change in Charpy USE is further support of the inaccuracy of the prediction curves at the lower fluence values. Al-though the prediction curves are conservative in that they predict a larger drop in upper shelf than is observed for a given fluence and copper con-tent, the conservatism can unduly restrict the operational limitations.

These data support the contention that the USE drop curves will have to be modified as more reliable data become available; until that time the design curves used to predict the decrease in USE of the controlling materials are considered conservative.

7-4 Babcock &WHcom a McDermott company

Table 7-1.

Comparison of Tensile Test Results Elevated temp Room temp test test (580F)

Unfrr Irrad Unirr Irrad Base Metal -- BCC-241 Transverse Fluence, 1018 n/cm2 (>l MeV) 0 5.92 0

5.92 Ult. tensile strength, ksi 90.7 91.1 86.3 87.5 0.2% yield strength, ksi 72.3 70.1 64.0 66.9 Elongation, %

28 26 26 21 RA, %

68 65 65 57 Weld Metal -- WF-182-1 Fluence, 1018 n/cm2 (>l MeV) 0 5.92 0

5.92.

Ul t. ' tensile strength, ksi 85.6 100.9 83.2 93.9 0.2% yield strength, ksi 70.2 85.5 67.6 77.8 Elongation, %

27 16 19 15 RA, %

64 54 50 42 i

l i

i l

l-l l

7-5 R h & Mfcom a McDermott company

Table 7-2.

Summary of Davis-Besse Reactor Vessel Surveillance Capsules Tensile Test Results Ductility, %

Strength, ksi of-Material 101 n/cm2 temp, F Ultimate

% "a (*) Yield

% a (*) Total F1 ence Test elon..

% a (*) A

% a (*)

28 68 Base metal 0

73 90.7 72.3 26 65 580 86.3 64.0 1.96 70 95.6

+5.4 75.0

+3.7 26.

-7.1 66

-2.3 580 88.8

+2.9 66.3

+3.6 22

-15.4 59

-9.2 5.92 76 91.1

+0.4 70.1

-3.0 26

-7.1 65

-4.4 580 87.5

+1.4 66.9

+4.5 21

-19.2 57

-14.0 Weld metal 0

73 85.6 70.2 27 64 50 78 580 83.2 67.6 19 m

1.96 70 98.1

+14.6 82.5

+17.5 25

-7.4 58

-9.4 580 90.0

+8.2 73.1

+8.1 21

+15.8 48

-4.2 5.92 76 100.9

+17.8 85.5

+21.8 16

-40.7 5A

-15.6 580 93.9

+12.8 77.8

+15.1 15

-21.0 42

-16.0 i

(*) Change relative to unirradiated.

n E

a5D Ii M

Table 7-3.

Obsterved Vs Predicted Changes in Irradiated Charpy Impact Properties Material Observed Predicted (a)

Increase in 30 ft-lb trans temp, F Base material (BCC-241)

Transverse Neg 43 Heat-affected zone (BCC-241) 57 43 Weld metal (WF-182-1) 125 154 Increase in 50 ft-lb trans temp, F Base material (BCC-241))

Transverse 16 43 Heat-affected zone (BCC-241) 36 43 Weld metal (WF-182-1) 194 154 Increase in 35 MLE trans tertp, F Base material (BCC-241)

Transverse 8

43(b)

Heat-affected zone (BCC-241) 33 43(b)

Weld metal (WF-182-1) 158 154(b)

Decrease in Charpy USE, ft-lb Base material (BCC-241)

Transverse 17 13 Heat-affected zone (BCC-241) 20 13 Weld metal (WF-182-1) 13 23 (a)These values predicted per Regulatory Guide 1.99, Revision 1.

(b) Based on the assumption that MLE as well as 50 ft-lb transition temperature is used to control the shift in RTNDT-7-7 h & Mfcom a McDermott company

Table 7-4.

Summary of Davis-Besse Reactor Yessel Surveillance Capsules Charpy Impact Test Results Trans temp increase, F Decrease in upper 0 ft-lb 50 ft-h she H ene m, ft-b F:

ence,2 n/cm Observ.

Predicted Observ.

Predicted Material 10 <

Base metal 1.%

Neg 25 1

25 7:

13 5.92 Neg 43 16 43 17 13 HAZ metal 1.96 43 25 13 25 15 13

_5.92 57 43 36 43 20 13 i

Weld metal 1.96 127 89 113 89 6

16 5.92 125 154 194 154 13 23

?

m i

1 4

et I

2

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' q

{

i 8.

DETERMINATION OF RCPB PRESSURE-TEMPERATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) of Davis-Besse Unit 1 are established in accordance with the require-ments of 10 CFR 50, Appendix G.

The methods and criteria employed to estab-lish operating pressure and temperature limits are described in topical re-port BAW-10046.6 The objective of these limits is to prevent nonductile i

failure during any normal operating condition, including anticipated opera-tion occurrences and system hydrostatic tests.

The loading conditions of interest include _ the following:

1.

Normal operations, including heatup and cooldown.

f 2.

Inservice leak'and hydrostatic tests.

L-3.

Reactor! core operation.

The major components ~ of the RCPB have been analyzed in accordance with 10 i

CFR 50, Appendix'G.

The closure head region, the reactor vessel outlet noz-zie, ' and the beltline region have been identified ' as the only regions of i

the reactor vessel.(and conseq'uently of the RCPB) 'that regulate the pres-

sure-temperature limits.

Since t~ne closure head region is significantly.

stressed at relatively low temperatures (due to mechanical loads resulting from bolt preload), this region largely controls the pressure-temperature limits of the first several service periods.-

The reactor vessel outlet

. nozzle also affects - the pressure-temperature limit curves of ~ the first several service periods.

This is ' due to the high local stresses at the l

inside corner. of :the nozzle, which can be two - to three times' the membrane

stresses of the shell.

After the first several years of neutron radiation

[.

exposure, the RTNDT of the beltline. region materials will be high. enough that the beltline _ region of the reactor vessel will start to control the

~

pressure-temperature limits of the RCPB. - For the service period for. which the limit curves are established, the maximum allowable pressure as a func-tion-of fluid temperature is obtained through a point-by-point comparison 8-1 Balseeck&WIIcom a McDermott comparty

of the limits imposed by the closure head region, the outlet nozzle, and the; beltline region.

The maximum allowable pressure is taken to be the lowest of three calculated pressures.

The limit curves for Davis-Besse Unit 1 are based on the predicted values of the adjusted reference temperatures of all the beltline region materials

' at the end of the fifteenth EFPY.

The fifteenth EFPY was selected because i

it is estimated that the third surveillance capsule will be withdrawn at the end of -the refueling cycle when the estimated capsule fluence corre-

- sponds ' to approximately the twenty-first EFPY.

The time difference between the withdrawal of the second and third surveillance capsule provides ade-quate time for re-establishing the operating pressure and temperature lim-its for the period period of operation between the third and fourth sur'v'eil-lance capsule withdrawals.

The unirradiated impact properties were detennined for the surveillance beltline region materials in accordance with 10 CFR 50, Appendixes G and H.

For the other beltline 1 region and RCPB materials for which. the measured properties are not available, the unirradiated impact properties and resid-ual elements, as originally established for the beltline region materials, f

are listed in Table A-1.

. The adjusted reference temperatures are calcu-

-lated by adding the predicted radiation-induced RTNDT and the unirradiated RTNDT.

The predicted ARTNDT is calculated using the respective neutron fluence and copper and phosphorus contents.

Figure 8-1 illustrates the cal-culated peak neutron fluence at several locations through the reactor ves-sel beltline region wall.

The supporting information -for Figure 8-1 is described in section 6.

The neutron fluence values of Figure 8-1 are the predicted fluences that have been demonstrated (section 6) to be conserva-ti ve.

The design curves of Regulatory Guido 1.99* were' used to predict the radiation-induced ARTNDT values as a function of the material's copper and phosphorus content and neutron fluence.

l t

Revision 1, April 1977.

i-M 8-2 l

a McDermott company

The neutron fluer.ces and adjusted RTNDT values of the beltline region ma-terials at the end of the fifteenth full-power year are listed in Table 8-1.

The neutron fluences and adjusted RTNDT values are given for the 1/4T and 3/4T vessel wall locations (T = wall thickness).

The assumed RTNDT Of the closure head region and the outlet nozzle steel forgings is 60F, in ac-cordance with BAW-10046.

Figure 8-2 shows the reactor vessel's pressure-temperature limit curve for normal heatup.

This figure also shows the core criticality limits as re-quired by 10 CFR 50, Appendix G.

Figures 8-3 and 8-4 show the vessel's pressure-temperature limit curve for normal cooldown and for heatup during inservice leak and hydrostatic tests, respectively.

All pressure-tempera-ture limit curves are applicable up to the sixteenth EFPY.

Protection against nonductile failure is ensured by maintaining tne coolant pressure below the upper limits of the pressure-temperature limit curves.

The ac-ceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve.

The reactor is not per-mitted to 90 critical until the pressure-temperature combinations are to the right of the criticality limit curve.

To establish the pressure-tem-perature limits for protection against nonductile failure of the RCPB, the limits presented in Figures 8-2 through 8-4 must be adjusted by the pres-sure differential between the point of system pressure measurement and the pressure on the reactor vessel controlling the limit curves.

This is necessary because the reactor vessel is the most limiting component of the RCPB.

8-3 Babcock &WHeos a McDermott company

.~

v T'

s

' Table 8-1.

Data for Preparation of Pressure-Temperature Limit' Curves for.

Davis-Besse -- Applicable Through 15 EFPY Weldent loca+1m Radiation-frvLced Neuttm flutnte at RTgT at W(M ) Adjusted RT0T at Oumistry Core Location end of 15 EFPY gypy* p c,d mi41ane fras Weld Copper Phosphorus (E > 1 M). n/on2 end of 15 UPY, F w,y gg Beltline to wid ' major axis, 1/4T

. thier

contert, rontert, n.

Heat No.

Type region locatim CL, on degrees location RTmT. F At 1/4T At 3/4T 4 1/4T At 1/4T At 1/4T. At 314T.

A08-203 963, C1. 2 kxtzle belt

+50 0.04 0.007 1.0E18

2. 1 17 13 6

63

'56

+20 -

0.04 0.004 4.2E18 9.6E17 26 12 46 32 XJ-233 5A508. C1. 2 l$per shell 450 0.E 0.011 4.2E18 9.6E17 36 17 86 67 BE-241 SMiOB, C1. 2 laer shell No

(+20)(8) 0.18(b) 0.016(b)

W-232 eld t4per circum sem (10 95)

+198 W-233 Weld L5per ciram sem (m 91%) +198 Yes

(+20)(a) 0.29(b) 0.021(b)

'1.0E18 2.3E17 X)0 48 120 68 W-182-1 Weld Hid ciram sem (1001)

-24 Yes

+2 0.24(b) 0.014(o) 4.2E18 9.6E18 l49 71 151 73 j

W-232 Weld Loer cirum sea (!012%) -247 No

(+3))(a) o,tg(b) 0.016(b)

Yes

(+20)(a) 0.29(b) 0,gt(b) 5.1E15 1.1E15 6

3 26'-

23 W-233 Weld Loer cirum sem ((D Bat) -247

?*

(al er BM-10046A, Rev.1. July 1977.

P (b)Per BM-1799, July 1983.

^

(c)Per Replatory Gste 1.99, Art.1, April 1977.

(d)Replatory Cuide 1.92 not valid for shifts less than SF.

e e

il O

8E ia3M

Figure 8-1.

Predicted Fast Neutron Fluences at Various locations Through Reactor Vessel Wall for First 15 EFPY -- Davis-Besse 1

8. 0 7.6x1018 nvt 7.0 C

5 6.0 c

E A

5.0 E

m i

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2 4.0

p e

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$\\%

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E 9.6x1017nv t h

I.0 g

Vessel 3/4T Location b

3.6x10l7nvth g

Vessel outside surface 3

j 0

i I

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g 0

1 2

3 4

8 9

10 11 12 13 14 15 g

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=

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>= x

a..a 2 W

r*

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6 sd 'eJnssaJd luegooo tassoA Jogosay 1

8-6 NSM8888 a McDermott company

4 Figure 8-3.

Reactor Vessel Pressure-Temperature Limit Curve for Normal Operation -- Cooldown Applicable for First 15 EFPY 2400 ASSUMED RTNDT, F P0lNT PRESSURE, PSIG TEMP., F E

2M

- BELTLINE REGION 1/4T 151 A

299 70 BELTLINE REGION 3/4T 73 8

625 162 2000 CLOSURE NEAD REGION 60 C

625 195 0UTLINE N0ZZLE 60 D

850 200 m

E 2250 303 1800 THE ACCEPTABLE PRESSURE. TEMPERATURE COMBINATIONS ARE BELOW AND TO THE RIGHT a

OF THE LIMIT CURVE (s). THE LIMIT CURVES D0 NOT INCLUDE THE PRESSURE 1600 DIFFERENTIAL BETWEEN THE POINT 0F SYSTEM PRESSURE MEASUREMENT AND THE PRESSURE ON THE REACTOR VESSEL REGION CONTROLLING THE LIMIT CURVE, NOR D0

=

h THEY IN;LUDr. ANY ADDITIONAL I4ARGIN OF SAFETY FOR POSSIBLE INSTRUMENT IWO E""08-a.

T 1*

.?

8 IMO e'

APPLICABLE FOR C00LDOWN RATES UP T0100F/h to 270F AND THEN h

1000 50F/h 0

800 ac 600 B

C f~

e 400 k

i S

g Cooldown Rates

~

Up to 50F/hr Up to 100F/hr

D l

0 I

I

'l i

I I

I 40 80 120 160 200 2%

280 320 360 4

ft Reactor Vessel Coolant Temperature, F i

h

oe A4 o

3 2

am 03 moa-<

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sm

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> 0

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meuo

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i i

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61sd 'aJnsseJa ;ustooo tassaA Jogseeg 8-8 Batwock&Micos a McDermott company

9.

SUttiARY OF RESULTS The analysis of the reactor vessel material contained in the second surveil-lance capsule, TEl-B, removed for evaluation as part of the Davis-Besse Unit 1 Reactor Vessel Surveillance Program, led to the folicwing conclu-sions:

1.

The capsule received an average fast fluence of 5.92 x 1018 n/cm2 (E >

1.0 MeV). The predicted fast fluen::e for the reactor vessel T/4 loca-tion at the end of the third fuel cycle is 8.4 x 1017 n/cm2 (E > 1 MeV).

2.~ The fast fluence of 5.92 x 1018 n/cm2 (E > 1 MeV) increased the RTNDT of the capsule reactor vessel core region shell materials a maximum of 125F.

3.

Based on the calculated fast flux at the vessel wall, an 80% load fac-tor and the planned fuel management, the projected fast fluence that the Davis-Besse Unit I reactor pressure vessel will receive in 40 calendar years' operation is 1.6 x 1019 n/cm2 (E > 1 MeV).

4.

The increase in the RTNDT for the base plate material was not in good agreement with that predicted by the currently used design curves of RTNDT versus fluence, but the prediction techniques are conservative.

5.

The increase in the RTNDT for the weld metal was not in good agreement with that predicted by the currently used design curves of RTNDT versus fluence but the prediction techniques are conservative.

6.

The current techniques used to predict the change in weld metal Charpy upper shelf properties due to irradiation are conservative.

7.

The analysis of the neutron dosimeters demonstrated that the analytical techniques used to predict the neutron flux and fluence were accurate.

8.

The capsule design operating temperature may have been exceeded during operating transients but not for times and temperatures that would make the capsule data unusable.

Balseeck&WIIsos 9-1 a ucoermote con,any

10.

SURVEILLANCE CAPSULE REMOVAL SCHEDULE Based on the postirradiation test results of capsule TEl-B, the following schedule is recommended for examination of the remaining capsules in the Davis-Besse Unit 1 RVSP:

Evaluation schedule (a)

Est. vessel Est. capsule thg'n" 2

Est. date Capsule Quencon/cm2 Surface 1/4T available(b) data ID 10 TEl-A 1.1 0.20 0.11 1985 TEl-C 1.6 0.32 0.17 1988 TEl-D(c) 1.5 0.44 0.23 1991 TEl-E(d) 1.5 0.62 0.32 1996 (alIn accordance with BAW-10100A and ASTM E 185-79 as modified by BAW-1543, Rev. 2.

(b) Estimated date based on 0.8 plant operation fac-tor.

(c) Capsules contain weld metal compact fracture tough-ness specimens.

(d) Capsules designated as standbys and may not be evaluated when removed.

O 10-1 a McDermott company

11. CERTIFICATION

_ The specimens were tested, and the data obtained from Davis-Besse Nuclear Power Station Unit 1 surveillance capsule TEl-B were evaluated using ac-cepted techniques and established standard methods and procedures in ac-cordance with the requirements of 10 CFR 50, Appendixes G and H.

M e/

] ~ll'TY y/ E. Lowe, J r., P.E.

Date Vroject Technical Manager This report has been reviewed for technical content and accuracy.

frC u'

c' b l "N

L. B. Gross, P.E.

Date Materials and Chemical Engineering Services l

l l

11-1 BabcockA WHees a McDermott company

~

APPENDIX A Reactor Vessel Surveillance Program --

Background Data and Information A-1 mggg 3 A4CDersnott company

1.

Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with E-185-73, are shown in Table A-1.

The loca-tions of these materials within the reactor vessel are shown in Figures A-1 and A-2.

2.

Definition of Beltline Region

=The beltline region of Drvis-Besse Unit I was defined in accordance with the data given in BAW-10100a.

3.

Capsule Identification The capsules used in the Davis-Besse Unit i surveillance program are iden-tified below by identification number, type, and location.

Capsule Cross Reference Data Number Type Location TEl-A III Upper TEl-B IV Lower TEl-C III Upper TEl-D IV Lower TEl-E III Upper TEl-F IV Lower 4.

Specimens for Determining Material Caseline Properties See Table A-2.

S.

Specimens per Surveillance Capsule See Tables A-3 and A-4.

1 NEEb l

A-2 a McDermott company

Table A-1.

Unirradiated Impact Properties and Residual Element Content -

Data of Beltline Region Materials Used for Selection of Surveillance Program Materials -- Davis-Besse Unit 1 Charpy data, CVN

Olstance, core midplane Transverse 50 35 Chemistry,sfal Material 8eltline to weld Longitudinal

ident, Material region centerline.

Drop wt, ft-Ib,

MLE, USE,

RTupi, heat No.

type location cm Tuni, F At 10F ft-lb F

F ft-lb F

Cu_

P Ni 134 50 0.04 0.007 0.009 50 61 ADB-203 SA508, C1 2 Nozzle belt 20 136,179,130 30 144 20 0.04 0.004 0.006 MJ-233 SA50e, C1 2 upper shell 8

107, %,81 118 50 0.02 0.011 0.011 50 60,62,47 27 8CC-241 SA508, C1 2 Lower shell A

47,62 59 0.14 0.011 0.007 f

25,31,35 W-232 Weld Circus seam

+198 upper (9510) 0.22 0.015 0.016 43,30,26 E-233 Weld Circus seam

+198 upper (91100)

W-182-1 Weld Circum seam

-24

-20 36,33,44 62 81 2

0.18 0.014 0.015 ' --

middle L

0.14 0.011 0.007 25,31,35 W-232 Weld Circus seam

-247 loter (12510) 0.22 0.015 0.016 43,30,26 W-233 Weld Circus seam

-247 lower (885 00) i k

P R5o 1

=

Table A-2.

Test Specimens for Determining Material Baselir.e Properties No. of test specimens Tension Material description 70F 600F(a) CVN impact Compact-fracture (b)

Heat SS Base metal Transverse direction 3

3 15 Longitudinal direction 3

3 15 Heat-affected zone (HAZ)

Transverse direction 3

3 15 Longitudinal direction 3

3 15 Total 12 12 60 Heat TT Base metal Transverse direction 3

3 15 Longitudinal direction 3

3 15 Heat-affected zone (HAZ)

Transverse direction 3

3 15 Longitudinal direction 3

3 15 Total 12 12 60 Weld metal Longitudinal direction 3

3 15 8 1/2 TCT 41 TCT (a) Test temperature to be the same as irradiation temperature.

(b) Test temperature to be determined from shift in impact transition curves after irradiation exposure.

A-4 Babcock &WHeent a McDermott company

Table A-3.

Specimens in Upper Surveillance Capsules (Designation A, C, and E)

No. of test specimens Material description Tension CVN impact Weld metal 2

12 Weld, HAZ Heat SS, transverse 12 6

Heat TT, transverse Base metal Heat SS, transverse 2

12 6

Heat TT, transverse Correlation material 6

Total per capsule 4

54 Table A-4.

Specimens in Lower Surveillance Capsules (Designations B, D, and F)

No. of test specimens 1/2 T Material description Tension CVN impact compact fracture (a)

Weld metal 2

12 8

Weld, HAZ Heat SS, transverse 12 Base metal Heat SS, transverse 2

12 Total per capsule 4

36 8

(a) Compact fracture toughness specimens precracked per ASTM E399-72.

A-5 N8h a McDermott company

Figure A-1.

Location and Identification of Materials Used in Fabrication of Reactor Pressure Vessel

(

\\

r ADB-203 (Lower Nozzle Belt)

[

W-232 (9% ID)

W-233 (91% OD) p AKJ233 (Upper Shell)

I

)

]

W-182-1 (100%)

\\

k BCC241 (Lower Shell) p

/

W-232 (12% ID)

W-233 (88% OD)

~

Dutctman j

"" E Mb A-6 a McDermott company

APPENDIX B Preirradiation Tensile Data l

l Babcock &WI8com B-1 a McDermott company

Table B-1.

Preirradiation Tensile Properties for Gase Metal Heat No. BCC241 Test Strength, psi Elongation, %

Specimen

temp, Red'n of No.

F Yield Ultimate Uniform Total area, %

i SS-601 73 75.6 91.9 12.7 27.0 67.3 SS-603 73 69.4 90.0 13.1 27.2 67.0 l

.SS-604 73 71.9 90.3 13.0 28.8 71.1 i

Mean-73 72.3 90.7 12.9 27.7 68.5 Std dev'n 73 3.12 1.02 0.21 0.99 2.29 SS-606 580 64.4 86.3 14.4 25.7 65.4 SS-611 580 64.4 86.3 13.6 26.0 63.7 SS-615 578 63.1 86.3 16.3 25.5 67.0 Mean 580 64.0 86.3 14.8 25.7 65.4 Std dev'n 580 0.75 0

1.39 0.25 1.65 Table B-2.

Preirradiation Tensile Properties for Weld Metal, WF-182-1 Strength, psi Elongation, %

l Specimen t

Red'n of No.

F Yield Ultimate Uniform Total area, %

SS-003 73 70.6 85.6 14.8 26.0 63.7 SS-007 73 69.7 85.6 15.4 27.3 64.7 Mean 73 70.2

~85.6 15.1 26.7 64.2 Std dev'n 73 0.64 0

0.42 0.92 0.71 SS-009 582 64.4 80.6 14.8 20.0 50.1 SS-015 582 67.8 83.1 11.4 17.4 49.7 SS-016 579 770.6 85.9 12.5 18.9 50.9 Mean 580 67.6 83.2 12.9 18.8 50.2 Std dev'n 580 3.10 2.65 1.73 1.31 0.61 l

""Eh B-2 x'

a A4cDermott company

..~..

i l

APPENDIX C Preirradiation Charpy Impact Data 1

f i

l l

1 l

he&Micou C-1 a McDermott company

s Table C-1.

Preirradiation Charpy Impact Data for Shell Forging Material -- Transverse Orientation, Heat BCC-241 Test Absorbed Lateral Shear Specimen

temp, energy, expansion,

fracture, t

No.

F ft-lb 10-3 in.

SS-616

-79 5.5 10 0

SS-636

-40 17.5 14 0

SS-609

-2 19.5 18 0

SS-617 0

16.5 16 0

SS-621

+21 39.0 33 2

SS-666

+40 53.0 45 15 SS-667

+40 73.0 57 20 SS-672

+40 88.0 69 60 SS-643

+70 76.0 60 25 SS-646

+70 87.0 70 25 SS-652

+74 109.0 79 85 SS-627

+106 99.0 74 80 SS-663

+130 111.5 85 90 SS-686

+171 120.0 88 100 SS-656

+213 128.5 92 100 SS-658

+278 116.0 89-100 SS-681

+338 113.5 88 100 SS-630

+585 113.0 83 100

?

C-2 Babcock &WHeon a McDermott company

Table C-2.,Dreirradiation Charpy Impact Data for Shell

' Forging Material -- Heat-Affected Zone, Heat BCC-241 Test Absorbed Lateral Shear i

Specimen

temp, energy, expansion,

fracture, No.

F ft-lb 10-3 in.

SS-331

-120 27.0 19 0

SS-307

-80 30.5 16 0

SS-309

-80 60.0 36 0

SS-310

-80 28.0 17 2

SS-325

-59 67.0 37 20 SS-346

-40 56.0 31

<10 SS-320

-20 62.0 37 25 SS-337

-20 94.0 54 30 SS-341

-2 97.5 57 60 SS-329

+40 114.5 69 40 SS-305

+74 133.0 76 90 SS-333

+106 135.5 88 100 SS-304

+130 110.5 77 100

-SS-315

+176 138.5 82 100 SS-335

+223 110.0 79 100 SS-343

+338 112.0 83 100 SS-322

+406 135.5 84 100 SS-348

+578 101.0 78-100 b'

t-

^

l 4

c NMicos C-3 a Mr.Dermott company

i

.n Table C-3.

Preirradiation Charpy Impact Data for Weld Metal, WF-182-1 Test Absorbed Lateral Shear Specimen

temp, energy, expansion,

fracture, No.

F ft-lb 10-3 in.

SS-046

-80 15.5 16 0

l SS-060

-40 16.0 15 2

SS-077

-2 37.5 35 10 SS-084

-2 28.0 27 25 SS-053 0

33.0 29 20 SS-055 0

33.5 29 15 SS-027

+40 40.0 40 50 SS-028

+40 40.0 38 35 SS-029

+40 37.5 34 15 SS-071

+70 45.5 44 50 SS-081

+70 58.0 55 70 SS-092

+74 55.0 56 75 l

SS-056

+130 70.5 64 100 SS-067

+145 36.5 35 40 SS-036'

+169 69.5 64 100 l

SS-063

+223 72.5 71 100 SS-085

+228 66.5 65 100 SS-016

+338 72.0 70 100 SS-040

+583 68.5 72 100 l

l-l-

l i

C-4 sh&WNcom a McDermott company

l i

i Figure C-1.

Impact Data for Unirradiated Shell Forging Material, Heat BCC-241 100 i

i g

l e

aM 1

~

5 y,0 l

1 l

33 l

I f

I I

I l

v.,-

1 85 { 88 42

) 88

{ 88 83{

l

. 08 g

g 3-I e

i.0s e

I W.

5 l

k e

l

"* 04-3 l

4.02 I

I I

I l

0 I

I I

I i

i DATA SIN MRY 180- T,,,

+50F

+28F T, (35 m )

c l

TCV (50 n-ts) +25F 160

+18F l

Tcy (30 n-t.s) l C -USE (avs) 127 FT-L85 y

.140 RT

+50F y

NOT

$120-

=: S l

E e

5 3100-3=

e 5 80-e W

t 1

g e

SA508.C12 40-

,g Oniantarian TRANSVERSE ~~

Fluence NONE 20 -

e Hur No.

8CC241 I

I I

I 0-00

-40 0

to 80 120 160 200 240 280 320 360 400 Testinwenarume,F C-5 Babcock &Wilcoat a McDermott company i

Figure C-2.

Impact Data for Unirradiated Shell Forging Material, Heat-Affected Zone, Heat BCC-241 W

i i

2 2

i t

i i

i 75 -

.I e

y, 2

JI25 e

t i

I i

i i

i I

I 4

. Os 3

i-r i

E 8.Os e

3 ti

" 04 -

_A.-_____

__2 5

o agnr lE I

I I

0^

I i

l i

i i

DATA SUN WlY 180. T,,7

+50F T 05 m.a) 45F y

-57F

,Tn (50 FT-ts) g

100F T, 00 n-t.s)

(-USE(avs) 130 FT-LBS 140 RT

+50F e

7 88 7 e

e 7

3 120 E

e e

I

' 4 100 I

.3so -

W t

m-e

~

navsnias, SA508.C12 p_

yg7 s,r,ttc, pws ca NOME 20 Muar no.

BCC241 I

I I

I# '

0-80

-40 0

40 80 120., 150 200 240 " s o o goo soo soo Test Toesnaruse, F C-6 Batscock&WHoom a McDermott company

\\

Figure C-3.

Impact Data for Unirradiated Weld Metal, WF-182-1 W

I I

~

i-I -- i i

i ~ l

.< 75 a

5 0

l 50 -------o-e--------------------------

w e

j m

Ji 25 -

e l

e I

I I

I 0

. 08 g

g j

i i

g g

f, a

i e.06 2

2 0,og_

5

-E e

4.02-5 a

E I

I I

O I

I I

I DATA SumARY 90-T

-20F uor

+33F T, (35 not) e gg, ICV (50 rt-ts) 45F Tey (30 FT-ts) -IlF _

C -USE (avo) 70 FT-LES y

70 S

RT

+5F 2

-,583F wer e

m

$ 60-

%o}% - - - - _ _ _ - - __________--__-------

a e

1 5

1 d@-

e V

t 30 _---

i 1

e 20 -

weld Metal MATERIAL Ca:Entation 10-FLutnet

~

HEAT No.

WF-182-3 I

I I

I 0-60

-40 0

40 80 120 160 200 240 280 320 360 400 Test Tenernarunt, F C-7 mahea,ar &WNcom a MCDermott comparty

APPENDIX D Fluence Analysis Procedures l

1 l

l l

D-1 Batacock&Wilcon a McDermott company

Analytical Method Neutron flux values at the capsule position are caculated from measured cap-sule dosimeter activities using the following equation:

A" 2

(AT-Ate-Ajt )

9 +3 = 3.7 x 104 i

Nfj oPI A"

= 6.14414 x 104 (AT-A1e-Aj t )

i fjo PI where

- A tj ),- Aj (T - Tj )

(Known as: Power Integral or l

PI =

Fj(1 - e j=1 Activity Saturation Factor) o = average cross section, barns, AT = total dosimeter activity, MCI, A1 = dosimeter activity for cycle 1, mci, t1 = time from cycle i shutdown to cycle 3 shutdown, N = Avogadro's number, A = atomic weight of target material n, n

fj = fission yield of target isotope,1.0 if not fissionable iso-e

tope, Fj = fraction of full power during jth time interval, tj, Aj = decay constant of the ith isotope, T = sum of total irradiation time, i.e., residual time in reac-tor, and wait time between reactor shutdown and counting, tj = interval of power history, Tj = cumulative time from reactor start-up to end of jth time pe-riod, i.e.,

j Tj k1 Flux values at other locations are determined using ratios which account for well established azimuthal, axial and radial variations.

Reactor vessel fluence values are calculated as the sum of cycle 1 and cycles 2 plus 3 values: with the conservative assumption that the maximum value for all cycles fall at the same point on the reactor vessel ID.

0-2

" " & WllCENE a McDermott company

. - - - ~ _..,.

The cycle 1 fluence values was obtained from the calculation for Capsule TEl-F.1 Vessel Fluence Extrapolation For up-to-date operation, fluence values in the pressure vessel are calcu-lated as described above.

Extrapolation to future operation is required for prediction of vessel life based on minimum upper shelf energy (USE) and for calculation of pressure-temperature operation curves.

Three time periods are considered:

(1) to-date operation for which vessel fluence has been cal-culated, (2) designed future fuel cycles for which PDQ calculations have been performed for fuel management analysis of reload cores, and (3) future fuel cycles for which no analyses exist.

Data from time period 1 are extrapolated through time period 2 based on the premise that excore flux is proportional to the fast flux that escapes the core boundary.

Thus for the

vessel,

e

v,C

o,C X *v,R e,R where the subscripts are defined as y = vessel, e = core escape, R = refer-ence cycle, and C = a future fuel cycle.

Core escape flux is available from PDQ output.

Extrapolation from time period 2 through time period 3 is based on the last fuel cycle in 2 having the same relative power distribution as an " equilibrium" fuel cycle.

Generally, the designed fuel cycles include l

several cycles into the future.

Therefore, the last cycle in time period 2 l

should be representative of an " equilibrium" cycle.

Data for TEl-B are listed in Table D-1.

l This procedure is considered preferable to the alternative of assuming that lifetime fluence is based on a single, hypothetical " equilibrium" fuel cycle because this procedure accounts for all known power distributions.

In addi-l tion, errors that may result from the selection of a hypothetical "equilib-rium" cycle are reduced.

l l

l L

o l

D-3 Babcock &WHeos acom,ouc - y

Table D-1.

Extrapolation of Reactor Vessel Fluence Core escape

' Cumul ative "C

flux, Time, time,.

Vessel flux, Time 2

2 Cycle n/cm -s EFPY EFPY n/cm -s interval Cumulative 1(a) 1.02 1.02 1.61(+10)-

5.19(+17) 5.19(+17) 283(a) 0.648(+14) 1.56 2.58 2.01(+10) 9.88(+17)

-1.51(+18) 4(b) 0.685(+14) 0.71 3.29 2.12(+10)(d) 4.8(+17) 2.0(+18) i 5(b) o,488(+14) 1.07 4.36 1.51(+10)(d) 5.1(+17) 2.5(+18)

X(c) 0,488(+14) 3.64 8

1.51(+10)(d) 1.7(+18) 4.2(+18)

X(c) 0.488(+14) 7 15 1.51(+10)(e) 3.3(+18) 7.6(+18)

X(c) 0,488(+14) 6 21 1.51(+10)(e) 2.9(+18) 1.0(+19)

X(c) 0.488(+14) 11 32 1.51(+10)(e) 5.25(+18) 1.6(+19)

?

j (a) Calculated.

l (b) Predicted.

]

(c) Extrapolated.

0.685 x 10M Value from x (2.01 x 1010) = 2.12 x 1010, 0.648 x 1014 (e) Cycle 5 assumed to be equilibrium cycle fur future operation.

i O

2 55D i

h 4

4

APPENDIX E Capsule Dosimetry Data E-1 Babcock &WHcom a McDermott company

Table E-1 lists the compostion of the threshold detectors and the equivalent cadmium thickness used to reduce competing thermal reactions.

Table E-2 shows capsule TEl-F measured activity per gram of target material (i.e., per gram of uranium, nickel, etc.).

Activation cross sections for the various materials were flux-weighted with a 235U fission spectrum (Table E-3).

Table E-1.

Detector Composition and Shielding Monitors Shielding Reaction 10.4% U-Al (99.27% 238 )

Cd-Ag 0.02676" Cd 238 (n,f) 0 U

1.44% Np-Al (100% 237Np)

Cd-Ag 0.02676" Cd 237Np(n,f)

Ni 100% (67.77% 58Ni)

Cd-Ag 0.02676" Cd 58Ni(n,p)58Co 59 o)

Cd-0.040" Cd 59Co(n,Y)60 o 0.66 wt % Co-Al (100%

C C

59 o(n,Y)60 o 0.66 wt % Co-Al (100% 59 Co)

None C

C 58 e)

None 54Fe(n,p)54Mn Fe 100% (5.82%

F l

\\

l l

i l

I i

l l

'T E-2 Babcock &WHcom a McDermott company

Table E-2.

Dosimeter Specific Activities Nuclide Specific

Activity, Dosimeter Post irrad.

activity activity

-uCi/g of material wt, grams Reaction Radionuclide uCi uCi/g target Dosimeter: TE1 B01 60 o 37.26 2361 357800 59 o(n,Y)

Co-Al(Bare) 0.01578 C

C 59 o(n,Y) 60C0 8.508 436.8 66180 Co-Al(Cd) 0.01748 C

Ni 0.13543 Shi(n,p) 58 o 188.5 1392 2054 C

Fe 0.15641 54Fe(n,p) 54Mn 10.64 68.03 1169 238 (n,f) 137Cs 0.06298 0.7379 7.095 238 -Al 0.08535 U

0 237Np-Al 0.06960 237Np(n,f) 137Cs 0.04056 0.5828 40.47 m

Dosimeter: TE1 BD2 Ea Co-Al(Bare) 0.01456 59Co(n,Y) 60Co 35.50 2301 348600 59 o(n,Y) 60Co 7.566 402.9 61040 Co-Al(Cd) 0.01878 C

Ni 0.13392-58Ni(n,p) 58 o 190.4 1422 2098 C

Fe 0.16241 54Fe(n,p) 54Mn 11.09 68.28 1173 238 (n,f) 137 s 0.03515 0.7653 7.359 238 -Al 0.04593 0

C U

237Np-Al 0.06833 237Np(n,f) 137Cs 0.03914 0.5728 39.78 Dosimeter: TE1 BD3 h

Co-Al(Bare) 0.01487 59 o(n,Y) 60Co 21.88 1471 222900 C

59 o(n,Y) 60 o 5.390 273.9 41500 P

Co-Al(Cd) 0.01968 C

C Ni 0.13768 58Ni(n,p) 58 c 137.5 998.7 1474 C

D l jg

-Fe 0.15176 54Fe(n,p) 54Mn 7.341 48.37 831.1 238 (n,f) 137 s 0.02710 0.5338 5.132 238 -Al 0.05077 j{

U C

0 4

A 237Np-Al 0.06876 237Np(n,f) 137 s 0.03209 0.4667 32.41 C

~

a Table E-2.

(Cont'd)

Nuclide Specific

Activity, Dosimeter Post irrad.

activity activity pCi/g of material wt, grams Reaction Radionuclide uCi uCi/g target Dosimeter: TE1 BD4 60 o 41.64 2664 403700 59 o(n,Y)

C Co-Al( Bare) 0.01563 C

59 o(n,T) 60Co 10.34 546.2 82760 Co-Al(Cd) 0.01693 C

58 o 226.0 1728 2550 Ni 0.13079 58Ni(n,p)

C 54Mn 12.93 84.63 1454 Fe 0.15279 54Fe(n p) 238 (n,f) 137 s 0.08709 0.9480 9.115 238 -Al 0.09187 U

C 0

237Np-Al 0.07075 237Np(n,f) 137 s 0.05345 0.7555 52.46 C

Footnotes - Dosimeter Material Dcta These data are the disintegration rates per gram of wire, as of 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />, July 25, 1983.

These data are the disintegration rates per gram of target nuclide: viz., 238, 237Np, 58Ni, U

54 e.

59C0, and F

The following abundances and weight percents were used to calculate the disintegration rate per gram of target element.

2380 - 10.4 wt %; 99.27% target nuclide.

237Np - 1.44 wt %; 100% target nuclide,

=

h Ni - 100 wt %; 67.77% 58Ni target nuclide.

Co - 0.66 wt %; 100% 59Co target nuclide.

N I' Fe - 100 wt %; 5.82% 54Fe target nuclide.

4N

Table E-3.

Dosimeter Activation Cross 3ections, b/ atom (a)

Energy range, U

54Fe(n,p)

?38 (n,f) 58Ni(n,p)

G_,

MeV 237Np(n,f) 1 12.2 - 15 2.323 1.050 4.830(-1) 4.133(-1) 2 10.0 - 12.2 2.341 9.851(-1) 5.735(-1) 4.728(-1) 3 8.18 - 10.0 2.309 9.935(-1) 5.981(-1) 4.772(-1) 4 6.36 - 8.18 2.093 9.110(-1) 5.921(-1) 4.714(-1) 5 4.96 - 6.36 1.541 5.777(-1) 5.223(-1) 4.321(-1) 6 4.06 - 4.96 1.532 5.454(-1) 4.146(-1) 3.275(-1) 7 3.01 - 4.06 1.614 5.340(-1) 2.701(-1) 2.193(-1) 8 2.46 - 3.01 1.689 5.272(-1) 1.445(-1) 1.080(-1) 9 2.35 - 2.46 1.695-5.298(-1) 9.154(-2) 5.613(-2) 10 1.83 - 2.35 1.677 5.313(-1) 4.856(-2) 2.940(-2) 11 1.11 - 1.83 1.596 2.608(-1) 1.180(-2) 2.948(-3) 12 0.55 - 1.11 1.241 9.845(-3) 6.770(-4) 6.999(-5) 13 0.111 - 0.55 2.34(-1) 2.432(-4) 1.174(-6) 1.578(-8) 14 0.0033-0.111 6.928(-3) 3.616(-5) 1.023(-7) 1.389(-9)

(a)ENDF/B5 value.5s have been flux-weighted (over CASK energy groups) based on a 23 0 fission spectrum in the fast energy range plus a 1/E ' shape in the intermediate energy range.

E-5 Baducck&M8ssNE a McDermott company

t APPENDIX F References l

l Babcock &WIfcom F-1 a McDermott company

1 A.

L.

Lowe, Jr.,

et al., Analysis of Capsule TEl-F Toledo Edison Company, Davis-Besse Nuclear Power Station Unit 1, BAW-1701, Babcock &

Wilcox, Lynchbitrg, Virginia, January 1982.

2 H. S. Palme, G. S. Carter, and C. L. Whitmarsh, Reactor Vessel Material Compliance With 10 CFR 50 Appendix H,

for Surveillance Program Oconee-Class

Reactors, BAW-10100A, Babcock

Wilcox, Lynchburg, Virginia, February 1975.

3 A. L. Lowe, Jr., K. E. Moore, and J. D. Aadiand, Integrated Reactor Ves-sel Material Surveillance Program, BAW-1543, Rev 1, Babcock & Wilcox, Lynchburg, Virginia, October 1983.

4 A.

L. Lowe, Jr., et al., Fracture Toughness Test Results From Capsule TEl-B. The Toledo Edison Company, Davis-Besse Nuclear Power Station Unit 1, BAW-1835, Babcock & Wilcox, Lynchburg, Virginia (to be published).

5 J. D. Aadland, R. J. Futato, and W. A. Pavinich, Babcock & Wilcox J-R Test Procedure for Compact Fracture Toughness Specimens, BAW-1808, Babcock & Wilcox, Lynchburg, Virginia, October 1983.

6 H. S. Palme and H. W. Behnke, Methods of Compliance With Fracture Tough-ness and Operational Requirements of Appendix G

to 10 CFR 50, BAW-10046A, Rev 1 Babcock & Wilcox, Lynchburg, Virginia, July,1977.

7 A. L. Lowe, Jr., et at., Irradiation-Induced Reduction in Charpy Upper Shelf Energy of Reactor Vessel Wel ds, BAW-1511P, Babcock & Wilcox, Lynchburg, Virginia, October 1980.

I' 6

g a McDermott company

v t e Letter Sequence Other
TAC:56041, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML20094J654, ML20094J664, ML20137W348
MONTHYEAR1984-08-10 [Table View]

ML20094J664

Text

SUMMARY

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