Text

Analysis of Capsule TE1-D,Toledo Edison Co,Davis Besse Nuclear Power Station Unit 1 Reactor Vessel Matl Surveillance Program
	ML20072U421
Person / Time
Site:	Davis Besse
Issue date:	12/31/1990
From:	Aadland J, Lowe A, Johari Moore; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML20072U414	List: ML20072U411; ML20072U421;
References
	BAW-2125, NUDOCS 9104190133
	Download: ML20072U421 (96)
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December 1990 I

I I ANALYSIS OF CAPSULE TEl-D THE TOLEDO EDISON COMPANY l DAVIS BESSE NUCLEAR POWER STATION UNIT 1

-- Reactor Vessel Material Surveillance Program --

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I BGW NUCLEAR I BWSERVICE COMPANY

I BAW 21/5 December 1990 I

ANALYSIS OF CAPSULE TEl-D I THE TOLEDO EDISDN COMPANY DAVIS BESSE NUCLEAR POWER STATION UNIT 1

-- Reactor Vessel Material Surveillance Program --

I I by A. L. towe, Jr., PE J. D. Aadland J. W. Moore, 111 I L. Petrusha W. R. Stagg I

I I

I B&W Document No. 77-2125-00 B&W Contract No. 579-7735 (See Section 11 for document signatures)

B&W Nuclear Service Company Engineering and Plant Services Division I P. O. Box 10935 Lynchburg, Virginia 24506-0935 I IBWit?vn?s!?;m I

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I

SUMMARY

This report describes the results of the examination of the fourth capsule of the '

Toledo Edison Company's Davis Besse Nuclear Power Station Unit I reactor vessel ;

surveillance program. The capsule was removed and examined at the end of the l sixth fuel cycle. The objective of the program is to monitor the effects of neutron irradiation on the tensile and fracture toughness properties of the reactor vessel materials by the testing and evaluation of tension, Charpy impact, and compact fracture toughness specimens. The program was designed in accordance <

with the requirements of Appendix H to 10CFR50 and ASTM Specification E185-73. '

The capsule received an average fast fluence of 9.62 x 10 18 n/cm2 (E > 1.0 Mev) I and the predicted fast fluence for the reactor vessel T/4 location at the end of '

the sixth cycle is 1.36 x 10 18 n/cm2 (E > 1 MeV). Based on the calculated fast flux at the vessel wall, an 80% capacity factor, and the planned fuel management, the projected fast fluence that the Davis Besse Unit I reactor pressure vessel inside surface will receive in 40 calendar years of operation is 1.11 x 10 I9 n/cm2 (E > 1 MeV). i The results of the tension tests indicated that the materials exhibited normal behavior relative to neutron fluence exposure. The Charpy impact data results 1 exhibited the characteristic behavior of shift to higher temperature for the 30 l

ft-lb transition temperature as a result of neutron fluence damage and a decrease in upper shelf energy. These results showed that the current techniques used for '

predicting the change in both the increase in the RT NDT and the decrease in upper '

shelf properties due to irradiation are conservative. The recommended operating period was extended to 32 effective full power years. These new operating limitations are in accordance with the requirements of Appendix G of 10CFR50.

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I I .{0NTENTS I

Page

1. INTRODUCTION ........................... 11

2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

3. SURVEILLANCE PROGRAM DESCRIPTION , . . . . , . . . . . . . . . . 3-1 j l 4. PRE-lRRADIATION TESTS . . . . . . . . . . . . . . . . . . . . . . .

4.1. Tension Tests ...... . . . . . . . . . . . . . . . . .

4-1 4 l 4.2. Impact Tests . . . . . . . . . . . . . . . . . . . . . . . . 4-1 l

5. POST-IRRADIATION TESTS . . . . . . . . . . . . . . . . . ... , . . 5-1 5.1.

I 5.2.

5.3.

Thermal Monitors . . . . . . . . . . . . . . . . , ,

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

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

. . . 5-1 52 52

)

3 5.4. Compact fracture Toughness Tests . . . . . . . . . . . . . . 5-3

6. NEUTRON FLUENCE . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 I 6.1. 6-1 l

Introduction . . . . . . . . . . . . . . . . . . . . . . . . 1 6.2. Vessel Fluence . . . . . . . . . . . . . . . . . . . . . . . 6-5 l 6.3. Capsule fluence . . . . . . . . . . . . . . . . . . . . . . . 6-5 1 6.4. Fluence Uncertainties . . . . . . . . . . . . . . . . . . . . 6-6 -!

7. DISCUSSION OF CAPSULE RESULTS . . . . . . . . . . . . . . . . . . . 7-1 7.1. Pre-Irradiation Property Data . . . . . . . . . . . . . . . . . 7-1 I 7.2. Irradiated Property Data . . . . . . . . . . ... . . . . . .

7.2.1. Tensile Properties . . . . . . . . . . . . . . . . .

'7.2.2. Impact Properties . . . . . . . . . . . . . . . . . .

7-1 7-1 7-2 7.3. Reactor Vessel Fracture Toughness . . . . . . . . . . . . . . 7-4

8. DETERMINATION OF REACTOR COOLANT PRESSURE BOUNDARY PRESSURE -

TEMPERATURE LIMITS . . . . . . . . . . . . . . . . . . . . . . . . 8-1 I

9.

SUMMARY

OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . . 9-1 I -

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I Contents (Cont'd)

Page

10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE . . . . . . . . . .

11. CERTiflCATION

, , . . 10-1 l

. . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 AppENDlXES A. Reactor Vessel Surveillance Program Background Data and information . . . . . . . . . . . , . . . . . . . . . . . . . . . A-1 B. Pre-Irradiation Tenaile Data .,,...... ...........B1 C. Pre-lrradiation Charoy impact Data ........... . . . . C-1 D.

E.

F.

Fluence Analysis Methodology Capsule Dosimetry Data . . .

References

... .................D1

. . . . . . . . . . . , . . . . . . . . E-1

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1 l

I List of Tables 3-1. Specimens in Surveillance Capsule TEl-A . , , . . . . . . . . . . 3-2 3-2. Chemical Composition and Heat Treatment of Surveillance Materials . . . . . . . . . . . . . . . . . . . . . 3-3 5-1. Tensile Properties of Capsule TEl-D Irradiated Base Metal 3 and Weld Metal . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5 5-2.

Charpy Impact Irradiated Data to 9.62 x 10for gpsule n/cm TEl D Base Metal, Heat BCC241,2 5 4 (E > 1 MeV) 5-3.

54.

Charpy impact Data for Capsule TEl-03 geat-Affected Zone Metal, Heat BCC241, Irradiated to 9.62 x 10 n/cm2 (E > 1 MeV) . . . .

Irradiated Charpy Impac Data ror Capsule TEl-D Weld Metal 5-4 l Irradiatedto9.62x10}8 n/cmg (E > 1 MeV) . . . . . . . . . . . 5-5 6-1. Surveillance Capsule Dosimeters . . . . . . . . . . . . . . . . 6-6 6-2. Davis Besse Unit 1 Reactor Vessel f ast flux . . . . . . . . . . . 67 6-3. Calculated Davis Besse Unit 1 Reactor Vessel Fluence ..,,,.6-8 E 6-4 Surveillance Capsule TEl-D fluence, flux, and DPA . . . . . . . . 6-9 5 6-5. Estimated Fluence Uncertainty . . . . . . . . . . . . . . . . . . 6-9 7-1. Comparison of Tensile Test Results ............... 77 7-2. . Summary -of Davis Besse Unit 1 Reactor Vessel Surveillance Capsules Tens il e Tes t Resul ts . . . . . . . . . . . . . . . . . . . . . . . 7-8 7-3.

Observed Properties -Vs. Predicted 9.62 x 10 3 n/cmghangeg(In E > 1 MeV) irradiated

. . .Charpy

. . . . . impact

.... 7-9 I-

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j Jables (Dnt'd)

Table Page 7 4. Summary of Davis Besse Unit 1 Reactor Vessel Surveillance 1 Capsules Charpy Impact Test Results ............... 7-10 7 5. Evaluation of Reactor vessel End of Life Fracture Toughness and Pressurized Thermal Shock Criterion ............. 7 11 1 7 6. Evaluation of Reactor Vessel End of Life Upper Shelf Energy ... 7 12

81. Data for Preparation of Pressure Temperature Limit Curves for Davis Besse Unit 1 Applicable Through 32 Ef PY . . . . . .... B-5 i A-1. Unirradiated Impact Properties and Residual Eletent Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials -- Davis Besse Unit 1 . . . . . . . A-3 I A 2.

A-3.

Test Specimens for Determining Material Baseline Properties Specimens in Upper Surveillance Capsules (Designations A C, and E) . . . . . . . . . . . . . . . . ...........

... A4 A5 A 4. Specimens in Lower Surveillance Capsules (Designations B, D, and f) . . . . . . . . . . . . . . . . . . . . . . . . . . . A5 B 1. Pre irradiation lensile Properties of Shell Plate Material, Heat BCC 241 ...........................B2 B-2. Pre 1rradiation Tensile Properties for Weld Metal WF 182 1 ....B2 0 1. Pre lrradiation Charpy impact Data for Shell forging Material - Transverse Orientation, Heat BCC 241 . . . . . . . . . . C 2 I C 2.

C-3.

Pre trradiation Charpy Inipact Data for Shell forging Material Heat Affected Zone, Heat BCC 241 . . . . . . . . . . . . C 3 Pre Irradiation Charpy Impact Data for Weld Metal WF 1B2-1 . . . . C-4 D-1. Normalization Factor .......................D7 D-2. Davis Besse Unit 1 Reactor Vessel fluence by Cycle ........D8 E-1. Detector Composition and Shielding . . . . . . . . . . . . . . . . E-2 E 2. Measured Specific Activities (Unadjusted) for Dosimeters in I E-3.

Capsule TEl D . . . . . . . . . . . . . . . . . . . . . . . . . . . E-2 Dosimeter Activation Cross Sections, b/ atom . . . . . . . .....E3 I list of Fiaures figure 3 1. Reactor Vessel Cross Section Showing Location of Capsule I 3 2.

5-1.

TEl D in Davis Besse Unit 1 . . . . . . . . . . . . . . . . . . .

Loading Diagram for Test Specimens in TEl-D , . . . . . . . . . .

Impact Data for Irradiated Shell Forging Material, Heat BCC 241.

34 35 56 5-2. Impact Data for Irradiated Shell forging Material, Heat Affected I 5 3.

Zone, Heat BCC 241 . . . . . . . . . . . . . . . . . . . . . . .

Impact Data for Irradiated Wold Metal, WF 182-1. . . . . . . . .

5-7 58 6-1. General Fluence Determination Methodology . ........... 62 h -v-I BW!!?vML%

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I Fioures (Cont'dl figure Page 6 2. Fast flux, Fluence, and DPA Distribution Through Reactor Vessel Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 10 6-3. Azimuthal Flux and Fluence Distributions at Reactor Vessel inside Surface . . . . . . . . . . . . . . . . . . . . . . . . . 6 11 8 1. Predicted Fast Neutron fluence at Various locations Through Reactor Vessel Wall for 32 EFPY - Davis Besse Unit 1 .......B6 8 2. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation - Heatup, Applicable for first 32 EFPY - Davis Besse Unit 1 ............... 87 8-3. Reactor Vessel Pressure Temperature Limit Curves 3 for Normal Operation Cooldown, Applicable for 5 First 32 EFPY Davis Besse 1 . . . . . . . . . . . . . . . . . . . C-8 8-4. Reactor Vessel Pressure Temperature Limit Curves for Inservice Leak and Hydrostatic lests, Applicable for First 32 EFPY Davis Besse Unit ! ............. 89 A 1. Location and Identification of Materials Used in Fabrication of Reactor Pressure Vessel ....................A6 C-1. Impact Data for Unirradiated Shell Forging Material, Heat BCC-241 ...........................C5 C 2. Impact Data for Unirradiated Shell Forging Material, Heat Affected Zone, Heat BCC 241 .................C6 C 3. Impact Data for Unirradiated Wold Metal, WF 1821. . . . . . . . . C 7 D-1, Rationale for the Calculation of Dosimeter Activities and Neutron Flux in the Capsule . . . . . . . . . . . . . . . . . . D 9 D-2. Rationale for the Calculation of Neutron Flux in the Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . . D 10 D 3. Plan View Through Reactor Core Midplene 3 (Reference R 0 Calculation Model) . . . . . . . . . . . . . . . . . D 11 B I

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1. INTRODUCTION This report describes the results of the examination of the fourth capsule of the I 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 sixth f uel cycle (5.45 EfPY). The first capsule of the program was removed and evaluated af ter the first year of operation; the results are reported in BAW 1701.I The second capsule of the program was removed ad evaluated af ter the third fuel cycle (2.58 EfPY); the results are reported in BAW 1834. The third capsule was removed and evaluated after the (curth fuel cycle (3.33 EfPY);

the results are reported in BAW 1882.

The objective of the program is to monitor the effects of neutron irradiation on the tensile and impact properties of reactor pressure vessel materials under actual operating conditions. The surveillence program for Davis Besse Unit I was designed and furnished by Babcock & Wilcox (B&W) as described in BAW 10100A.

I The program, designed in accordance with the requirements of 10CFR50, Appendix 5

H and ASTM Specification E185 736 , is being conducted in accordance with BAW-8 15437 and ASTM specification E185 82 and was planned to monitor the effects of neutron irradiation on the reactor vessel materials for the 40 year design life, i.e., 32 effective full power years (EFPY), of the reactor pressure vessel. The future operating limitations established after the evaluation of the surveillance capsule are also in accordance with the requirements of 10CFR50, Appendixes G9 and HS . The recommended operating period was extended to 32 EFPY as a result of

( the fourth capsule evaluation.

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I il I 2. BACKGROUND I

The ability of the reactor pressure vessel to resist fracture is the primary 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 properties of such low alloy ferritic steeis as SA508, Class 2, used in the fabrication of the Davis Besse Unit I reactor vessel, are well characterized and documented in the literature.

1he low.611oy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a correspond-ing decrease ia ductility after irradiation. The most significant mechanical l property change in reactor pressure 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 10CFR50, " fracture Toughness Requirements," specifies minimum l fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of water-cooled power reactors, and provides specific guidelines for determining the pressure-temperature limitations on operation of the RCPB. The toughness and operational requirements are specified to provide adequate safety margins during I any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix G to 10CFR50 became effective on August 13, 1973, the requirements are applicable to all l boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on the effective date.

I 21 I 13WfteWE % r l I

I Appendix H to 10CFR50,

Reactor Vessel Materials Surveillance Program Requirements," defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of water cooled reactors resulting from exposure to neutron irradiation and the thermal environment, fracture toughness test data are obtained f rom material specimens withdrawn periodically f rom the reactor vessel. These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life.

A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code, Section 111,

" Nuclear power Plant Components." This method utilizes fracture mechanics concepts and the reference nil ductility temperature, RTNDT, which is defined as g the greater of the drop weight nil ductility transition temperature (per ASTM g E 208) or the temperature that is 60f below that at which the material exhibits 50 f t lbs and 35 mils lateral expansion. The RT ND1 of a given material is used l

to index that material to a reference stress intensity f actor curve (K curve),

lR which appears in Appendix G of ASME Section 111. 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 g K

IR curve, allowable stress intensity f actors can be obtained for this material 5 as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

The RTNDT 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 materials. The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored 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 increase in the Charpy V-notch 30 f t-lb temperature is added to the original RT NDT to adjust it 22 I

I amam- g

I for radiation embrittlement. This adjusted RTg) is umi to index the material i

to the K jg curve which, in turn, is used to set operating limits for the nuclear lE P **" "" '""'* "'" "*"' '' '"' ""' 'h' " "'d"" "

W the reactor vessel materials, iI

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3. SURVEILLANCE PROGRAM DESCRIPTION l The surveillance program comprises six surveillance capsules designed to monitor the effects of neutron and thermal environment on the materials of 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 I 3-1. The six capsules, originally designed to be placed two in each holder tube, are positioned near the peak axial and azimuthal neutron flux. However, with the use of Davis Besse Unit I as one of the irradiation sites of the 177-fuel-assembly integrated reactor vessel material surveillance program, the capsules l are irradiated on a schedule integrated with the capsules of the other participating reactors. This integrated schedule is described in BAW 1543. BAW-g 10100A includes a full description of the capsule design.

Capsule TEl D was removed during the sixth refueling shutdown of Davis Besse Unit

1. This capsule contained Charpy V-notch impact test specimens fabricated from one base metal (SA508, Class 2), a weld metal, and heat affected zone material.

Tensile specimens were fabricated from one base metal and the weld metal only, I in addition, there are compact fracture specimens fabricated from the weld metal.

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. lhe chemical composition and heat treatment of the surveillance material l in capsule TEl D are described in Table 3 2.

All test specimens were machined from the 1/4-thickness (1/4T) location of the l forging material. Charpy V-notch and tension test specimens from the vessel material were oriented with their longitudinal axes perpendicular to the I

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principal working direction of the forging. Capsule TEl D contained dosimeter '

wires, described as follows:

Dosimeter Wire lhieldina V Al alloy Cd Ag alloy Np Al alloy Cd Ag alloy Nickel Cd Ag alloy 0.66 wt % Co Al alloy Cd Ag alloy 0.66 wt % Co Al alloy None fe None Thermal monitors of low-melting metals and alloys were included in the capsule.

The metals and alloys and their melting points are as follows:

Allov Meltina Point. E 90% Pb, 5% Ag, 5% Sn 558 g

97.5% Pb, 2.5% Ag 580 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Cadmium 610 Lead 621 Table 3-1. Snecimens in Surveillance Capsule TEl-D Material No. of Test Specimens Material Description Identity lension ,CVN Impact 1/2T CT Weld Metal WF 182-1 2 12 8 Weld, HAZ Heat SS, Transverse BCC241 -

12 -

Base Metal Heat SS, Transverse BCC241 2 12 :.

Total Per Capsule 4 36 8 32 lBWE?Ar??SUnv l

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E Tab 1_e 3-2. Chemical Composition and Heat Trea.tment of Surveillante Materials Chemical Analysis Heat Weld Me Element BCC 241(a) WF-1821g' C 0.22 0.09 Mn 0.63 1.70 P 0.011 0.014 I S 51 fii 0.011 0.27 0.81 0.013 0.42 0.63 Cr 0.32 0.15 1 Mo 0.63 0.40 Cu 0.02 0.21 Heat Treatment Heat tio, lets_f T_ime. h Coolina BCC241 1640110 4 Water quenched Water quenched I 1590110 4 1240110 5 Air cooled 1125125 15-l/2 Furnace cooled g WF-182 1 1125125 15 l/2 Furnace cooled (a)Percertified Materials Test Reports (b)PerLicensing Document BAW 150010 5

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4. PRE-IRRADIATION TESTS I 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 10CFR50. l 4.1. Tension Tests I

l 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 I 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 0.0$0 inch per minute. A 4-pole extension device with a strain gaged extensometer was used to determine the 0.2% yield point. Test conditions were in accordance with the applicable requirements of ASTM A370 72.II For each 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 pre-irradia-tion tensile specimens are given in Appendix B.

4.2. Imoact Tests I Charpy V notch impact tests were conducted in accordance with the requirements of ASTM Standard Methods A370 72 and E23-72 I2 on an impact tester certified 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 from -85 to +550F. Specimens I

41 I BWlt%MW I

3 were removed from the baths and positioned in the test frame anvil with tongs specifically designed for the purpose. The pendulum was released manually, l

j allowing the specimens to be broken within 5 seconds from their removal from the g

i temperature baths.

Impact test data for the unirradiated baseline reference materials are presented

, in Arpendix C. Tables C-1 through C-3 contain the basis data that are plotted in figures C 1 through C 3.

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I 5. POST-lRRADIATION TESTS I l 5.1. Thermal Monitors I Capsule TEl D contained three temperature monitor holder tubes, each containing I

five fusible alloy wires with melting points ranging from 558 to 621F. All the I thermal monitors at 558, 580, and 588F had melted while those at the 610F f

location showed no signs of melting of slumping; the monitor at the 621F location i melted in all three holder tubes, it was heretofore assumed that the 610F and l 621F monitors were placed in the wrong locations in the holder tubes, and based !

on these observations, it was concluded that the capsule had been exposed to a l

peak temperature in the range of 610 to 621F during the reactor operating prriod.

l In the case of TEl-D the original loading diagram was consulted. This drawing ;

lists the five materials used in the monitors, and showed the position in which g each wire was loaded. Both show the lead wire (621F melting point) to be in the fourth position, with the cadmium wire (610F melting point) in the fifth position.

This supports the observation that the 610F monitor did not melt while the 621F I monitor did melt. It is believed that the lead wire sof tened (and presented the appearance of melting) due to long term exposure to elevated temperatures, which were not sufficient to melt the cadmium wire. Therefore, it is probable that the capsules was exposed to temperatures in excess of 588F but not as high as 610F, and that this was sufficient to cause the lead wires to slump and appear to have melted.

These peak temperatures are attributed to operating transients that are of short durations as described in BAW 2040 I3 and are judged to have insignificant effect on irradiation damage. Short duration operating transients cause the use of I

51

,I l 13 W fte W a h r I

B thermal monitor wires to be of limited value in determining the nsximum steady state operating temperature of the surveillance capsules; however, it was l

calculated that the maximum steady state operating temperature of specimens in the capsule was held within +25f of the 1/4T vessel thickness location temperature of 561 as described in BAW 2040, it is concluded that the capsule design temperature may have been exceeded daring operating transients but not for times and temperatures that would make the capsule data unusable.

5,2, Tensien Test ResuMJ The results of the postirradiation tension tests are presented in Table 51. g Tests were performed on specimens at both room temperature and 550 using the same B test procedures and techniques 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, The results of the pre irradiation tension tests are presented in Appendix B, 5,3. Charny V-Notch Impact Test Results l

The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material are presented in Tables 5-2 through 5-4 and Figures 51 through 5 3. The test procedures and techniques were the same as those used to test the unirradiated specimens (Section 4.2). The data show that the materials exhibited a sensitivity to irradiation within the values to be expected from their chemical composition and the fluence to which they were exposed.

Scatter in the heat-affected zone material normally prevents a serious interpretation of the data. Although the data contained in this capsule results g appear to follow a smooth trend, the data at 550f is significantly below the upper-shelf trend and, therefore, must be classified as abnormal scatter, The results of the pre irradiation Charpy V-notch impact tests are given in Appendix C.

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I 5.4 Cemnact Fracture Touchness Tests The compact fracture toughness specimens fabricated from the weld metal, which i

were a part of the capsule specimen inventory, were tested by a recognized single j specimen J-integral testing procedure. The results of the testing of these specimens is reported in a separate report, BAW 2128.I4 I l l Table 5 1. Tensile Properties of Capsule TEl-D Irradiated Base Metal and Weld Metal l

4 g Red'n, 3 Specimen Test Temp, Sin, ngl.h . n s i [10ng11.ipm 3; in Area, No. F Yield Ultimate Uniform 19111 9;. , _

Base Metal, Transverse, 9.62 x 10 18 n/cm2 (E > 1 MeV) 55-612 70 73,800 95,200 10 25 61 55 605 550 69,500 91,900 9 22 58 '

l Weld Metal, 9,62 x 10 I0 n/cm2 (E > 1 MeV) 55 008 70 87,300 103,300 10 25 56 I SS-005 550 78,100 94,800 9 18 48 I

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Table 5 2. Charpy Impact Data for Capsule 310 Bgse Metal, Heat BCC24), Irradiated to 9.62 x 10 I

n/cm (E > 1 MeV)

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Specimen Energy, Expagsion, fracture, No. F ft b 10 in. %

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SS 677 550 103.5 87 100 g

Values used to determine upper shelf energy value per ASTM E185.

Table 5 3. CharpyImpactDataforCapsuleTElDgestA{fectedZoneMetal, 3 Heat BCC241, Irradiated to 9.62 x 10 n/cm (E > 1 MeV)

E Absorbed Lateral Shear Specimen Test Temp., Energy, Expagsion, fracture.

No, f ft-lb 10 in. %

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Values used to determine upper-shelf energy value per ASTM E185.

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' Absorbed Lateral thear Specimen lest Temp., Energy, lxpagston, fracture, j No. f ft lb 10 in. %

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' Values used to determine upper shelf energy value per ASTM E185.

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6. NEUTRON FLUENCE 6.1. Introduction The neutron fluence (time integral of flux) is a quantative way of expressing the cumulative exposure of a material to a pervading neutron flux over a specific I period of time. Fast neutron fluence, defined as the fluence of neutrons having j

energies greater than 1 MeV, is the parameter that is presently used to correlate :

radiation induced changes in mateM61 properties. Accordingly, the fast fluence must be determined at two locations: (1) in the test specimens located in the j surveillance capsule, and (2) in the wall of the reactor vessel. The former is l- used in develtwing the correlation between fast fluence and changes in the

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material properties of specimens, and the latter is used to ascertain the point g

of maximum fluence in the reactor vessel, the relative radial and azimuthal distribution of the fluence, the fluence gradient through the reactor vessel I wall, and the corresponding material properties.

The accurate determination of neutron flux is best accomplished through the I simultaneous consideration of neutron dosimeter measurements and analytically derived flux spectra. 00simeter measurements alone cannot be used to predict the fast fluence in the vessel wall or in the test specimens because (1) they cannot measure the fluence at the points of interest, and (2) they provide only j

rudimentary information about the neutron energy spectrum. Conversely, reliance on calculations alone to predict fast fluence is not prudent because of the g length and complexity of the analytical procedures involved, in short, measurements and calculations are necessary complements of each other and together they provide assurance of accurate results.

Therefore, the determination of the fluence is accomplished using a combined analytical-empirical methodology which is outlined in figure 61 and described I in the following paragraphs. The details of the procedures and methods are pre-sented in general terms in Appendix 0 and in BAW 1485p.15 I 61

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I floure 6 1. General fluence Determination Methodoloav l

MEASURDiDRS OF NEMRON NMLYTICAL DETGMDATION OF 3 00SDETm ACTIVITIES DOSIMETER ACTIVITIES NO 5 NElfTRON FLUX I'

ADJUSTED ENERGY DEPDOENT NEMRON 3 FLUX 5 I

REACTOR OPERATING NEIITRON HISTORY No PRE-R.UDOE DICTED FUTlRE a OPERATION 5 Analytical Determination of Dosimeter Activities and Neutron flux I

The analytical calculation of the space and energy dependent neutron flux in the h test specimens and in the reactor vessel is performed with the two dimensional ;

discrete ordinates transport code, D011V.16 The calculations employ an angular quadrature of 48 sectors (58), a third order leGendre polynomial scattering II approximation (p3), the CASK 23E cross section set with 22 neutron energy groups and a fixed distributed source corresponding to the time weighted average power distribution for the applicable irradiation period.

In addition to the flux in the test specimens, the 00TIV calculation determines

, the saturated specific activity of the various neutron dosimeters located in the surveillance capsule using the ENDF/B5 dosimeter reaction cross sections.I8 The saturated activity of bach dosimeter is then adjusted by a factor which corrects

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for the fraction of saturation attained during the dosimeter's actual (finite) irradiation history. Additional corrections are made to account for the following effects:

o Photon induced fissions in V and Np dosimeters (without this correction the results underestimate the measured activity).

e fissile impurities in V dosimeters (without this correction the results underestimate the measured activity).

e Short half life of isot epes produced 4 iron and nickel dosimeters (303-day Mn 54 and 71 day Co48, respectiyOy). (Without this correction, the results could be biased high or low depending on the long term versus short term power histories.)

Measurement of Neutron Dosimeter Activities I The accuracy of neutron fluence predictions is improved if the calculated neutron flux is compared with neutron dosimeter measurements adjusted for the effects I noted above. The neutron dosimeters located in the surveillance capsules are listed in Table 61. Both activation type and fission type dosimeters were used.

The ratio of measured dosimeter activity to calculated dosimeter activity (M/C) is determined for each dosimeter, as discussed in Appendix D. These M/C ratios are evaluated on a case by-case basis to assess the dependability or veracity of h

each individual dosimeter response. Af ter carefully evaluating all f actors known l to affect the calculations or the measurements, an average M/C ratio is calculated and defined as the

normalization factor." The normalization factor is applied as an adjustment f actor to the DOT-calculated flux at all points of interest, llqutron fluence The determination of the neutron fluence from the time averaged flux requires only a simple multiplication by the time in EFPS (effective full-power seconds) over which the flux was averaged, i.e.

f 4) (AT) = 3 4 4)g AT 9

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I where g fluence at (i,j) accumulated over time AT (n/cm'),

f g) (61) g = Energy group index, Time average flux at (1,j) in energy group g, (n/cm' sec),

g)g al e Irradiation time, EFPS.

tieutron fluence was calculated in this analysis for the following components over the indicated oMrating time:

l Test Specimens: Capsule irradiation time in EfPS Reactor Vessel: Vessel irradiation time in EfPS Reactor Vessel: Maximum point on inside surface extrapolated to 32 effectite full power years The neutron exposure to the reactor vessel and the material surveillance specimens was also determined in terms of the iron atom displacements per atom of iron (DPA). The iron DPA is an exposure index giving the fraction of iron atoms in an iron specimen which would be displaced during an irradiation. It is l

considered to be an appropriate damage exposure index since displacements of atoms from their normal lattice sites is a primary source of neutron radiation damage. DPA was calculated based on the AS1'i Standard E693-79 (reapproved 1985).I9 A DPA cross section for iron is given in the ASTM Standard in 641 energy groups. DPA per second is determined by multiplying the cross section at a given energy by the neutron flux at that energy and integrating over energy.

DPA is then the integral of DPA per second over the time of the irradiation. In the DPA calculations reported herein, the ASTM DPA cross sections were first collapsed to the 22 neutron group structure of CASK-23E; the DPA was then l

determined by summing the group flux times the DPA cross section over the 22 g energy groups and multiplying by the time of the irradiation.

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l l 6.2. Vessel Fluence The maximum fluence (E > 1 MeV) exposure of the Davis Besse Unit I reactor vessel during Cycles 16 was determinta tb be 2.44 x 100 n/cm' based on a maximum neutron flux of 1.42 x 10 n/5lm2 -s (Tables 6 2 and t ,1). The maximum fluence j g occurs at the cladding / vessel interface at an azimuthal location of approximately 11 degrees from a major horizontai Akis of the cert.

Fluence data were extrapolated to 32 EFPY of operation based on two assumptions:

(1) the future fuel cycle o;a' rations do not differ significantly from their I current designs, and (2) the latest chhulated (or extrapolated) flux remains constant from that time through 32 UPt. The extrapolttion was carried out in two states, (1) frot EOC 6 to [00 7, and (2) from EOC 7 to 32 EFPY. In the first I stage, Cycle 7 fuel design information is used to calculate assembly averaged relative power distributions. These power distributions are used with DOT adjoint factors to determine the average fast flux for Cycle 7. In the second stage, the 32 EfPY fluence was calculated by assuming a constant flux over the l period which was equal to the average flux for cycle 7.

Relative fluence and DPA (displacement per atom) as a function of radial location l in the reactor vessel wall is shown in figure 6 2. Reactor vessel neutron fluence lead factors, which are the ratio of the neutron flux at the clad inter-face to that in the vessel wall at the T/4, T/2 and 3T/4 locations, are 1.79, 3.54, and 7.28, respectively. DPA lead factors at the same locations are 1.59 2.64, and 4.57, respectively. The relative fluence as a function of azimuthal I angle is shoe in figure 6 3. A peak occurs in the fast flux ([ > 1 HeV) at about 11 degrt with a corresponding value of 1.42 x 10'0 n/cm' s.

6,3. Capsule Fluence I The capsule was irradiated for 1989.8 EFPD in the top holder tube position during Cycles 1-6 of Davis Besse located 26.9' degrees off the major horizontal axis at about 202 cm from the vertical axis of the core. The cumulative fast fluence at the center of the surveillance capsule was calculated to be 9.62 x 10 a n/cm'.

1 This fluence value represents an average value for the center location of the 1 65 I

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I Charpy specimens in the capsule. It includes an axial peaking factor in the capsule of 1.14 and a normalization factor of 1.03. ,

6.4. Fluence Uncertainties Uncertainties were estimated for the fluence values reported herein. The results l are shown in Table 6 5 and are based on comparisons to benchmark experiments, when available; estimated and measured variations in input data; and on ll The values in Table 6 5 represent best estimate values engineering judgement.

g!

based on past experience with reactor vessel fluence analyses.

Surveillante.Dngle Dosimeters l!I Table 6-1.

L* er Energy limit for Isotope Half Life Dosimeter Reactions (a) 54fe(n,p)54Mn Reaction, MeV 2.5 312.5 days I

5BNi(n.p)58Co 2.3 70.85 days l 1.1 30.03 years 238U (n.f)l370s 237Np(n,f)I37Cs 0,5 30.03 years (a) Reaction activities measured for capsule flux evaluation. I I

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I l Iable 6 2. Dtvis Besse Unit 1 Reactor Vessel fast flux I f ast flux (E > 1 MeV), n/cmi -s Inside Surface flux n/cm' s (E > 0.1 McV) inside Surface Cycle (Max Lont.inal T/4 1/2 3T/4 (Max location)

Cycle 1 1.61E+10 8.9E+9 4.4E+9 2.0E+9 3.4E+10 (374 EFPD)

Cycles 2 6 1.38E+10 7.71E+9 3.90E+9 1.90E49 2.88E410 (1615.8 EfPD)

Cycle 7** 1.03E+10 0.58E+10* 0.29E+10* 1.41E+9*

(390 EFPD) l 8 EFPY 15 EfPY 1.03E+10 1.03E410 0.58E+10*

0.58E410*

0.29E+10*

1.41E+9*

1.41E+9' 21 EfPY 1.03E410 0.58E+10* 0.29E410* 1.41E+9* l 1

32 EfPY 1.03E+10 0.58E410* 0.29E410* 1.41E49* i

Divide flux at inside surface by the appropriate lead factors on p. 6 8 to obtain these T/4, T/2, and 3T/4 fast flux values.

- Assumed cycle length of 390 EfPD for flux extrapolation for Cycle 7.

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I Table 6 3. Calculated Davis Besse Unit 1 Reactor Vessel Fluence Cumulative Inside Surface Fast Fit" t re, n/cm' (E > 1 MeV) l Irradiation Time (Max location) T/4 T/2 3T/4 End of Cycle 1 5.19E+17 2.9E+17 1.4E+17* 6.8E+16 (374 EFPD)

End of Cycle 6 2.44E+18 1.36E+18 6.89E+17 3.35E+17 (1989.8 EFPD)

End of Cycle 7 2.79E+18 1.56E+18* 7.88E+17* 3.83E+17* l (2379.8 EFPD) 8 EFPY 3.27E+18 1.83E+18* 9.24E+17* 4.49E+17*

l 15 EFPY 5.55E+18 3.10E+18* 1.57E+18* 7.62E+17*

21 EFPY 7.50E+18 4.19E+18* 2.12E+18* 1.03E+18*

32 EFPY 1.llE+19 6.20E+18* 3.14E+18* 1.52E+18*

Calculated using these 1.0 1.79 3.54 7.28 lead factors g Conversion Factors Fluence (E > 1 MeV) 1.45E-21** 1.63E-21** 1.94E-21** 2.31E-21**

to DPA 2

- Multiply fast fluence values (E > 1 MeV) in units of n/cm by these factors to obtain the corresponding DPA values.

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I Table 6 4. Surveillance Cansule TEl-D Fluence. Flux and DPA I Irradiation Time flux (E >21 MeV),

n/cm -s Fluencp, n/cm DPA DB1, Cycles 1-6 5.60E+10 9.62E+18 1.40E-2 (1989.8 EFPD)

! Table 6-5. Estimated Fluence Uncertainty I Estimated Calculated Fluence Uncertainty Basis of Estimate In the capsule 1 15% Activity measurements, cross section fission yields, satu-I ration factor, deviation from average fluence value in the reactor vessel i 21% Activity measurements, cross at maximum location for sections, fission yields, fac-cycles 1 through 6 of tors, axial f actor, capsule Davis Besse Unit I location, radial / azimuthal ex-trapolation, normalization I factor In the reactor vessel i 23% Factors in vessel fluence above I at the maximum location for vnd of-life extra-plus uncertainties for extra-polation to 32 EFPY polation I

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I O.100 -

l I '

O.137 DPA h 0.00 I = 7.28 1

I

> 0.08 Te - I I I I i y 0.07 -

i ;

Q I

, 0.00 -

I l

c I e > 1.0 Mev 4

s 0.05 -

1/4T I

I 1/2T I

3/4T I

I l E 0.04 -

223.01 cm 228.37 cm 233.72 cm 0.03 ' ' ' '

215 220 225 230 235 240 245 Distance From Core Center (cm) 6-10 I

I an==_ l

- M M M M - K _ M M . W W M M Fioure 6-3. Azimuthal Flux and Fluence Distributions at Reactor Vessel Inside Surface 1.10 1.08 -

1.06 -

1.04 -

1.02 -

1.00 0.98 -

_5 0.96 u.

? e 0.94 -

.?

j O.92 -

E O.90 -

0.88 -

0.86 -

0.84 -

D O.82 -

E p O.80 I

d N@ O.78 -

"E I I I I ' I I I I

![ O.76 O 5 10 15 20 25 30 35 40 45 E

s Degrees Off Major Axis

I I

7. DISCUSSION OF CAPSULE RLSULTS 7.1. Pre-1rradiation Property Data A review of the unirradiatec properties of the reactor vessel core beltline region materials indicated no significant deviation from expected properties except in the erse of the upper shelf properties of the weld metal. Based on the predicted end t -service peak neutron fluence value at the 1/4T vessel wall location and the copper content of the weld metal, it was predicted, using Regulatory Guide 1.99, Rev. 2 methodology, that the end of-service Charpy upper shelf energy (USE) will be below 50 ft-lb. However, plant specific material I surveillance data show this value will remain above 50 ft-lbs.

This weld was selected for inclusion in the surveillance program in accordance I with the criteria in effect at 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 I 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 I ductility are within the limits observed for similar materials. There is some strengthening, as indicated by increases in ultimate and yield strengths and 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 550F in the properties of the weld metal are larger than those observed for the base metal, indicating a greater sensitivity of the weld metal

-to irradiation damage. In either case, the changes in tensile properties are insignificant relative to the analysis of the reactor vessel materials at this period in service life.

I 7-1 I l

1 l

A comparison of the tensile data from the first three capsules (Capsules TEl-F, TEl A, and TEl-B) with the corresponding data from the capsule reported in this report is shown in Table 7 2. The currently reported capsule experienced a fluence that is 3/4 that of the previous capsule and five times greater than the first capsule.

The general behavior of the tensile properties as a function of neutron irradiation 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 exhibited 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 compares the observed changes in irradiated Charpy impact properties with the predicted changes.

The 30 ft-lb transition temperature shift for the base metal is not in good agreement with the value predicted using either Regulatory Guide 1.99, Rev. 220 or Rev. 2 plus margin. It would be expected that these values would exhibit better agreement when it is considered that the data used to develop Regulatory Guide 1.99, Rev. 2, was taken at the 30 ft-lb temperature.

The transition temperature measurements at 30 ft-lbs for the weld metal is in good agreement with the results using Regulatory Guide 1.99, Revision 2, but is in poor agreement with the results using Regulatory Guide 1.99, Revision 2 plus margin. At the 30 ft-lb level, the shift relationship to the predicted value demonstrates that the estimating curves of Regulatory Guide 1.99, Rev. 2 are conservative for predicting the 30 ft-lb transition temperature shifts.

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, Rev. 2. The less than ideal comparison may be attributed to the spread in the data of the material was based only on the 30 ft-lb transition data ccmbined 7-2 l

g:

sw==_ g 1 -

I g with a small number of data points to establish the irradiated curve. Under 5 these conditions, the comparison indicates that the estimating curves in RG 1.99 for low-copper materials are not conservative for predicting the 50 f t-lb transition temperature shifts.

l The data for the decrease in Charpy USE with irradiation showed good agreement l with predicted values for the base metal and only fair agreement for the weld metal. However, a good comparison of the measured data with the predicted value is not expected in view of the lack of data for low , medium , or high-copper content materials at these fluence values that were used to develop the estimating curves, A comparison of the Charpy impact data from the first, second and third capsules (Capsules TEl-F, TEl-A and TEl-B) with the corresponding data from the capsule l reported in this report is shown in Table 7-4. The currently reported data l experienced a fluence that is 3/4 that of the previous capsule and five times ;

greater than the first capsule. I 1

'I The base metal exhibited shif ts at the 30 f t-lb level for the latest capsule that were greater than those of the second capsules and less than that recorded for l

the third capsule. The corresponding data for the weld metal showed about a 20%

increase at the 30 ft-lb level.

Both the base metal and the weld metal exhibited decreases in tne upper shelf I

values comparable to that observed in the previous capsules, These data confirm t that the upper shelf drop for this weld metal may have reachet a stabilized condition (saturation) as observed in the results of capsules evaluawi by others. This lack of further decrease in Charpy USE drop for this weld meta' should not be considered indicative of a similar lack of decrease of upper shelf region fracture toughness properties. This behavior indicates that other i reactions may be taking place within the material besides simplo neutron damage.

Verification of this relationship must await the testing and evaluation of the data from conpact fracture toughness test specimens, I

I l 7-3 I BW!!nn?a!Lr I

I Results from other surveillance capsules also indicate that RT NDT estimating curves have greater inaccuracies than originally thought. These inaccuracies are a function of a number of parameters related to the basic data available at the g

time the estimating curves are established. These parameters may include a inaccurate fluence values, poor chemical composition values, and variations in data interpretation. The change in the regulations requiring the shif t measurement to be based on the 30 f t lb value has minimized the errors that result from using the 30 f t-lb data base to predict the shif t ft-lbs.

behavior at 50 l

The design curves for predicting the shif t will continue to be modified as more data become available; until that time, the design curves for predicting the RT NDT shift as given in Regulatory Guide 1.99, Revision 2, are considered adequate for predicting the RT NDT shift f those materials for which data are not available. These curves will be used to establish the pressure-temperature g operational limitations for the irradiated portions of the reactor vessel until W the time that new prediction curves are developed and approved.

The lack of good ag eement of the change in Charpy USE is further support of the inaccuracy of thn prediction curves as included in Regalatory Guide 1.99, Revision 2. Ahhough the prediction curves are conservative in that they generally predict a larger decrease in upper-shelf energy than is observed for a given fluence and copper content, the conservatism can unduly restrict the operational limitations. These data support the contention that the USE drop l

carves will have to be revised as more reliable data become available; until that time the design curves used to predict the decrease in USE of the controlling materials for licensing applications are considered conservative.

7.3. Reactor Vessel Fracture Touahness An evaluation of the reactor ssel end of-life fracture toughness and the g

pressurized thermal shock criterion was made and the results are presented in Table 7-5.

I 7-4 BW!!An?a%m I

B

I The fracture toughness evaluation shows that the controlling weld metal may have I an end-of-life RT DTof N 216f based on Regulatory Guide 1.99, Revision 2. This predicted shift is excessive since the latest capsule (Lapsule TEl D) exhibited I a measured shif t of 150F for a fluence of 9.62 x 10 18 n/cm2 , Ratioing this measured shift to the T/4 wall location fluence, it is estimated that the end of-life RTNDT shift will be significantly less than the value predicted using i Regulatory Guide 1,99, Revision 2. This reduced shift permits the calculation l of less restrictive pressure temperature operating limitations than if Regulatory Guide 1,99, Revision 2, was used.

I The pressurized thermal shock evaluation demonstrates that the Davis Besse reactor pressure vessel is well below the screening criterion limits and, therefore, need not take any additional corrective action as required by the regulation, An evaluation of the reactor vessel end-of-life upper shelf energy for each of the materials used in the fabrication was made and the results are presented in I Table 7-6. This evaluation was made because the weld metals used to fabricate the reactor vessel are Linde 80 flux, l ow-upper-shel f-energy, relative high copper and are expected to be highly sensitive to neutron radiation damage. Two I methods were used to evaluate the radiation induced decrease N upper shelf energy. The method of Regulatory Guide 1,99, Revision 2, which is the same procedure as used in Revision 1, and the method presented in BAW-1803 2I which was developed specifically to address the need of an estimating method for this class of weld metals.

The methods of both Regulatory Guide 1,99, Revision 2, and BAW-1803 show that none of the materials used in the fabrication of the reactor vessel will have an upper-shelf energy below 50 f t-lbs through 32 EFPY design life based on the T/4 wall location. Regulatory Guide 1,99 method predicts a decrease below 50 ft-lbs for the controlling weld metal at the vessel inside wall, Contrary to this prediction, the actual weld metal surveillance data does not support such a large I decrease in upper-shelf energy.

I I 7-5

,I I

I Based on the Davis Besse surveillance data and the prediction techniques presented in BAW-1803, it is concluded that none of the reactor vessel material upper-shelf energies will decrease to below 50 f t-lbs during the vessel design life.

I I

I 7-6 I

I BWeina?%%r g

=

I Table 7-1. Comnarison of Tensile Tes' Rtsults lI Elevated Temp.

Room Temo. Teit_ Test (550F)

Unirr IIf_ad Unirr irred Base Metal -- ANK 191. Transverse fluence, 10 I8 n/cm2 (E > 1 MeV) 0 9.62 0 9.62 Ultimate tensile strength, ksi 90.7 95.2 86.3 91.9 0.2% yield strength, ksi 72.3 73.8 64.0 69.5 Total elongation. % 28 25 26 22 RA, % 68 61 65 58 Weld Metal -- WF-182-1

~

Fluence, 10 18 n/cm2 (> 1 MeV) 0 9.62 0 9.62 I Ultimate tensile strength, ksi 85.6 103.3 83.2 94.8 0.2% yield strength, ksi 70.2 87.3 67.6 78.1 Total elongation, % 27 25 19 18 RA, % 64 56 50 48 I

I I

I 7-7 I BW!! nevi?e%~r I

Table 7-2. Summary of Davis 8 esse Unit i Reactor Vessel Surveillance Capsules Tensile Test Results Strenath, ksi Ductility. %

Test , Total Reduction Fjgence 2 Yield %(,' Elon. %{.) of Area %(.I Material 10 n/cm Temp, F Ultimate %( )

90.7 72.3 -

28 -

68 -

8ase Metai 0 73 -

86.3 64.0 -

26 -

65 -

(BCC-241) 580 -

95.6 + 5.4 75.0 + 3.7 26 -?I 66 - 2.3 1.96 70 l 580 88.8 + 2.9 66.3 + 3.6 22 -15.4 59 - 9.2 l

5.92 76 91.1 + 0.4 70.1 - 3.0 26 - 7.1 65 - 4.4 87.5 + 1.4 66.9 + 4.5 21 -19.2 57 -14.0 580 95.2 + 5.0 73.8 + 2.1 25 -10.7 61 -10.3 9.62 70

-10.8 550 91.9 + 6.5 69.5 + 8.6 22 -15.4 58 i

69 96.4 + 6.3 74.7 + 3.3 25 -10.7 65 - 4.4 7 12.90 0.0

' 580 92.4 + 7.0 72.2 +12.8 23 -11.5 65 70.2 27 64 -

85.6 Weld Metal 0 73 -

83.2 67.6 -

19 -

50 -

(WF-182-1) 580 -

98.1 +14.6 82.5 +17.5 25 - 7.4 58 - 9.4 l.96 70

+15.8 48 - 4.2 580 90.0 + 8.2 73.1 + 8.1 21 110.9 +17.8 85.5 +21.8 16 -40.7 54 -15.6 5.92 76

-21.0 42 -16.0 580 93.9 +12.8 77.8 +15.1 15 D 9.62 70 103.3 +20.7 87.3 +24.4 25 - 7.4 56 -12.5 gg 550 94.8 +13.9 78.1 +15.5 18 - 5.3 48 - 4.0

$E 104.1 +21.6 88.8 +26.5 23 -14.8 53 -14.1 12.90 69 R$ 550 96.4 +15.9 79.4 +17.5 17 -37.0 49 - 2.0

80 b

(*IChange relative to unirradiated.

l EW W W W M M M M M M M M M M m m m m M l

1.

I Table 7-3. ObservedVs.PredictedChangesinIrygdiateg I Charpy impact Properties - 9.62 x 10 n/cm Predicte Predicted I Material Observed RG1.99/2p#I RG 1.99/2+M(b)

Increase in 30 ft-lb Trans. Temo.. F Base material (BCC-241)

Transverse 3 20 40 Heat-affected zone (BCC-241) 101 20 40 Weld metal (WF-182-1) 150 167 223 Increase in 50 ft-lb Trans. Temo.. F Base Material (BCC-241)

Transverse 30 20 40 Heat-affected zone (BCC-241) 67 20 40 Weld metal (WF-1821) 149 167 223 Decrease in Charov USE. ft-lh Base material (BCC-241)

Transverse 5 12 N.A.

Heat-affected zone (BCC-241) 7 12 N.A.

Weld metal (WF-1821) 16 25 N.A.

(a)Per R.G. 1.99, Revision 2, May 1988.

(b)Per R.G. 1.99, Revision 2, May 1988, shift plus 2 x margin.

I I

.I 7-9 I BW!!nMa % v I

l

. Table 7-4. Su;nmary of Davis Besse Unit 1 Reactor Vessel Surveillance Capsules Charpy Impact Test Results Transition Temperature Increase. F Upper Shelf Enerny. ft-lb Fgence 2 30 ft-lb Predicted (a) Predicted (b) Predicted IC) Predicte

Material 10 n/cm Observed RG 1.99/2 w/o M RG 1.99/2 w/M BAW-1803/1 Observed Fredictef RG 1.99/2 'I BAW-1803/1{C) ,

Base Metal 1.96 Neg. 12 24 N.A. 113 113 N.A.

(BCC-241) 5.92 Neg. 17 34 N.A. 113 111 N.A. ,

9.62 3 20 40 N.A. 117 110 N.A.

12.90 28 22 44 N.A. 118 108 N.A.

HAZ Metal 1.96 43 12 24 N.A. 118 115 N.A.

, (BCC-241) 5.92 57 17 34 N.A. 105 113 N.A.

9.62 101 20 40 N.A. 117 112 N.A.

12.90 34 22 44 N.A. 111 110 N.A.

Weld Metal 1.96 127 96 152 80 62(d) 53 64 (WF-182-1) 5.92 125 144 200 140 55(d) 48 61 9.62 150 167 223 166 46 60 1 12.90 175 181 237 182 54(d) 54 44 59 7 '

o (*)Per Regulatory Guide 1.99, Revision 2, dated May 1988.

IDI Per Regulatory Guide 1.99, Revision 2. plus margin.

i ICI Per BAW-IS03. Revision 1, dated October 1990, plus margin.

(d) Upper-shelf energy value re-defined per ASTM E185.

D E5 t

I P.

d l

i l

k g

l L 1 s l

1 i

! EM M M M M M M M M M M M M M M M M M

M M M M M M M M W W W W W W W M M M M Table 7-5. Evaluation of Reactor Vessel End-of-Life fracture Toughness and Pressurized Thermal Shock Criterion Estimated Material Chemical E0t Flu 3nce T/4 Wall EOL RTNDT' I PTS Evaluation. F Composition Inside T/4 Wall Material Descrfotion Surface locatien Per Per E Screeng Heat Copper, Nickel, j RG 1.99/2 BAW-1083/1 RT p73 Criterion Reactor Vessel w/o n/cd n/cm Beltline Region location Number Type w/o N.A. 101 270 2.64 Ele 1.59E18 86 ADR-203 5A508 C1.2 0.04 0.68 Nozzle Belt N.A. 81 270 1.11E19 6.67E18 65 AKJ-233 5A508, C1.2 0.04 0.77 Upper Shell 105 270 56 N.A.

SA503 C1.2 0.02 0.81 1.11E19 6.67E18 Lower Shell BCC-241 N.A. 169 300 2.64EI8 -- N.A.

WF-232 Weld 0.18 0.64 Upper Circum. Seam (109%) N.A. 300 1.59E18 168 208 WF-233 Weld 0.29 0.63 -

Upper Circum. Seam (0D91%) 300 216 167 211 0.24 0.63 1.11E19 6.67E18 WF-182-1 Weld Mid. Circum. Seam (100%) 70 300 N.A. N.A.

0.18 0.64 1.34E16 --

WF-232 Weld Lower Circum. Seam (1012%) N A. 300 8.05E15 Neg. Neg.

WF-233 Weld 0.29 0.68 --

tower Circum. Seam (0088%)

(a)Per Regulatory Guide 1.99, Revision 2 dated May 1988.

(b)Per Regulatory Guide 1.99. Revision 1, dated April 1977.

ICIP er 10CFR50, Section 50.61, Fracture Toughness Requirements for Frotection Agait.tt Pressurized Thermal Shock E it consistant with Regulatory Guide 1.99, Pevision 2, May 1988 (proposed revision published in Federal Register. V 1989).

D E

=

EE 53 iP d

a I

Table 7-6. Evaluation of Reactor Vessel End-of-Life Upper Shelf Enerav Material Estimated Estimated E Chemical _(OL Fluenqt,__ PerRG1.99/2gtg USE EstimatedE0gfjSE Per BAW-1803 Estimated EFPY Material Description . Composition Inside I/4 Wall Initial to 50 ft-lbs Reactor Vessel Heat Copper. Nickel, Surfa7 Locatign USE. Inside T/4 Wall Inside 1/4 Wall at T/4 Wall l

Oeltline Region Location Number Type w/o w/o n/cm n/cm ft-lbs Surface Location Surface Location RG :.99/2 l

Nozzle Belt ADB-203 SA508, C1.2 0.04 0.68 2.64E18 1.59E18 133 IC) 121 122 N.A. N.A. >32 Upper Shell AKJ-233 SA508, 01.2 0.04 0.77 1.IIE19 6.67E18 140 ICI 12C 124 N.A. N.A. >32 tower Shell BCC-241 SA508, C1.2 0.02 0.81 1.llE19 6.67E18 Il8 IC) 105 107 N.A. N.A. >32 Upper Circtri. Seam (109%) WF-232 Weld 0.18 0.64 2.64E18 --

(70)(D) 53 -

67 --

>32 Upper Circum. Seam (OD91%) WF-233 Weld 0.29 0.68 --

1.59e1B (70)(b) -

50 --

58 >32 81 IC)

I Mid. Circum. Seam (100%) WF-182-1 Weld 0.24 0.63 1.IIE19 6.67E18 49 53 58 59 >32 tower Circum. Seam (ID12%) WF-232 Weld 0.18 0.64 1.34E16 --

(70)(b) N.A. -

N.A. --

>32 Lower Circum. Seam (0088%) WF-233 Weld 0.29 0.68 --

8.05E15 (70)(b) -

N.A. --

N.A. >32 l I")Per Regulatory Guide 1.99, Revision 2. dated May 1983.

(b)Per BAW-1803 Revision 1 dated October 1990.

ICI Per Certified Materials Test Reports.

N EE F

sE M

EF mE 2"

uns uma sus aus amm aus aus aus sus sus sus aus aus sus sus aus sus

_ _ _ _ _ _ _ _ _ . s u m _ __u

__ _ m a. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

I I

I I 8. DETERMINATION OF REACTOR COOLANT PRESSURE BOUNDARY 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 recuirements of I 10CFR50, Appendix G. The methods and criteria employed to establish operating pressure and temperature luits are described in topical report BAW 10046A.24 The objcctive of these limits is to prevent nonductile failure during any normal I operating condition, including anticipated operation occurrences and system hydrostatic tests. The loading conditions of interest include the following:

I 1. Normal operations, including heatup and cooldown.

2. Inservice leak and hydrostatic tests.

g

3. Reactor core operation.

The major components of the RCPB have been analyzed in accordance with 10CFR50, Appendix G. The closure head region, the reactor vessel outlet noule, and the l beltline region have been identified as the only regions of the reactor vessel (and consequently of the RCPB) that regulate the pressuretemperature limits.

Since the 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

.I 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 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 RT HDT f 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 I 8-1

- "" " ~

I

I limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained through a point by-point comparison of the limits impused by the closure head region, the outlet nozzle, and the beltline region.

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

pressures.

The limit curves for Davis Besse Unit I are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of the thirty-second EFPY, The thirty second EFPY was selected because it is estimated that the fourth surveillance capsule will be withdrawn at the end of the refueling cycle when the estimated capsule fluence corresponded to appro-ximately the inside surface end of-life value. The time difference between the withdrawal of the third and fourth surveillance capsule provides adequate time for re establishing the operating pressure and temperature limits for the period of operation beyond the current limits.

The unirradiated impact properties were determined for the surveillance beltline l

region materials in accordance with 10CFR50, Appendixes G and H. For the other beltline region and RCPB materials for which the measured properties are not l

available, the unirradiated impact properties and residual elements, as originally established for the beltline region materials, are listed in Table A-1. The adjusted reference temperatures are calculated by adding the predicted radiation-induced RT NDT and the unirradiated RT NDT. The predicted RT NDT is calculated using the respective neutron fluence and copper and nickel contents.

Figure 8-1 illustrates the calculated peak neutron fluence at several locations through the reactor vessel 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 l

conservative. The design curves of Regulatory Guide 1.99, Rev. 2, were used to predict the radiation-induced RT NDT values as a function of the material's copper and nickel content and neutron fluence.

-The neutron fluences and adjusted RT NDT values of the beltline region materials at the end of the thirty-second full-power year are listed in Table 8-1. The 82 I

asu==_ ,

I neutron fluences and adjusted RT values are given for the 1/4T and 3/4T vessel NDT wall locations (T = wall thickness). The assumed RT NDT of the closure head region and the outlet noale steel forgings is 60F, in accordance with BAW-10046. l The chemistry factors for the controlling metals (WF 1821) in the beltline region was recalculated in accordance with the procedures described in Regulatory I Guide 1.99, Revision 2, Regulatory Position 2.

The data used to calculate a new chemistry factor for weld metal WF 182-1 was I obtained from the B&WOG Integrated Reactor Vessel Surveillance Program. The data !

for the weld metal WF-182-1 which has the weld wire Heat No, 821T44. A summary of the available data is as follows.

Capsule Weld Metal Fluence, n/cm 2 RTNDT,F Reference TEl-F WF-182-1 1.96E+18 127 1 TEl-B WF-182 1 5.92E+18 125 2 TEl-D WF-182-1 9.62E419 150 (This report)

TEl-A WF-182-1 1.29E+19 175 3 The analysis of these data produced a new chemistry factor for WF 182-1 of 162, Similarly, the data used to calculate a new chemistry factor for weld metal WF-233 was obtained from the B&WOG IRVSP (Integrated Reactor Vessel Surveillance Program). The data for weld metal WF-233 has the weld wire Heat No. T29744. A summary of the available data is as follows.

Capsule Weld Metal Fluence,n/cm 2 tRTNDT,F Reference I Ko-Ri 1-V WF-233 4.67E+18 191 22 l Ko-Ri 1-T Ko-Ri 1-5 WF-233 WF-233 1.08E+19 1.21E+19 187 222 22 22 The analysis of these data produced a new chemistry factor for WF-233 of 207.

!I 83 lI BWeinEY%%r

I Figure 8 2 shows the reactor vessel's pressure-temperature limit curve for normal heatup. This figure also shows the the core criticality limits as required by 10CFR50, Appendix G. Figures 8-3 and 8-4 show the vessel's pres-g sure-temperature limit curve for normal cooldown and for heatup during inservice 5' leak and hydrostatic tests, respectively. All pressure-temperature limit curves are applicable up to the thirty-second EFPY, Protection against nonductile failure is ensured by maintaining the coolant pressure below the upper limits of the pressure-temperature limit curves. The acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not permitted to 90 critical until the pressure-te-mperature combinations are to the right of the criticality limit curve. To establish the pressure temperature limits for protection against nonductile g

failure of the RLPD. the limits presented in Figures 8 2 through 8-4 must be 3 adjusted by the pressure dif ferential 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.

I I

I 8-4 13Wifeefa h r Il El

l

EEM m m m W ~

mW W W W W M M M. .

Table 8-1. Data for Preparation of Pressure-Temperature Limit Curves for Davis Besse Unit 1 - Applicable Throuah 32 EFPY Radsatson ladored Adjest ed W *DI pf

. f._cr_e aldmeet (etation _ _ _ Inside i c< at een

$gr (gr a "(t) at(Waf gt (m f MatertaI Mtdplane item " ' I U "P** "

t hami st ry 1s t t ial JZ (fPV f _JZ,(fff F hittiaa to Weld eatar Avis Wald I/4 flua* U_

_ _ Identifio tton_ _.

I_'.' I '#"" ""

' I *EE_

" DIE'_ I'1'_" _"$"ye """***O****'*"#

UNUf ' ' 2 '_._* "' "* ' 3"__

+50 I9I il 0 7/4 TO 60 2.64f l8 0.04 0 68 26 ADR 203 5A508. (1 2 b rile t*ft 0 04 0 77 26 +20 I9I 23 16 12/8 55 44 1.11t l9

+50 M A8J 233 SA508. (1 2 Upper shell 18 12 9/6 1T 68

- 1.11f 19 0.02 0.81 20 I4C 241 SA508. ( 1 2 Ic *r shell 6.64 161 6 83 51 69/65 14 *> 150

+f99 % 2 64ffB 0.18 WT 232 ASA/tinde 80 Upper r irtue weld (50 9%)

0.68 2*4 6 W 105 64 69/69 168 177

198 Yes 2.64118 0.29 Wf 233 ASAftsade 60 tipper cir(no wid (00 9tt) -

+I I9I 158 109 56/56 (Zi6)td);ggrytd)

Tes 1. 8 t( 19 0.24 0 63 178 WV 182 1 A9/t ende 80 Middle g irtue w*ld (100%) 24 -

0 64 161 6

-24T No I . 34f l6 0.18 m Wf-232 ASAftende PO t er ctr<um w*ld (TD 121) 0 68 -6 N -

I.34fl6 0 29

{ lit l i'll129l(# }

ves h WF 23) ASAft inde 80 t ower c trttne weld (00 884 ~247 162 +f i9' 444 99 28/28 af 6bn<sh6f t of limiting weld metal tiased 1.11(19 WF-182 1 A54/lind* S0 (alculated use of serve e tapsele data per Segulatory Codde en 1.99. Sevas ten 2. 9 9vlatery pos6tten 2.1. e ICT 65 34/34 IIS 93 2.64ff8 207 WI 233 shif t of iteltine w*1d **tal t>awd on A9/linde to Calculated RIese af servesbate < apsule dat a par Regaf atory Guide 1.99 p*vtsten 2. 9 9wtatory position 2.1

$*IEf t altulated per 9eggiatery Calde 1.99. Perision 2. dat ed May 19'M.

We st emat*d initial af, of weld omtals per BAW-j90), Sevision I, Ortoter 1990: One Standard Dentation - 70f II and PMW 1799. July 1983.24 I'IMaterials d e=ical r upesitions per BAW 1820. December 1984 j (d)lnitial f alculated Bi g valves f or ese in c alculation of p**ssera temperature Itails.

I'I Rf y

values eswd in t alf ulat e pressure-tear *rature limits.

Maa,n t er vessel . alt tht< 6 ness - 8 5 in< hes .

per f art 6fied Material f est Report s.

EE sE na

%C "O

$2 o

2n R

( .

Figure 8-1. Predicted Fast lieutron Fluence at Various locations Through Reactor Vessel Wall for 32 EFPY - Davis Besse Unit I 1.25

' ' 1 X 10'* nicm' Se 2 1.00 ~

^

E o 0.75 ~

C 6.20 X 10's n/cm' l

?'

2 O G*

v- \s ci O.50 - sot l o 60 c \

e yes*e\ "' aOO" 3.14 X to'* n/m>

2 4 tcc u-0.25 -

eg W agg T I 8 Vess 1.52 X 10's n /cm's D

g Vessel Wall T /2 Location Z

Vessel Wall 3/4T Location Q , ' I y ' I '

0 g O 4 12 16 20 24 8 32

$E N EFPY 8D 25 f*

x M M m g , M M m

m M M M M M MM M -

W 'm M M M M Figure 8-2. Reactor Vessel Pressure-Temperature limit Curves for Normal Operation -

Heatup. Applicable for First 32 EFPY - Davis Besse Unit 1 2400 Assveed RIssi T ' " " ' " ' ' ' ' "

H J 2200 -

Beltline tegles 1/41 174 a 450 73 Beltline Region 3/4T 129 s 450 130

.9 2000 - closure mens regnee 60 c 500 175 M%

to Outline terrie 60 5 675 Q.-

E 675 730 t T 751 7 35 O 1800 -

t 1307 320 !

$m a

I 7750 1100 355 340 m 160o - J 250 M5 8

Q, The acceptence pressure-teeperature ceabinettons are below and to the right of the llelt cerve(s), the IIelt cerves de est _

1400 - ll'8' *** $$ di*r'atist between the pelat et syste.

cm E pressere senserement and the pressere es the reacter vessel reglen centrolling the llelt carve, ser de they laclede any additleast G h -d **'d ' Y "' " ' ' "-

o 1200 -

o iy O Criticality

-c 1000 -

Applicable for Heatup Limit

$ Rates up to 50 F/h j 800 -

u F O -

g 600 -

D E Q

=

tr 400 -

=

A B

c E,E e2 9 200 -

d NE, i i ge o i I i g5 50 100 150 200 250 300 350 400 45n .

4 x Reactor Vessel Coolant Temperature. F

Figure 8-3. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation -

Cooldown. Applicable for First 32 EFPY - Davis Besse 1 2400 Folet Pressure, psig feep., T Assumed ifRDI, T E

30s 70 settilee regles 1/41 174 a 2200 Beltline Region 3/4T 129 8 450 150 625 19 5 Closure Head reglos 60 C u e 1503 7n 8'"I" sonle0 CD 2000 [ 7250 g Q. The acceptable pressore-temperature combinations are below and to the

                                     ' * ' " "" "' "( 5 ) ' "' " " " "' " 5 d' " ' '  ' 'd' O

1800 r'e's'u'r' p s e dif f erential between the point of systes pressere seasurement and the pressure en the reactor vessel region contrelling 7 the 11 11 curve, nor do they include say additiesal margin of

</>

u> 1300 3,,,,, g,, possible instrument error. - O u 1 1400 - E D m m o 1200 - m O O

      -       1000   -

Applicable for Cooldown

       $                                                                                     - Rates up to 100F/h to 270F o      800   -

and then 50F/h u 600 - C e B D $ 400 - Cooldown Rates EE A

==

200 - Up to SOF/hr Up to 100F/hr

 %e                                       '              '                8                '            '         '

O 300 350 400 lh 35 50 100 150 200 250 I

Reactor Vessel Coolant Temperature. F m m m m m e em e em a e e g

Figure 8-4. Reactor Vessel Pressure-Temperature Limit Curves for Inservice Leak and Hydrostatic Tests, Applicable for First 32 EFPY - Davis Besse Unit 1 2600 Assumed RINDi' I Foint Pressure, psig ? esp., I F CD Beltline Region 1/41 17% A 407 70 3 Beltline Region 3/%I 179 8 (17 5 150 a 2200 - closure nead Region 0 c 675 705

                                  .                  Outline Rozzle           60        0              845         710

2000 _

t 1544 78 0 3 T YA0 5\0 M Ihe acceptance pressure-temperature combinations are below and to the o 1800 - right of the tielt corve (s). The limit cerves do not include the pressere g y differential between the point of systes pressere seasurement and the Q. 1600 - pressure oa the reactor vessel regina controlling the limit cerve, nor do y they include any additional margin of safety for possible instrement error. C cc 1400 -

           ,                   3                                                                                                 .-E a                    0   1200         -

O Applicable for Heatup and

                               'i$  1000         -                                                                                   Cooldown Rates up to g                                                                                                    50F/h (<50F in any 1/2h

800 -

D Period) . o 600

                                -                                                    B                     C O

m 400 -

A Q~

g 200 - EE ' ' ' ' ' '

me= , 0 sE 50 100 150 200 250 300 350 400 R$ 89 Reactor Vessel Coolant Temperature. F d n

I 1 I 9.

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in the third surveillance I capsule, TEl D, removed for evaluation as part of the Davis Besse Unit 1 Reactor Vessel Surveillance Program, led to the following conclusions:

1. Tha capsule received an average fast fluence of 9.62 x 10 18 n/cm2 (E >

1.0 MeV) . The predicted fast fluence 18 3 location at the end of the sixth fuel cycle foristhe reactor 1.36 x 10 vesse} n/cm E> (T/4 E 1 MeV).

2. The fast fluence of 9.62 x 10 18 n/cm2 (E > 1 MeV) increased the RT of the capsule reactor vessel core region shell materials a maximum $I 150F.

I 3. Based on the calculated fast flux at the vessel wall, an 80% capacity factor and the planned fuel management, the projected fast fluence that the Davis receive Besse in 32 EFPYUnit I reactor operation pressure is 1.11 x 10 gesse} g n/cm (inside surface will E > 1 MeV).

4. The increase in the RT for the shell forging material was not in good agreement with thayDhredicted by the currently used design curves l 5.

of RT NDT versus fluence (i .e. , R.G.1.99/Rev. 2) . The increase in the RT for the weld metal was in good agreement with that predicted by thMDdurrently used design curves of RT . versus

   -I                 fluence (i.e., R.G.1.99/Rev. 2) and the prediction tech $ dues are conservative.

6. values decreased for 32 EFPY because of a decrease in the The RT[dh estima EOL fluence values and are well below the PTS screening criteria.

7. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change in weld metal Charpy upper-shelf properties due to irradiation are conservative.

I 9-1 I B W !!nin # L r I

Il

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

9. 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.

I' l Ii ! I, I I I l I I I I I I l I 9-2 awu==- I a

1 I 1I I 10 *i e .utANCE CAPSULE REMOVAL SCHEDULE I Based on the postirradiation test results of capsule TEl-D, the following I schedule is recommended for examination of the remaining capsules in the Davis Evaluation Schedule (a) Estimated Vessel F Capsule 10} gence'2 n/cm Estimated Date Estimatedgpsulg ID Fluence, 10 n/cm Surface 1/4T Data Available(b) TEl-C(c) 1.9 0.24 0.14 Removed /In Storage TEl-E(C) 1.7 0.42 0.24 1996 (a)ln accordance with BAW-10100A and ASTM E 185-79 as modified by BAW-1543A, Rev. 3, September 1989. (b) Estimated date based on 0.8 plant operation factor. (c) Capsules designated as standbys and may not be evaluated when removed. I I I I 10-1 I [3WElNNESEmy I

l 'I I 11. CERTif! CATION I The specimens were tested, and the data obtained from Davis Besse Nucicar Power Station Unit I surveillance capsule TEl D were evaluated using accepted I techniques and established standard methods and procedures in accordance with the requirements of 10CfR50, Appendixes G and H.

                                                           <      f PLT        F/Oc/P70 li. L. Lowe"Jr., f.f,7                           0;.t e I                                    Project Technical Manager This report has been reviewed for technical content and accuracy.

W ha hlb I4 Det HkD I M. J. De/an (Material Analysis) M&S uni" Date OMM b Mdwe3 /9 $EC 3O L. B. Wimmer (fluence Analysis) Date Perfo ante Anal s Unit 19 h. Jo K. K. Yoon, P.E. (Fr ure Analysis) Date M&SA Unit Verification of independent review. k110 Q() .s J V ]N6 Y0 K.' E. Moore,' Manager Dat'e M&SA Unit This report has been approved for release. l se' AK T. L. Baldwin w~ss Date Program Manager I I 11-1 I < swame_

l I 'I I I I I I I APPENDlX A I Reactor Vessel Surveillance Program Background Data and Information I . I I I I

  .I I

A1 I awanua-

i i I

1. fig.1.erial Selection Data l 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-l. The locations of these materials within the reactor vessel are shown in figure A-1,

2. Definition of Beltline Reaion The beltline region of Davis Desse Unit I was defined in accordance with the data E

!                    given in BAW 10100A.                                                                                                                    E

3. Capsule Identification The capsules used in the Davis Besse Unit I surveillance program are identified below by identification number, type, and location.

I D.nu.1e Cross Refgrann_Egla l fiumber TEl A lyne 111 l l TEl B IV , TEl C 111 TEl D IV TEl-E 111 g i TEl F IV

4. Specimens for Determinina Material Baseline Properties See Table A 2.

5 Specimens er Surveillance Caniple See Tables A 3 and A-4. I I . A-2 I

                                                                                                                              /3WfisN h r                    g i

M M M M M M M M M M M M M M M M M M M Table A-1. Unirradiated Impact Properties and Residual Elemont Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials -- Davis-Besse Unit 1 Distance. Charry Data. CYM Core Midplane Iraasverse Material Beltilne to Weld 50 35 Chemistry. % Ident. Material Region Centerline. Drop wt. tgmiludinal ft-Ib. MtE. USE. RT Heat No. Type cm FT* C. p s ni location Tg7. F At IDF ft-Ib F F ft-Ib ADB-203 SA508. CI 2 Nozzle belt -- 50 -- 61 -- 134 50 0.04 0.007 0.009 -- AKJ-233 SA508. C1 2 Upper Shell 8 -- 20 136.179.130 30 -- 144 20 0.04 0.004 0.006 -- 107.95.81 BCC-241 SASC3, C1 2 tower Shell A -- 50 60.62.47 27 -- 118 50 0.02 0.011 0.011 -- 47.62.59 WF-232 Weld Circum seam +193 - 25.31.35 -- -- - -- 0.14 0.011 0.007 -- p tmper (9% ID) WF-233 Weld Circum seal +198 -- 43.30.26 -- -- - -- 0.22 0.015 0.016 -- tepper (91% 00) WF-182-1 Weld Circus sers - 74 -20 36.33.44 62 -- 81 2 0.18 0.014 0.015 -- , Middle WF-232 Weld Circum seam -247 -- 25.31.35 -- -- - -- 0.14 0.011 0.007 - lower (12% ID) WF-233 Weld Circum seam -247 -- 43.30.26 -- -- - -- 0.22 C.015 0.016 -- lower (88% 00) D E RR

     %E 55 on 4

.I I I lable A 2. Test Stecimens for Determinino Material Baselirtt Propert ies No, of Test Specimens _ ! Tension , N Material Description 70f 600F CVN 1mpact Compact fracture (b)

                                                         !!t01 }_$

Base Metal Transverse Direction 3 3 15 -- u longidutinal Direction 3 3 15 - Heat Affected Zone (HAZ) Transverse Direction 3 3 15 --

Longitudinal Direction Total J

12

                                                                                                                                                              ,J 12 li 60
                                                                                                                                                                                                          =

l Heat TT Base Metal Transverse Direction 3 3 15 -- I. Longitudinal Direction 3 3 15 - Heat-Affected Zone (HAZ) Transverse Direction 3 3 15 - 3

longitudinal Direction ,J J 15 = 5 Total 12 12 60 -

Longitudinal Direction 3 3 15 8 1/2 TCT E 4 1 TCT E (8)Testtemperature to be the same as irradiation temperature. (b)Testtemperature to be determined from shif t in impact transition curves af ter irradiation exposure. l 1 i A-4 I I l 13WffsMahr g-l

I I Table A 3, Specimens in Upper Surveillance Capsules (Eesionations A. C and E) I .fiterial Descrio_ tim t No. of Test SntcisinL lension CVN Ironi Weld Metal 2 12 Weld, HAZ I Heat SS, Transverse 12 Heat T1, Transverse - 6 Base Metal I Heat SS, Transverse 2 12 Heat T1, Transverse - 6 Correlation Material ;, _0 Total per Capsule I 4 54 Table A 4, Specimens in Lower Surveillance Capsules l- . (Dnignations B. D. and F) I Material Description Tension No. of Test Specirnens CVN Impact 1/21 Compact fracture (a) Weld Metal 2 12 8 Weld, HAZ Heat SS, Transverse - 12 - l Base Metal Heat SS, Transverse - 2 12 Total Per Capsule 4 36 8 (a)Compactfracture toughness specimens procracked per ASTM E399 72. I I A-S I BWlinnM% 1

.I

I ! Figure A-1. Location and Identification of Materials Used in iat!rication of Reetter Pressure Vessel l I

I t

i 1 ( I T ( , l

                                        ~ ADB-203 (Lover Nozzle Belt)

T - kT-2 32 (9% 1D) kT-233 (91% OD) I p ATJ233 (Upper Shell) I [: kT-192-1 (100%)

                                     - BCC241 (Lover Shell)

I I

              /                           kT-232 (12, 1D)               I kT-233 (8-8% OD) j Dutchman I

I A6 I awuma- y

i lI j lI !I !l I APPENDlX B Pre Irradiation Tensile Data I I . I B-1 g B W !!?vE H %

Table B 1. Pre Irradiation Tensile Properties of Shell Plate Material. Heat BCC-241 ll Test Strenathu l010 p11 _Elonaation. % Red'n of Specimen No. Temp. F . Yield Ultimate LigLleg lotal Area. % S5 601 73 75.6 91.9 12.7 27.0 67.3 603 73 69.4 90.0 13.1 27.2 67.0 6)4 73 71.9 90.3 13.0 28.8 71.1 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 55 606 580 64.4 86.3 14.4 25.7 65.4 h 611 580 64.4 86.3 13.6 26.0 63.7 615 578 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. P_re lrradiation Tensile Properties for Weld Metal WF-182-1 Test Strenoth. 1000 psi Elonaation. % Red'n of Epecimen No. Temp. F Yield Mitimate Uniform Total Area. % SS-003 73 69.7 85.6 14.8 26.0 63.7 007 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 S5 009 582 64.6 80.6 14.8 20.0 50.1 015 582 67.8 83.1 11.4 17.4 49.7 016 579 770.6 85.9 12.5 18.9 50.9 Hean 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 1 B2 13WtinnH%r l _

!I i il i i lI \ I LI l il 'I I

APPENDIX C Pre lrradiation Charpy impact Data lI I

I E I I C-1 I BW!!nn?afeiar I. I,w,---*,.-wr w ww-was--ae.- - e. m v. __q._ _ _ , , , .i- . ., ...y,,_- ,pm% ,,w e n m., , , , _ , , , , , .,,,.,9..pg, ,,.p_en-m,,, -,- -c.yyyw -w,q%,-.gs,+,.m 9m-y--wt yy -

u q I Table C-1. Pre lrradiation Charpy impact Data for Shell forging Material - Transverse Orientation. Heat BCC 241 Asborbed Lateral Shear E Specimen No. Test Temp., F Energy, ft lb Expagsion, 10 in. Fracture, 5 SS-616 79 5.5 10 0 55 636 40 17.5 14 0 55-609 - 2 19.5 18 0 55 617 0 16.5 16 0 g SS-621 21 39.0 33 2 S5 666 40 53.0 45 15 55-667 40 73.0 57 20 SS-672 40 88.0 69 60 SS 643 70 76.0 60 25 55-646 70 87.0 70 25 55 652 74 109.0 79 85 SS-627 106 99,0 74 80 SS-663 130 111.5 85 90 S5 686 171 120.0 88 100 SS 656 213 128.5 92 100 5S-658 278 116.0 89 100 SS-681 338 113.5 88 100 SS-630 585 113.0 83 100 I I C-2 I 13 W i!#vefe h r 5

  - . - . - . . - - - - - -     . . - . . . -                  . -          - . . - . -              . - - .--- - ~

I I Table C 2. Pre lrradiation Charpy lmpact Data for Shell forging Material Heat Affected Zone. Heat 14C-241 Asborbed lateral Shear lI Specimen No. Test lemp., F Energy, ft lb Exptig sion, 10 in. Fracture, l SS 331 120 27.0 19 0 S5 307 80 30.5 16 0 l SS-309 - 80 60.0 36 0 28.0 I SS-310 55 325 80

                                                  - 59              67.0 17 37          20 2

SS 346 - 40 56.0 31 10 55 320 - 20 62.0 37 25 55-337 20 94.0 54 30 SS-341 - 2 97.5 57 60 , SS-329 40 114.5 69 40 i SS-305 74 133.0 76 90 . SS-333 106 135.5 88 100 55 304 130 110.5 77 100 SS 315 176 138.5 82 100 SS 335 223 110.0 79 100 SS-343 338 112.0 83 100 55 322 406 135.5 84 100 'l 55 348 578 101.0 78 100 !I I I C3 I sW!!aMiftJar I

 .-... - - - - . - . - . - - -                      - . _ - - -          - - . = - - .                 - . - _ _ - - - _ - - . - . - .

I Teble C-3. Pre-Irradiation Charny impact Dala for Weld Metal Wi 182-1 Asborbed Lateral Shear Specimen Test Temp., Energy, Expagsion, fracture, 5 No. f ft-Ib 10 in. 7 3 SS 046 -80 15.5 16 0 55 060 -40 16.0 15 2 SS 077 SS 084

                                                -2
                                                -2 37,5 28.0 35 27 10 25 l

SS-053 0 33.0 29 20 55 055 0 33.5 29 15 SS-027 40 40.0 40 50 , 55-028 40 40.0 38 35 55-029 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 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 I E C-4 l I 13W!isefa % v l

I fioure C-1. Impact Data for Unirrtdiated Sht 11 Forairo Material. Heat KL211 100 i r : .: : , , , E ., 3 , p ,0 - g ?, . ., .

                                                         '        '            f             f         f 0   :       :     ba 1

0.10 , , ,* , , , i . i ge. _ . I w h0.06 , [0.04 b _ E f, 0.02 8 e i i

                                              '           i        i            i 0

220 , , ' ' ' '

                               - DAT A SumARY -

200 - T,n A b f__ icy (35 met) *?f r 180 -icy (50 et-a) +?$r i cy (30 rt ts) 41tr . g If0 c .ust (ayo) 1?? II.t Br

                    .*                          . sot RT""                                                                               .
                      , 140  -
                   =

gm -

                                                                            *       .                          e l                                                  .     .

5 . a 1* - i . . - g" g - l . 40 -

                                                                                                 $ASM,Ct.? (I)

P'at ta At 20

FLutNct Of -

ig Htat No. PCC-?'1 5 , , , im o 2. 2w = om sw . Telt-Temperature, F C-5 'I awamm-g

I figure C-2. Impact Data for Unirradiated Shell forging g Material. Heat Affe0ted 70ne. Heat BCC 241 3

i i

r , 75 - E $0

                         ,           #             e            ,            t           t 0.

s 0.10 ' ' *

e i ,

0.08 , i0.06 - a . E 1- .. g n ~ g p0.02 , t t i 0 220 , , ' i i #

               - DATA

SUMMARY

200 -T,,, 450f _ Igy 05au) 45f 180 ' icy (50 pt-u) -57f Icy Wnts)-W , a AN 'CyllSC (avol 1?4 ff-tBf O

   , 1%

RI

             - ut
                              *50f
   $120                                                               .                           -
 .p a 200 e*

5 80 - t . w -.

                                                                                  $8508,Cl.? (Mil)          W Patta:AL g   ,

Flutmet W - HtAt No. PCC-?At W p

            -100          o          100        . 200          500           h          h             g Itst fin eroture, F C-6 B W!!# # Wasar I

E I

I fi(lure { 3. }rDact Data for Unirradjated Weld Met,g], W[,.}) g J K' , ,

, O ,

                                                                                                ;          ,                   ,       O I

H 75 - - g M y s0 * . - I

25 -

p ; _ i e i i f i 0.10 i i , , , , 5 i 7 (D.06 * - : D . Qf, - - I :

        ' O.04       -

I a , S

0. 0.'

2 g 1 I i ! I i 110 , , , , , ,

                        - DAT A $U4%RY -

I 100 -i ?Of - Igy(35 Mit) !llf_ 93 "igy (M M.W ff _ I g y 8J I g, (30 FT LA)_*11I

                    'C,-ust (ave.) '/D fi lif j              ,U401 -
                                                                     ,              e          e I

E o E. f tD - I O [ 50 - - e C - - I 30 f I ?D 10

                        '        '                                                                P.Af tRI AL ASA/lladt 80 FLutNet       heat I             g
                -100 i

0 100 t f 200 f 3JJ Test Tecerature, F HtAt No. WI+Y ? i i 400 i 500 t.00 C-7 I SW!2Wahr ,I

I I I I !I i

.I                                              ,

lI APPENDIX D l Fluence Analysis Methodology

I I

lI I I I I I D-1 I SWi!##S h y I

I

1. Analytical Method A semiempirical method is used to calculate the capsule and vessel flux. The method employs explicit modeling of the reactor vessel and internals and uses an average core power distribution in the discrete ordinates transport code D011V, version 4.3. CSTIV calculates the energy and space dependent nedtron flux for the specific reactor under consideration. This semiempirical method is conven-l iently outlined in Figures 01 (capsule flux) and D 2 (vessel flux). g The two dimensional transport code DOTIV was used to calculate the energy and space-dependent neutron flux at all points of interest in the reactor system.

DOTIV uses the discrete ordinates method of solution of the Boltzmann transport equation and has multi-group and asymmetric scattering capability. The reference calculational model is an R-e geometric representation of a plan view through the reactor core midplane which includes the core, core liner, coolant, core barrel, g thermal shield, pressure vessel, and concrete. The material and geometry model, 5 represented in Figure D-3, uses one eight core symmetry. In order to include reasonable geometric detail within the computer memory limitations, the code parameters are specified as P 3 order of scattering, S quadrature, e and 40 energy groups. The P3 order of scattering adequately describes the predominately forward scattering of neutrons observed in the deep penetration of steel and l water media, as demonstrated by the close agreement between measured and calculated dosimeter activities. The S, symmetric quadrature has generally produced accurate results in discrete ordinates solutions for similar problems, g and is used routinely in the B&W R s 00T analyses. 5 Flux generation in the core was represented by a fixed distributed source which the code derives based on a 235 0 fission spectrum, the input relative power distribution, and a normalization factor to adjust flux level to the desired 5 power density. B 9eometrical Confiauration for modeling purposes, the actual geometrical configuration is divided into three parts, as shown in Figure D 3. The first part, Model *A," is used to generate D-2 I 13Wtisen%r Bl 4

=!

I l the energy dependent angular flux at the inner boundary of Model "B," which begins at the outer surface of the core barrel, Model A includes a detailed representation of the core baffle (or liner) in R d geometry that has been checked for both metal thickness and total metal volume to cr'ure that the 001 approximation to the actual geometry is as accurate as possible for these two I very important parameters. The second, Model B, contains an explicit represen-tation of the surveillance capt.ule and associated components. The B&W Owners Group's Flux Perturbation Experiment O verified that the surveillance capsule must be explicitly included in the DOT models used for capsule and vessel flux calculations in order to obtain the desired accuracy. The magnitude of the perturbations in the fast flux due to the presence of the capsule was determined in the Porturbation Experiment to be as high as 47% at the capsule center and as high as 10% at the inner surface of the reactor vessel. Detailed explicit modeling of the capsule, capsule holder tube, and internal components is I therefore incorporated into the DOT calculational models. The third, Model "C," is similar to Model B except that no capsule is included. Model C is used in determining the vessel flux in quadrants that do not contain a surveillance capsule; typically these quadrants contain the azimuthal flux peak on the inside l surface of the reactor vessel. An overlap region of approximately 33.0? cm or 17 radial intervals is specified l between Model A and Models B or C. The width of this overlap region, which is fixed by the placement of the Model A vacuum boundary and the Model B boundary source, was determined by an iterative process that resulted in close agreement g between the overlap region flux as predicted by Models A and B or C. The outer boundary was placed sufficiently far into the concrete shield (cavity wall) that I the use of a " vacuum" boundary condition does not cause a perturbation in the flux at the points of interest. Macrosconic Cross Sections Macroscopic cross sections, required for transport analyses, are obtained with the mixing code GIP. Nominal compositions are used for the structural metals. Coolant compositions were determined using the average boron concentration over I lI D3 I IBW!!MY8%%

Il a fuel cycle and the bulk temperature of the region. homogeneous mixture of fuel, fuel cladding, structure, and coolant. The core region is a l The cross section library presently used is the (22 neutron group and 18 gamma group) CASK 23E coupled set. The dosimeter reaction cross sections are based on the ENDF/B5 library, and are listed in Table E-3. The measured and calculated g dosimeters activities are compared in Table D-1. Distributed Source The neutron population in the core during full power operation is a function of neutron energy, space, and time. The time dependence is accounted for in the analysis by calculating the time weighted average neutron source, i.e. the neutron source corresponding to the time-weighted average power distribution. The effects of the other two independent variables, energy and space, are accounted for by using a finite but appropriately large number of discrete intervals in energy and space, in each of these intervals the neutron source is assumed to be invariant and independent of all other variables. The space and energy dependent source function can be considered as the product of a discretely expressed " spatial function" and a magnitude coefficient, i.e. Svgjg = [v/K Po ) x [RPDj )X,) (D 1) v- - v-- magnitude spatial where: Sv gjg

                                                                   =   Energy-and space-dependent neutron source, n/cc-sec,                                                                                                 I v/K                    =   fission neutron production rate, n/w sec, P,           =  Average power density in core, w/cc, RPD;)
                                                                   =   Relative power density at interval (i.j), unitiess,
                                                                   =  fission spectrum, fraction of fission neutrons having energy X,

in group "g," l i - Radial coordinate index, I D4 13W!!MWar I I

I j = Azimuthal coordinate index, g - Energy group index. The spatial dependence of the flux is directly related to the RPD distribution. Even though the entire (eighth core symmetric) RPD distribution is modeled in the g analysis, only the peripheral fuel assemblies contribute significantly to the ex-core flux. The axial average pin by pin RPD distribution is calculated on a quarter core symmetric basis for 8 to 12 times during each core cycle for the I entire capsule irradiation period. The time weighted average RPD distribution is used to generate the normalized space and energy dependency of the neutron source. Calculations for the energy and space dependent, time averaged flux were performed for the midpoint of each DOT interval throughout the model. Since the l reference model calculation produced fluxes in the R e plane that are averaged over the core height, an axial correction factor was required to adjust these l fluxes to the capsule elevation. The factor used (1.14) was prescribed in BAW-148SP.I4 l 1.1. Capsule flux and fluence Calculation As discussed above, the D0TIV code was used to explicitly model the capsule assembly and to calculate the neutron flux as a function of energy within the capsule. The calculated fluxes were used in the following equation to obtain g calculated activities for comparison with the measured data. The calculated activity for reaction product Di , in (pCi/gm) is:

                                                                            'j) 3on (E) 4 (E) 1 f) (1 e' Al t;) , - A i(T D=              '                                                     (D2) i (3.7 x 10') An E                              j where:

N = Avogadro's number, An Atomic weight of target material n, l f, = Either weight fraction of target isotope in n-th material or the fission yield of the desired isotope, lI D5 i E BWitsam!L, l l

I un(E) Group averaged cross sections for material n (listed in lable E 3) l ((f.) - Group averaged fluxes calculated by DOTIV analysis, F; Fraction of full power during j th time interval, t 3

                          *i - Decay constant of the ith isotope.

T - Sum of total irradiation time, i.e., residual time in reactor, e and the wait time between reactor shutdown and counting times, g ej - Cumulative time from reactor startup to end of j th time period. t) - Length of the j th time period I Adjustments were made to the calculated dosimeter activities to correct for the effects listed below: Short half-life adjustments to Ni and Fe dosimeter activities 238 237 Np dosimeter activities Photorission adjustments to 0 and 238 Fissile impurity adjustments to U dosimeter activities Af ter making these adjustments the calculated dosimeter activities were used with the corresponding measured activities to obtain the flux normalization factors: C,= 0 (measured) 1 , g 0,(calculated) B These normalization factors were evaluated, averaged, and then used to adjust the g calculated test specimen flux and fluence to be consistent with the dosimeter 5 measurements. Additionally, the normalization factor was used to update the average normalization factor which had been derived from previour, analyses. The updated normalization factor was then used to adjust the calculated vessel flux and fluence. The flux normalization factors are given in Table D 1. Vessel Fluence Extrapolation l 2. for past core cycles, fluence values in the pressure vessel are calculated as described above. Extrapolation to future cycles is required to predict the I D6 I IBW!!nMW 5 ....i......... .. ,,,

I l useful vessel life. Three time periods are considered in the extrapolation: 1) operation to date for which vessel fluence has been calculated, 2)futurefuel cycles for which PDQ calculations have been performed, and 3) future cycles for which no analyses exist, for the Davis Besse Unit I analysis, time period 1 is through cycle 6, time period 2 covers cycle 7, and time period 3 covers from the end of cycle 7 through 32 EFPY. The flux and fluence for time period 2 was estimated by calculating the I vessel flux using an adjoint D01 calculational procedure with the appropriate assembly average power distributions and integrating these values over time period 2. The extrapolation of the fluence through time period 3 was accom-plished by assuming that the average flux during period 3 was equal to the l average flux for period 2 (cycle 7). Table 01. Normalization factor Measured Calculate flux Activity,I8) Activity, b) Normalization uti/a nCi/o fag. tor 0 fe(n.p)54Mn 853.73 1152.79 0.89(c) 58 Hi(n.p)58Co 0.93(d) 1464.39 2001.51 l 1.14 11.86 10.36 238V (n,f)l37Cs 70.55 61.69 1.15 237Np(n,f)l37Cs Averaged: 1.035 ') I (a) Average of four dosimeter wires. (b) Average at four calculated activities. (c) Average at four ratios (one for each dosimeter wire) corrected by short half-life f actor of 1.195. (c) Average of four ratios (one for each dosimeter wire) corrected by short half life factor of 1.264. (d) Average of all four dosimeters was selected as the normalization constant. I I 0-7 E 13WtisefWem

I

I Table D 2. Davis Besse Unit 1 Reactor Vessel Fluence by Cycle Cycles incremental Time, EFPY Cumulative Time, EFPY Vessel n/cm [s 1ux, Vessel Fluence, n/cm 2(C) incremental Cumulative l 1 1.02 1.02 1.61E+10 5.19E+17 5.19E+17 26 4.43 5.45 1.38E+10 1.92E+18 2.44E+18 7 1.07 6.52 1.03E+10 3.47E+17 2.79E+18 l 1,48 8.00 1. 03 E + 10(*) 4.80E+17(b) 3.27E+18(b) 5.55E418(b) l 7.00 15.00 1.03E+10(') 2. 28E+ 18(b) 1 6.00 21.00 1. 03 E + 10(') 1. 95 E + 18(') 7.50E+18(b) 11.00 32.00 1.03E+10(*) 3.60E418(b) 1.llE+19(6) (a) Maximum neutron flux at inside surface of reactor vessel, based on fuel cycle designs for future cycle 7, used for extrapolation of fluence to future times. (b) Extrapolated values. (c) Peak fluence at inside surface of reactor vessel. I I I I I l I I D8 I B W !! M e!A hur y

='

I I figure 0 1. Rationale for the Calculation of Dosimeter Activities and Neutron Flux in the Caostt].g DOF/B4 010$$ SECTICNS GEDETRY & G.MDRATlRE PGER DISTRI-DCF/B5 DQSDETER REAC- FIR H] DEL A DUT BlftIDNS SDCE I IIR 01055 SECTIDNS CAPSULE D6ER-TICN (PDCD W ,r I omSSm .

So m

                                                                      TDE AVE DISTRI-I           GEIMETRY &

OLMDRATlRE K) DEL B DUT 4 K) DEL A 1 BUTED SOURCE Sy (E, R,e ) ir ir ir DOT 4 ANGLAR FUDC KlDQ. B AT BARREL PDWER HIST (RY 00SIMETER DF CAPSULE

                                >    ACTIVITIES                       (PRHIST CIDE) g I                                        ,r FIM.

cAtatATED 4 I ACTIVITIES a MEASLRED 'I DaSvertR ACTIVITIES I 0]RRELTION FACitR n <r 5 CAPSULE FU R KWALIZATION FACTDR

                             >                            4              M/C RATIO

.I l D-9 I BW!!bMi%r l I

I Figure D.2. Rationale for the Calculation of 9 5 Neutron Flux in the Rearlor Vessel GEDETRY & Ol%DRARRE PGER DISTRI. g FtR KDE1. "A" DOT

 -- KIO10 SCOPIC CROSS                                               BlfrIONS SDCE STARRP (PD0   m SECTIONS DOF/B4 D0F/85                                                     OR E31IVALIND g v                                         v           v H OTOSCOPIC CROSS                                     SORRD. CIDE SECTIONS BY E-Gl0lP                                      *
           " GIP" (IDE                 - - -

TDE AVE DISTRIBlfiTD SOURCE Sv (E, R,e ) g u v 4 m GEDETRY #0 OLMDRATLRE E 5 FOR 00T KDEl. B OR C u v DOT 4 PODCL 8 #GLAR F1.lK AT

 -+          #0/0R P000. C      4-          EARRfl. SLRFACE I

I I NORMM.12ATIIN FACTOR FR04 AXIAL CORRECTION CAPSlLE FUENCE #MLYSIS FACTOR l (FROM T}E DIAGMM ON 1)E

   - ram                                                                           I if     if        9P l

TDE-AVDtAGE VESSO. FUK AT

                             >%XIMN VESSEL. LOCATION (E, R,e) 0-10 13W!!s&f/4Lw I

I,

m m m m m m m m m m m m M -h m m m m m Figure D-3. Plan View Through Reactor Core Midplane - i (Reference R-9 Calculation Model) Y

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1 l I I I I I I I APPENDIX E Capsule Dosimetry Data I I I I I I I E-1 I - BWitnM%% I

Il Table E 1 lists the characteristics of the neutron dosimeters. Table E-2 shows the measured activity per gram of target material (i.e., per gram of uranium, nickel, etc.) for the capsule dosimeters. Activation cross sections for the ! 235 various materials were flux-weighted with the 0 fission spectrum shown in Table E-3. g Table E 1. Detector Composition and Shieldina Detector Material % Tarnet Shieldina Reaction 230 238U (n,f)I370s U Al 10.38% 0 Cd Ag Np-Al 1.44% 237 Np Cd Ag 237Np(n,f)l37c3 l 58 58 Ni 67.77% Ni Cd-Ag Ni(n,p)58Co 59 59Co(n,y)60Co Co Al 0.66% 0o Cd Co-Al 0.66%5900 None 59Co(n,7)60Co 54 54Fe(n.p)S4Mn Fe 5.82% Fe None I Table E-2. Measured bpecific Activities (Unadjusted) for Dosimeters in Caosule TEl-D Dosimeter Activity. (nCi/am of Taraet) Detector Material Dosimeter Reaction DD1 DD2 003 004 Ni 58Ni(n,p)58Co 1507.23 1433.60 1071.19 1845.52 Fe 54Fe(n,p)54Mn 878.81 852.94 627.96 1055.20 U-Al 2380 (n,f)l37Cs 13.80 12.41 9.26 16.97 237 Np(n.f)l37Cs Np-Al 73.37 64.10 57.95 92.54 I I E-2 I I IBWitnEVais%r E

I Table E 3. Dosimeter Activation Cross Sections, b/ atom (a) 237 58 54 G Energy Range, MeV Np(n,f) 238V (n,f) Ni(n,p) Fe(n.p) 1 12.2 - 15 2.323 1.051E+0 4.830E-1 4.133E-1 2 10.0 - 12.2 2.341 9.851E-1 5.735E-1 4.728E-1 3 8.18 - 10.0 2.309 9.935E 1 5.981E-1 4./72E-1 4 6.36 - 8.18 2.093 9.110E-1 5.921E-1 4.714E-1 5 4.96 - 6.36 1.542 5.777E-1 5.223E-1 4.321E 1 6 4.06 - 4.96 1.532 5.454E-1 4,146E 1 3.275E 1 7 3,01 - 4.06 1.614 5,340E-1 2.701E-1 2.193E 1 8 2,46 - 3.01 1.689 5.325E-1 1.445E-1 1.080E-1 2.46 1.695 5.399E-1 9.154E 2 5,613E-2 9 2.35 - 10 1.83 - 2.35 1.676 5.323E-1 4.856E-2 2.940E-2 11 1.11 - 1.83 1.596 2.608E-1 1.180E-2 2.948E 3 12 0.55 - 1.11 1.241 9,845E-3 1.336E-3 6.999E-5 l 13 14 0.111 - 0.0033 - 0.111 0.55 2.352E-1 1.200E-2 2.436E 4 6.818E-5 5.013E-4 1.512E-5 6.419E-8 0 I (a)ENDF/B5 valgg that have been flux weighted (over CASK energy groups) based on a U fission spectrum in the fast energy range plus a 1/E shape I in the intermediate energy range. I I I I E-3 I 13W!!seniMam I

, I: 'I I I

.I I

I I APPENDIX F References I I I I I I I F-1 I BW!!nEW%%r l I

Il

1. A. L. Lowe, Jr., et al., Analysis of Capsule TEl-f from Toledo Edison Company, Davis-Besse Nuclear Power Station, Unit 1, Reactor Vessel Materi-als Surveillance Program, BAW-1701, Babcock & Wilcox, Lynchburg, Virginia, January 1982.

2. A. L. Lowe, Jr., et al., Analysis of Capsule TEl-B from Toledo Edison Company, Davis-Besse Nuclear Power Station, Unit 1, Reactor Vessel Materi-als Surveillance Program, BAW-1834, Babcock & Wilcox, Lynchburg, Virginia, May 1984.

3. A. L. Lowe, Jr., et al., Analysis of Capsule TEl-A from Toledo Edison Company, Davis-Besse Nuclear Power Station, Unit 1, Reactor Vessel Materi-l als Surveillance Program, BAW 1882, Babcock & Wilcox, Lynchburg, Virginia, September 1985.

4. H. S. Palme, G. S. Carter, and C. L. Whitmarsh, Reactor Vessel Material Surveillance Program -- Compliance With 10CFR50 Appendix H, for Oconee-Class Reactors, BAW-10100A, Babcock & Wilcox, Lynchburg, Virginia, February 1975.

5. Code of Federal Regulation, Title 10, Part 50, Fracture Toughness Require-ments for Light-Water Nuclear Power Reactors, Appendix H, Reactor Vessel M?terial Surveillance Program Requirements.

6. American Society for Testing and Materials, Standard Recommended Practice I for Surveillance Tests for Nuclear Reactor Vessels, E185-73, March 1,1973.

7. S. Fyfitch, L. B. Gross, and A. L. Lowe, Jr., Master Integrated Reactor Vessel Surveillance Program, BAW-1543. Rev. 3, Babcock & Wilcox, Lynchburg, g Virginia, September 1989. E

8. American Society for Testing and Materials, Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, E185 82, July 1,1982.

I I

F-2 I

l I 13W!!=fs h y 3

I

9. Code of Federal Regulation, Title 10, Part 50, Fracture Toughness Require-I ments for Light-Water Nuclear Power Reactors, Appendix G, Fracture Tough-ness Requirements.

10. K. E. Moore and A. S. Heller, Chemistry of 177-FA B&W Owners' Group Reactor Vessel Beltline Welds, BAW 1500P, Babcock & Wilcox, Lynchburg, Virginia, September 1978.

11. American Society for Testing and Materials, Methods and Definitions for I Mechanical Testing of Steel Products, A370-77, June 24, 1977.

12. American Society for Testing and Materials, Methods for Notched Bar impact I Testing of Metallic Materials, E23-82, March 5,1982.

13. A. L. Lowe, Jr., gL.J.1, Evaluation of Surveillance Capsule Temperatures, I BAW 2040, Babcock & Wilcox, Lynchburg, Virginia, March 1989.

A. L. Lowe, Jr., g.L3.1, Fracture Toughness Test Results from Capsule TEl-D, I 14. The Toledo Edison Company Davis Besse Nuclear Power Station Unit 1 Reactor Vessel Material Surveillance Program, BAW-2128 B&W Nuclear Service Company, Lynchburg, Virginia, To be Published.

15. S. Q. King, Pressure Vessel Fluence Analysis for 177-FA Reactors, BAW-1485P. Revision 1, Babcock & Wilcox, Lynchburg, Va., April 1988.

16. B&W's Version of 00TIV Version 4.3, One- and Two-Dimensional Transport Code I System," Oak Ridge National Laboratory, Distributed by the Radiation Shielding Information Center as CC-429, November 1, 1983.

17. " CASK-40-Group Coupled Neutron and Gamma-Ray Cross Section Data," Radiation Shielding Information Center, DLC-23E.

18. Dosimeter File ENDF/B5 Tape 531, distributed March 1984, National Neutron Data Center, Brookhaven National Laboratory, Upton, Long Island, NY.

I 19. American Society of Testing Materials, Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (DPA), E693-79 (Re-I approved 1985). I F-3 I BWunniu%m I

I

20. U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel Material, Reaulatory Guide 1.99. Revision 2, May 1988.

21. A. S. Heller and A. L. Lowe, Jr., Correlations for Predicting the Effe:13 of Neutron Radiation on Linde 80 Submerged-Arc Welds, DAW 1803, Babcock &

Wilcox, Lynchburg, Virginia, January 1984.

22. The Third Surveillance Test on The Ko-Ri Unit No. 1 Reactor Vessel Materials (Capsule S), June 1986.

23. J. D. Aadland, Babcock & Wilcox Owner's Group 177-Fuel Assembly Reactor Vessel and Surveillance Program Materials Information, BAW 1820, Babcock &

Wilcox, Lynchburg, Virginia, December 1984.

24. K. E. Moore and A. S. Heller, B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study, BAW-1799, Babcock & Wilcox, Lynchburg, Virginia, July 1983.

25. N. L. Snider and L. A. Hassler, B&WOG Flux Perturbation Experiment at ORNL, Measured and Calculated Dosimeter Results, BAW 1886, Babcock & Wilcox, Lynchburg, Virginia, September 1985.

I I I I I I I I F-4 I naam- g}}

ML20072U421

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