Text

Analysis of Capsule P from Northern States Power Co Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program
	ML20069H222
Person / Time
Site:	Prairie Island
Issue date:	05/31/1982
From:	Kaiser W, Tran K, Yanichko S; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20069H208	List: ML20069H205; ML20069H222;
References
	TAC-49027, TAC-59974, TAC-59975, WCAP-10102, NUDOCS 8210190429
	Download: ML20069H222 (90)
	v • d • e

'

e WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION E5

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ANALYSIS OF CAPSULE P FROM NORTHERN STATES POWER COMPANY PRAIRIE ISLAND UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko K. C. Tran W. T. Kaiser MAY 1982 Work performed under Shop Order No. ELEP 949 g

APPROVED:

I T. R. Mager, Manager Metallurgical and NDE Analysis Prepared by Westinghouse for Northern States Power Company Although information contained in this report is nonproprictary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230

' 8210190429 B21012 i

PDR ADOCK 05000202 P

PM

5 i

I TABLE OF CONTENTS 1

1 Section Title Page 1

SUMMARY

OF RESULTS 1-1 2

INTRODUCTION i

2-1 3

BACKGROUND 3-1 4

DESCRIPTION OF PROGRAM 4-1 5

TESTING OF SPECIMENS FROM CAPSULE P 5-1 5-1.

Overview 5-1 Charpy d-Notch Impact Test Results 5-2.

5-2 l

5-3.

Tensile Test Results 5-2 5-4.

Wedge Opening Loading Tests 5-2 6

RADIATION ANALYSIS AND NEUTRON DOSIMETRY.6-1 6-1.

Introduction 6-1 6-2.

Discrete Ordinates Analysis 6-1 6-3.

Neutron Dosimetry 6-7 6-4.

Transport Analysis Results 6-12 6-5.

Dosimetry Results 6-13 7

REFERENCES 7-1 APPENDlX A HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERAT!ON A-1 i

r t_

]

l 2

LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the Prairie island Unit No.1 Reactor Vessel 4-5 4-2 Capsule P Schematic Showing Location of Specimens, Thermal Monitors. and Dosimeters 4-7/4-8 5-1 Irradiated Charpy V-Notch Properties for Prairie Island Unit 1 Reactor Vessel Shell Forging C (Axial Orientation) 5-9 5-2 Irradiated Charpy V-Notch Properties for Prairie Island Unit 1 Reactor Vessel Shell Forging C (Tangential Orientation) 5-10 5-3 Irradiated Charpy V-Notch Properties for Prairie Island Unit 1 Reactor Pressure Vessel Weld Metal 5-11 5-4 Irradiated Charpy V-Notch Properties for Prairie Island Unit 1 Reactor Pressure Vessel Weld Heat-Affected-Zone Metal 5-12 5-5 Irradiated Charpy V-Notch Properties for A533 Grade B Class 1 Correlation Monitor Material 5-13 5-6 Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 1 Shell Forging C (Axial Orientation) 5-14 5-7 Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 1 Shell Forging C (Tangential Orientation) 5-15 5-8 Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 1 Weld Metal 5-16 5-9 Charpy impact Spacimen Fracture Surfaces for Prairie Island Unit 1 Weld Heat-Affected-Zone Metal 5-17 5-10 Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 1 A533 Grade B Class 1 Correlation Monitor Material 5-18 ii

i i

LIST OF ILLUSTRATIONS (cont)

Figure Title Page 5-11 Irradiated Tensile Properties for Prairie Island Unit 1 Reactor Pressure Vessel Shell Forging C (Axial Orientation) 5-19 5-12 Irradiated Tensile Properties for Prairie Island Unit 1 Reactor Pressure Vessel Shell Forging C (Tangential Orientation) 5-20 5-13 Irradiated Tensile Properties for Prairie Island Unit 1 Reactor Pressure Vessel Weld Metal 5-21 5-14 Typical Stress-Strain Curve for Tension Specimens 5-22 5-15 Fractured Tensile Specimens from Prairie Island Unit 1 Pressure Vessel Shell Forging C (Axial Orientation) 5-23 5-16 Fractured Tensile Specimens from Prairie Island Unit 1 Pressure Vessel Shell Forging C (Tangential Orientation) 5-24 5-17 Fractured Tensile Specimens from Prairie Island Unit 1 Pressure Vessel Weld Metal 5-25 6-1 Prairie Island Unit 1 Reactor Geometry 6-2 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-6 6-3 Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel Surveillance Capsule Geometry 6-26 6-4 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-27 6-5 Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-28 6-6 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within Surveillance Capsules P and V 6-29 6-7 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsule V 6-30 6-8 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsule P 6-31 6-9 Comparison of Measured and Calculated Fast Neutron Fluence (E > 1.0 Mev) for Capsules P and V 6-32 111

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LIST OF TABLES I

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Page 4-1 Chemistry and Heat Tr'estment o, Material Representing the Core Region Shell Forging and Weld Metal from the Prairie Island Unit 1 Reactor Vessel 4-3 l

4-2 Chemistry and Heat Treatment of Surveillance Material Representing 12-inch-thick A533 Grade B Class 1 Correlation Monitor Material 4-4 1

4-3 Welding Procedure and Associated Information for I

the Prairie Island Unit 1 Core Region Weldments 4-4 5-1 Charpy V-Notch Impact Data for the Prairie Island Unit 1 Pressure Vessel Shell Forging C Irradiated at 550* F, Fluence 1.25 x 10n/cm (E > 1 Mev) 5-3 a

5-2 Charpy V-Notch Impact Data for the Prairie Island Unit 1 Pressure Vessel Weld and Heat-Affected Zone Metal Irradiated at 550* F, Fluence 1.25 x 10n/cm2 (E > 1 Mov) 5-4 l

5-3

' Charpy V-Notch Impact Data for the Prairie Island

.j Unit 1 A533 Grade B Class 1 Correlation Monitor i

Material irradiated at 550*F Fluence 1.25 x

}

10'*n/cm2 (E > 1 Mev) 5-5 5-4 The Effect of 550* F Irradiation at 1.25 x 10n/cm2 l

(E > 1 Mov) on the Notch Toughness Properties of the Prairie Island Unit 1 Reactor Vessel Impact Test Specimens 5-6 5-5 Summary of Prairie Island Unit 1 Reactor Vessel Surveillance Capsule Charpy Impact Test Results 5-7 5-3 Irradiated Tensile Properties for the Prairie Island Unit 1 Pressure Vessel Shell Forging C and Weld Metal, Fluence 1.25 x 10n/cm2 (E > 1 Mev) 5-8 6-1 21 Group Energy Structure

'6-5 6-2 Nuclear Parameters for Neutron Flux Monitors 6-8 6-3 Calculated Fast Neutron Flux (E > 1.0 Mev) and Lead Factors for Prairie Island Unit 1 Surveillance Capsules 6-15 iv

LIST OF TABLES (cont)

Table Title Page 6-4 Calculated Neutron Energy Spectra at the Dosimeter Block Location for Prairie Island Unit 1 Surveillance Capsules 6-16 6-5 Spectrum Averaged Reaction Cross Sections at the Dosimeter Block Location for Prairie Island Unit 1 Surveillance Capsules 6-17 6-6 Irradiation History of Prairie Island Unit 1 Reactor ~

Vessel Surveillance Capsule P

. 6-18 6-7 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule P 6-21 6-8 Comparison of Measum,! and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule V 6-22 j

6-9 Results of Fast Neutron Dosimetry for Capsules P and V 6-23 i

6-10 Results of Thermal Neutron Dosimetry for Capsules l

P and V 6-24

i 6-11 Summary of Neutron Dosimetry Results for Capsules

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P and V 6-25 i

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SECTION 1

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in Capsule P, the second surveillance capsule removed from the Prairie Island Unit 1 reactor pressure vessel, led to the following conclusion:

a The capsule received an average fast-neutron fluence of 1.25 x 10' n/cm2 (E > 1.0 Mev) compared to a calculated value of 1.21 x 10 n/cm2 E Based on the fluence measurements for Capsule P, the vessel inner surface fluence after 4.6 effective-full-power years of operation is 6.44 x 10'8 n/cm2 com-pared to a calculated fluence of 6.23 x 10" n/cm2 e The fast-neutron fluence of 1.25 x 10" n/cm2 resulted in the following int reases in transition temperature and decreases in upper shelf energy for the various reactor vessel surveillance materials: (see table 5-4).

l 50 Ft Ib 30 Ft Ib Temperature Temperature Shelf Energy increase increase Decrease Material

(*F)

(* F)

(ft lb)

Forging C 25 20 7

(Tangential)

Forging C 51 37 16 i

(Axial)

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Weld Metal 60 42

+4.5

{

HAZ Metal 65 70 68 1-1

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e The results of the material surveillance tests indicate that th'e reactor pressure vessels beltline material is not very sensitive to radiation.

a The projected end-of-life fluences at various locations through the vessel wall are as follows:

Vessel Fast Neutron Fluence (n/cm )

a Wall Location Measured Calculated Inner Surface 4.17 x 10

4.30 x 10

1/4 Thickness 2.78 x 10

2.87 x 10

3/4 Thickness 8.19 x 10

8.46 x 10'8 G

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SECTION 2 INTRODUCTION i

This report presents the results of the examination of Capsule P, the second capsule of the continuing surveillance program which monitors the effects of neutron irradia-tion on the Northern States Power Company, Prairie Island Unit 1 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Prairie Island Unit 1 reactor pressure vessel mate-rials was des.'gned and recommended by the Westinghouse Electric Corporation. A descripton of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented in WCAP-8086.I'l The surveillance program was planned to cover the 40-year life of the reactor pressure vessel and is based on ASTM E-185-73," Recommended Practice for Survelocnce Tests for Nuclear Reactor Vessels."r21 Westingh' use Nuclear Energy Systems personnel were con-o tracted for the preparation of procedures for removing the first capsule from the reactor and its ahipment to the Westinghouse Research and Development Labora-tory,where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes testing and the postirradiation data obtained from the second material surveillance capsule (Capsule P) removed from the Prairie Island Unit 1 reactor vessel and discusses the analysis of these data. The data are also compared to results of the previously removed Prairie Island Unit 1 surveillance Capsule V reported by Davidson.I83 Using current methods I'l heatup and cooldown pressure-temperature operating limits are established for the nuclear power plant.

The heatup and cooldown pressure-temperature operating limits are presented in Appendix A.

2-1

i 11.~

i SECTION 3 i

BACKGROUND i

The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety l

in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bom-bardment.The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels such as SA508 Class 3 (base material of Unit j

1 reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a j

decrease in ductility and toughness under certain conditions of irradiation.

A method for performing analyses to guard against fast fracture in reactor pressure

{

vessels has been presented in " Protection Against Non-ductile Failure," appendix G to section ill of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature, l

RT NDT-NDT s defined as the greater of either the drop weight nil-ductility transition i

RT temperature (NDTT per ASTM E-208) or the temperature 60' F less than the 50 ft Ib (and 35 mils lateral expansion) temperature as determined from Charpy specimens j

oriented normal to the major working direction of the material. The RTNDTofagiven material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in appendix G of the ASME Code. The KIR curve is a lower j

bound of dynamic, crack arrest, and static fracture toughness results obtained from i

several heats of pressure vessel steel. When a given material is indexed to the KIR cu rve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

3-1

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RTNDT, ar;d in turn the operating limits of nuclear powar plants, can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure yassel steel can be monitored by a reactor surveillance program such as the Northern States Power Company, Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program,Ul n which a surveillance capsule is periodically removed f rom l

the operating nuclear reactor and the encapsulated specimens are tested. The increase in the Charpy V-notch 50 ft Ib temperature ( ARTNDT) due to irradiation is NOT or radiation embrittinment.This f

NDT o adjust the RT added to the original RT t

NOT nitial+ARTNOT) is used to index the materital to the KIR adjusted RTNOT (RT i

curve and in turn to set operating limits for the nuclear power plant wtilch take into account tiie effects of irradiation on the reactor vessel materials.

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F SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Prairie Island Unit 1 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.

Capsule P was removed after approximately 4.6 EFPY of plant operation. This capsule contained Charpy V-notch impac*, tensile, and wedge opening loading (WOL) fracture mechanics specimens from the intermediate shell ring forging (heat 21918/38566), weld metal from the core region of the reactor vessel, and Charpy V-notch specimens from weld heat-affected zone (HAZ) material. The capsule also contained Charpy V-notch specimens from the 12-inch-thick correlation monitor material (A533 Grade B Class 1) furnished by Oak Ridge National Laboratory. The chemistry and heat treatment of the surveillance material is presented in Tables 4-1 and 4-2.

All test specimens were machined from the %-thickness location of the forging. Test specimens represent material taken at least one forging thickness from the quenchad end of the forging. All base metal Charpy V-notch and tensile specimens were oriented with the longitudinalaxis of the specimen both normal to and parallel to the principal working (hoop) direction of the forging. The WOL test specimens were machined such that the simulated crack of the specimen would propagate normal to (tangential specimens) and parallel to (axial specimens) the hoop direction of the forging. All specimens were fatigue precracked per ASTM E399-70T.

Charpy V-notch specimens from the weld metal chamfer region were oriented with the longitudinal axis of the specimens transverse to the weld direction. Tensile l

specimens were oriented with the longitudinal axis of the specimen parallel to the 4-1

a

=

T weld. Table 4-3 lists the weld procedure and information associated with the Prairie

-[

Island Unit 1 cora region weldments. Capsule contained dosimeter wires of oure copper, iron, nickel, and aluminum-0.15 wt%-cobalt (cadmium-shielded and un-

}

shielded). In addition, cadmium-shielded dosimeters of Np287 and U28 were contained 9

in the capsule and located as shown in Figure 4-2.

Thermal monitors made from two low melting eutectic alloys and sealed in Pyrex tubes were included in the capsule and were located as shown in Figure 4-2. The two eutectic alloys and their melting points are:

[

2.5% Ag,97.5% Pb Melting Point 579'F 1

1.75% Ag,0.75% Sn,97.5% Pb Melting Point 590*F

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TABLE 4-1 CHEMISTRY AND HEAT TREATMENT OF MATERIAL REPRESENTING THE CORE REGION SHELL FORGING AND WELD METAL FROM THE PRAIRIE ISLAND UNIT 1 REACTOR VESSEL Chemical Analyses (Percent)

Intermediate Element Shell C Weld Metal 21916/38566 C

0.17 0.052 Mn 1.41 1.30 P

'O.013 0.017 S

0.005 0.014 Si 0.28 0.36 Ni 0.72 Cr 0.17 0.015 V

<0.002 0.001 Mo 0.48 0.51 C.o 0.010 0.001 Cu 0.06 0.13 Sn 0.007 0.007 Zn 0.001 0.001 'l l

Al 0.033 0.015 N

0.006 0.014 Ti 0.001(*I 0.001 Sb 0.001 '1 0.001 l

As 0.001 'l 0.061 1

B 0.003 '1 0.003('l l

Zr 0.001 'l 0.001 t

Heat Treatment Forging C Heated at 1652/1715'F for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , water-quenched; Ht. No.

tempered at 1175/1238'F for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , furnace-cooled; 21918/38566 heated at 1652/1724* F for 5% hours, water-quenched; tempered at 1202/1238*F for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , furnace-cooled; stress-relieved at 1022* F for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , furnace-cooled; stress-relieved at 1112* F for 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months , furnace-cooled Weldment Stress-relieved at 1022*F for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , furnace-cooled; stress-relieved at 1112* F for 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , furnace-cooled

a. Not Detecti d. The number indicates the minimum limit of detection.

4-3

6

~ - * -

TABLE 4-2 CHEMISTRY AND HEAT YREATMENT OF SURVEILLANCE MATERIAL REPRESENTING 12-INCH-THICK A533 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL Chemical Analysis C

Mn P

S Si Ni Mo Cu Ladle 0.22 1.45 0.011 0.019 0.22 0.S2 0.53 Check 0.22 1.48 0.012 0.018 0.25 0.68 0.52 0.14 Heat Thatment 1675 25'F - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months - Air-cooled 1600 25'F - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months - Water-quenched 1125 25'F - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months - Furnace-cooled 1150 25'F - 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months - Fumace-cooled to 600* F TABLE 4-3 WELDING PROCEDURE AND ASSOCIATED INFORMATION FOR THE PRAIRIE ISLAND UNIT 1 CORE REGION WELDMENTS(*I

^

Top r.nd Bottom of Chamfer - Automatic submerged arc welding with multiple passes. Preheat of 356*F. Four passes on each side of chamfer.

Wire: UM 40 - 2.5 mm dia. - lot: 3049 Flux: UM 89 lot: 1230 Welding Speed: 36 cm/ min I

Chamfer Filling -

Automatic submerged arc welding with multiple l

passes. Preheat of 356'F.

Wire: UM 40 - 4 mm dia. -lot: 1752 Flux: UM 89 tot: 1230 l

Welding Speed: 40 cm/ min

a. Wold Groove - Double U Configuration

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Arrangement of Surveillance Capsules in the Prairie Island Unit 1 Reactor Vessel (Updated Lead Factors for the Capsules are Shown in Parentheses) 4 4-5

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e SECTION 5 TESTING OF SPECIMENS FROM CAPSULE 5-1.

OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Hot Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel.

Upon receipt of the capsule at the Laboratory, the specimens and spacer blocks were carefully removed, inspected for identifi.:ation number, and checked against the master list in WCAP-8086.m No discrepancies were found.

A TMI Model TM 52004H impact test machine was used to perform tests on the irradiated Charpy V-notch specimens. Before initiating tests on the irradiated Charpy V specimens, the accuracy of the impact machine was checked with a set of standard specimens obtained from the Army Material and Mechanics Research Center in Watertown, Massachusetts. The results of the calibration testing showed that the machine was certified for Charpy V-notch impact testing.

The tensile tests were conducted on a screw-driven instron testing machine having a 20,000 pound capacity. A crosshead speed of 0.05 in/ min was used. The deformation of the specimen was measured using a strain gage extensomett.r. The extensometer was calibrated before testing with a Sheffield high-magnification drum-type exten-someter calibrator.

Elevated temperature tensile tests were conducted using a split-tube furnace. The specimens were held at temperature a minimum of 30 minutes to stabilize their temperature prior to testing. Temperature was monitored using a chromel-alumel thermocouple in contact with the upper and lower clevis-pin specimen grips. Temper-ature was controlled within plus or minus 5'F.

5-1

o The load-extension data were recorded on the testing machine strip chart. The yield strength, ultimate t. ensile strength, and uniform elongation were determined from these charts. The reduction in area and total elongation were determined from specimen measurements.

5-2.

CHARPY V-NOTCH IMPACT TEST RESULTS Charpy V-notch impact test results for the reactor vessel beitfir:e forging C material, we!d metal and haat-affected zone (HAZ) material and ASTM correlation monitor material from HSST Plate 02 irradiated to a fluence of 1.25 x 10 n/cm2 are presented in Tables 5-1 through 5-3 and Figures 5-1 through 5-5, respectively. A summary of the increases in transition temperature and decrease in the upper shelf energy of the various surveillance materials is presented in Table 5-4 and shows that 30 and 50 ft Ib transition temperature increases for the shell forging C and the weld metal and HAZ material are small (20 to 70* F) for a fluence of 1.25 x 10 n/cm2, therefore indicating that the materials are not very sensitive to radiation. A comparison of the transition temperature increases resulting from irradiation tests performed on the two Prairie Island Unit 1 capsules tested to date shown in Table 5-5 indicates the vessel materials are not very sensitive to radiation. Photographs of broken Charpy impact specimens from the surveillance forging, weld metal and heat-affected-zone are shown in Figures 5-6 through 5-10.

5-3.

TENSILE TEST RESULTS The results of tensile tests performed on specimens from shell forging C and the weld metal are shown in Table 5-6. A comparison of the unirradiated versus irradiated tensile properties is shown in Figures 5-11 through 5-13 for forging C and the weld metal. The smallincreases in yield strength of approximately 5 to 10 ksi resulting from irradiation to 1.25 x 10 n/cm2 tend to confirm that the reactor vessel beltline materials are not highly sensitive to radiation as also indicated by the Charpy V-notch tests. A typical load-displacement curve obtained for the tensile tests is shown in Figure 5-14. Photographs of broken tensile specimens from the surveillance forging and weld metal are shown in Figures 5-15 through 5-17.

5-4.

-WEDGE OPENING LOADING TESTS The Wedge Opening Loading fracture mechanics specimens that were contained in Capsule P have been stored at the Westinghouse Research Laboratory on the recommendation of the United States Nuclear Regulatory Commission and will be tested at a late,rdate. The results of these tests will be reported upon their completion.

5-2

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c TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE.cRAIRIE ISLAND UNIT 1 PRESSURE VESSEL SHELL FORGING C lRRADIATED AT 550* F, FLUENCE 1.25 x 10 n/cm2 (E > 1 Mov)

Sample Temperature impact Energy Lateral Expansion Shear j

Number

(. C)

(*F)

(J)

(ft Ib)

(mm)

(mils)

(*/*)

j Tangential Orientation N-69

-46

-50 13.0 9.5 0.08 3.1 3

N-72

-32

-25 23.5 17.5 0.40 15.7 7

N-67

-23

-10 66.0 48.5 0.88 34.6 18 N-71

-18 0

59.5 44.0 0.91 35.8 17 N-61

-4 25 65.0 48.0 0.94 37.0 28 N-64 10 50 101.0 74.5 1.70 66.9 34 N-70 24 75 131.5 97.0 1.37 53.9 53 N-62 38 100 143.5 106.0 1.95 76.8 64 N-63 66 150 175.0 129.0 1.86 73.2 78 N-68 93 200 200.0 147.5 2.30 90.6 100 N-65 135 275 185.0 136.5 2.27 89.4 100 N-66 177 350 193.0 142.5 2.29 90.2 100 Axial Orientation S-71

-46

-50 8.0 6.0 0.31 12.2 3

S-66

-23

-10 25.0 18.5 0.40 15.7 13 S-72

-18 0

54.0 40.0 0.80 31.5 18 S-62 10 50 79.5 58.5 1.40 55.1 31 S-64 24 75 95.0 70.0 1.35 53.1 45 S-65 38 100 74.5 55.0 1.20 47.2 33 S-67 66 150 123.5 91.0 1.31 51.6 66 S-70 93 200 139.5 103.0 1.67 65.7 87 S-69 121 250 198.0 146.0 1.86 73.2 100 S-68 177 350 170.0 125.5 1.62 63.8 100 S-63 204 400 186.5 137.5 1.45 57.1 100 5-3

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.k TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE PRAIRIE ISLAND UNIT 1 PRESSURE VESSEL WELD AND HEAT-AFFECTED ZONE METAL IRRADIATED AT 550* F, FLUENCE 1.25 x 10" n/cm8 (E > 1 Mev)

Sample Temperature Impact Energy Lateral Expansion Shear 4

(c )

(*F)

(J)

(ft Ib)

(mm)

(mils)

(%)

Number Weld Metal W-48

-46

-50 28.0 20.5 0.44 17.3 27 W-41

-32

-25 44.0 32.5 0.62 24.4 35 I

W-45

-18 0-48.0 35.5 0.71 28.0 43 W-44 10

'50 70.0 51.5 0.94 37.0 53 4

W-46 24 75 75.0 55.5 1.32 52.0 71 W-42 66 150 96.5 71.0 1.69 66.5 95 4

I W-47 121 250 129.0 95.0 1.96 77.2 100 W-43 149 300 113.0 83.5 1.97 77.6 100 l

HAZ Metal J

H-41

-101

-150 23.0 17.0 0.27 10.6 2

H-48

-73

-100 74.5 55.0 0.72 28.3 34 H-42

-18 0

111.0 82.0 1.16 45.7 79 14-43 24 75 252.0 186.0 1.68 56.1 96 H-47 66 150 158.0 116.5 1.70

$9.3 96 H-46 121 250 173.0 127.5 2.04 84.3 100 i

5-4

---<w-

,,-s,,..,.y,,

.,y--

,7.,,

,_._.,,,y

~.,, _ _ _,,,

h I

i TABLE 5-3 l

CHARPY V-NOTCH IMPACT DATA FOR THE PRAIRIP. ISLAND UNIT 1 A533 GRADE 8 CLASS 1 CORRELATION MONITOR MATERIAL j

IRRADIATED AT 550* F, FL UENCE 1.25 x 10 n/cm8 (E > 1 Mov)

Sample Temperature impact Energy Lateral Expansion Shear Number (C')

('F)

(J)

(ft Ib)

(mm)

(mils)

(*/')

R-47 52 125 12.0 9.0 0.13 5.1 9

R-44 93 200 34.0 25.0 0.54 21.3 38 R-48 99 210 38.0 28.0 0.84 33.1 34 R-43 107 225 67.0 49.5 0.92 36.2 58 R-45 121 250 91.0 67.0 1.47 57.9 65 R-41 149 300 118.5 87.5 1.20 47.2 100 R-42 177 350 99.5 73.5 1.75 68.9 100 R-46 218 425 127.0 95.5 1.56 61.4 100 5-5

l TABLE 5-4 THE EFFECT OF 550*F IRRADIATION AT 1.25 x 10" n/cm* (E > 1 Mov)

ON THE NOTCH TOUGHNESS PROPERTIES OF THE PRAIRIE ISLAND UNIT 1 j

REACTOR VESSEL BMPACT TEST SPECIMENS Aversee Awrage as em Average Average Energy ei: :;"r so N h Temp. (*F)

Lateral Espanolon Temp. (*F) so N h Temp. (*F) at Fun Shear (ft h) asassessi unstramased kramased AT unirradiated krmsated AT Unirramated krmAMed AT Unitradated irradiated A(N h) i Forging C 4

55 51

-12 28 40

-27 10 37 143.0 136 7

{

(Axial) l f

Forging C

-5 20 25

-25 7

32

-25

-5 20 158.0 142 16 j

(Tangential)

Weld

-14 46 60

-45 25 70

-57

-15 42 78.6 83 4.5 '8 1

Metal HAZ

-170

-105 65

-175

-65' 110

-200

-130 70 211.0 143 68 l

Metal Correlation 81 232 151 53 218 165 49 205 156 123.5

.85

'38.5 Monitor

a. Upper Shelf Energy increase I

I

y

~

TABLE 5-5 f

I

)

SUMMARY

OF PRAIRIE ISLAND UNIT 1 REACTOR VESSEL SURVEl!. LANCE CAPSULE CHARPY IMPACT TEST RESULTS 50 ft Ib 30 ft Ib Decrease in Trans. Temp Trans. Temp Upper Shelf Fluence increase increase Energy Material 10 n/cm8

(*F)

(* F)

(ft Ib)

Forging C 0.546 15 24 12 *l l

(Axial) 1.25 51 37 7

1 t

Forging C 0.546 39 38 15 (Tangential) 1.25 25 20 16 l

Weld Metal 0.546 3i 25 12.5 *I 1

1.25 60 42 4.5 'l 1

HAZ Metal 0.546 0

0 1.25 65 70 68 Correlation 0.546 113 110 32.5 Monitor 1.25 151 156 38.5

a. Upper Shelf Energy Increase 5-7

i

. TABLE 5-4 l

1RRADIATED TENSILE PROPERTIES FOR THE PRAIRIE ISLAND UNIT 1 PRESSURE VESSEL SHELL FORGING C AND WELD METAL FLUENCE 1.25 x 10" n/cm8 (E > 1 Mov)

Ultimate Test 0.2% Yield Tensile Fracture Fracture Fracture Uniform

~ Total -

Reduction Sample Temp.

Strenoth Strength Load Strou Strength Elongation Elongation in Ared

.!1 Material No.

(*F)

(ksi)

(kip)

(ksi)

(%)

de Forging C S-16 78 73.3 92.7 2.95 181.1 60.1 12.0 26.0 67 (Axial)

S-18 200 70.3 88.1 2.65 177.2 54.0 10.4 23.7 70 S-17 500 63.2 84.5 2.30 160.7 46.9 9.0 20.4 71 Forging C -

N-16 78 73.3 92.7 2.70 180.5 55.0 11.3 25.8 70 (Tangential)

N-18 200

-70.3 87.6 2.60 166.5 53.0 9.6 22.5 68 N-17 500 64.2 84.5 2.60 195.9 53.0 98 20.5 73 Weld W-18 78 86.6 97.8 3.30 197.1 E7.2 16.5 26.1 66 Metal W-17 200 80.5 91.7 3.15 190.8 64.2 12.5 25.5 66 W-16 500 67.2 86.1 3.10 145.0 63.2 11.3 20.9 56

(*C)

-100

-50' O

50 100 150 200 100 3g 3g'

~,

~

40 O g-20

&' e 0

100 2.5 80 24 E

/

1 60 Q/o..

1.5 1

e 40 0

1.0

,p 20

/

0.5 Uk0*

2m O

O 180 240 160 9

200 140 8_ e o

UNIRRADIATED q 120 3g o

[~ 1%

o O

/0 120 9

~

yM g/

=

a IRR ADI ATED (550* F)

E 60 O

J2-.-

51$

1.

x 10'nw 40 k

37'F

- 40 20 0

O

-100 0

100 200 300 400 TEMPERATURE (* F)

Figure 5-1.

Irradiated Charpy V-Notch Properties for Prairie Island Unit 1 Reactor Vessel Shell Forging C (Axial Orientation) 5-9

(*C)

-100

-50 0

50 100 150 200 3

6 100 80 o

I O

M g

y*

<w 5e 8*#

Oso 120 3.0 100 Q

2.5 g

o 2

g

/o.

2.0 j 80 S

- 1.5 f_

0 60 g

b AO 1.0 32 7

9/

0.5 20 o

0

- 2m 8

160 o

o

.P-2 ie

,[

UNIRRADIATED

~

160 g/.

1205

~

N g 80 IRRADIATED (550'F) w o

1.25 x 10n/cm8 6 80 O'

80 25' F l* 20* F e

40 O

5

/

20 ev#

0 o

-100 0

100 200 300 400 TEMPERATURE (* F)

Figure 5-2.

Irradiated Charpy V-Notch Properties for Prairie Island Unit 1 Reactor Vessei Shell Forging C (Tangential Orientation) 5-10

G

's

('C)

-100

-50 0

50 100 150 200 e

i i

i 3

, M-e 100 s

80 o

v e

e

~ 60

~

O O l

a-O O 8

O f

40 V,&

20 o

O 1

100 2.5 l

1 1

(

, 60' o

o p

2.0 e

0 Of A

I E

/O e

60 0

15 i

O g0 F

$ 40 0 _[

6

- 1.0 --

Q

. I,7 70*F l

20 0

- 0.5 9

0 0

1 120 160 100 UNIRRADIATED 0

j

- M I 80 O

o e

l 60 0 9 O

O RRADIATED (550* F)

- 80 9 O

O

-8 1.25 x 10a /cm8

~

n p

6 40 O/

42'F 20

] 40 0'

0 o

-100 0

100 200 300 400 TEMPERATURE (* F)

Figure 5-3.

Irradiated Charpy V-Notch Properties for Prairie Island Unit 1 Reactor Pressure Vessel Weld Metal S-11

(*C)

-150

-100

-50 0

50 100 150 i

i i

i 3

--b---*

100

^Y m

./

/

O i

5%

O O

E o

/

o h to O r g)<

%2 O O

j 20 0

100 M

g g

o e

- 2.0 g 80 o

5 O

O l

0 C0 j 40 o.110*F

'i

a 8o.[

20

- 0.5 0

~

0

'O 2M 240 0

320 O

o 200 UNIRRACIATED O

g

,- 160 0

0

,n E

O e

y 120 o

_ is0 5 8

E IRRADIATED (550* F) 80 p$

OO 0

1.25 x 10"n/cma

~/* ' '

40 O

8 o.

0 O

l

-200

-100 0

100 200 300 l

TEMPERATURE (* F) l Figure 5-4.

Irradiated Charpy V-Notch Properties for Prairie Island Unit 1 Reactor Pressure Vessel Wold Heat-Affected-Zone Metal 5-12

~

e

(*C)

-100

-50 0

50 100 150 200 i

a e

i i

i kW

^

100 e

/*2

/

0

/

80 o

Em g

e cr 5

e e

x m

e 20 2

3v 0

100 2.5 0

a U

2.0 80 o

0 e

E 60 o

e-1.5 OlO U

[ 165'F 1.0 40 2gf e

/0 i

20 e

0.5 M2 og O

0 140 9

120 l{

o 160 e

,00 UNIRRADIATED e

120 g

g

(

h **

~

$m o

_ wa

~

O--- 151

F -+*

o IRRADIATED (550* F) 40 g

1.25 x 10"n/cm8 156'F 40 f

20 O

2,y,

,e 0

0

-100 0

100 200 300 400 TEMPERATURE (* F)

Figure 5-5.

Irradiated Charpy V-Notch Properties for A533 Grade B Class 1 Correlation Monitor Material 5-13

m--

a m -

l t

n,.y.y _.,u. n s%.+ww' m' ';

s p

. f. 7, ;. 4

,3.

.x

.A..

~ -

...,.. w,a. *:.; -

. h...- : :...;<...

4..w v,.

..g~,f.:. f, :, ; -

g....g _

x.

n :.

~

...y...,

S-71 S-66 S-72 S-62 S-64 S-65 V'

S-67 S-70 3-69 S-68 S-63

-.. o; 's :g _, =..;.; T

[ l f ? : ~4f l,'

e.];, r:

*. %q.- ;

y

.7

9, jy.'r ~-

'% p.y. < :..,

; y.gQ, L.

k

.3

'.k, f..

.) ' ~

.-q

-j

' y

. 7.v;,8 c9

. k.rgtg:,., 9p&l; g.ai '

.l[ }'

',,4,. f'e.,.

. f'l, ?)f"%.,. J.'

z_~: -Q %w' f '; g:]h,,'l ';' 4.

/

1 e..:.

.: w [R

, s;.

Ii -

[,..

i fp,, '

\\,, ):. p >.*. M.

..;s k.u

~ *3 w-my; sq ;

,i>,

.._ e 3:91.. :<.

.,. ; ' - r.. er.,.

t.,

.s = n,? ;f([

g

< v.

.t y,Q.f. :

'g

e'. p

.y.

,: }.

.; yy;y;g.

,, J. {p, t;.;

~;.e..;,.

g.,.1.._ 8..;.;.

Figure 5-6.

Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 1 Shell Forging C (Axial Orientation)

M

-wusza mmemas j

E,S!E ggE!is N-69 N-72 N-67 N-71 N-61 N-64 a*

N-70 N-62 N-63 N-68 N-65 N-66

..; r g

- ; 5 **47*fg I

gh.

_ 3.'.['

: y 3 s,.

o _.. w. - +.

lIli _

l, y.

{

l ni g,

pc.k-. -.i ' W -.3

, J p., c.: 13 3 q.

gcy.j-ly4;jfe '% J^;. ' ;;.;..;

T x

Figure 5-7.

Charpy impact Specimen Fracture Surf aces for Prairie Island Unit 1 Shell Forging C (Tangential Orientation) l f

i l,

.=

i

],'A' p<, <.., y..

g

' - +

. y._...

~~

s -

3 e.

.m 4

.7 w

3. -

??,5L.;.~y.,,

[

1.

gl

' ~....

a g

. v; p

.i'.

blf QQ:R f

1

-..? 'ss

-r

?

w-48 w-41 w-45 W-44 w-46 w-42 W-47 w-43

. ;/ "

g. s. AT :...

I.ll i

5 Ih

'.lfe',. Y,i l

f. R*.

,.s

.,. m.i.

3.

7. 3.,...c k--; -4 J

FC-gg.

~

q

:, 4,;wf? d.,

k n.

. ' J,,,.

jj.**.,,.,e.

g

.'W t'

-2 g,.

4 4.

...s ;

7.. $'

D e

lll,$Y' 7g ifq-l ?b t

, ;. ~

a

-.~

r 3

w ti Figure 5-8.

Charpy impact Fracture Surfaces for Prairie Isla.1d Unit 1 Wald Metal S

5-16 8

M

d t

i 4

t p,3 kkd

..~;,.

. g.

7 t

H-41 H-48 H-42 H-43 H-47 H 46 i

1 e

.g 1-S L.-.

i

..g

~

2

,, g'. ? 3 '

O.te-4 3

.. ~. -.

t S

, k. n -

r i

I i

Figure 5-9.

Charpy impact Fracture Surfaces for i

Prairie Island Unit 1 Weld Heat-Affected-Zone Metal l

5-17

__._u.__.,..,..._____

o

7.,...,,,....

' 4 J ;.:..;.f,3,,

_m p-N f _

e.? -

'..t? _ -8....

.c, m

m m

. v.

..c. : :...

5 6

l c

.C.*: :

.', y <i.-- 2 g:.

w.

=f( j,

-q

nyly 9;.(,pi f.;J.,-

M_

.,r.1...

_m

.g g _-

r a 4..^. e g.

m 4

n

.t +

, f y,iq 2

M y,...,s.

m m

s.,.<;..,.

r

, l,4.,.;,. ^ -

m 1

1 l

R-47 R-44 R-48 R-43 s

R-45 R-41 R-42 R-46

~.

_ y 'Q., g..

.[ f f,.;--;.*

'{

g.< }_ +.yl k

p;g.y, '; ; y e'

1 4,p "1-

,,4 7;., ;.[g, s m.;. f. '- '.;

g j

j.y.g;:.4;-

A Q.

l

-f w

,., n g,,

s c. w :

...., j y.;..;.; <

M. y_, yw..

...s s

.sy k,. Q. 3

.. QI...

3..

.., w.. -

, %h [

f-

[

7 g4 v,. o g.:......-

9..

,s.v.

m xa.

.r.

a 1

.. y 4..

-e,-

js g-

.r--,,-l...%,--.

k*.... l

-W

_ [i,;'

.l%

i Figure 5-10.

Charpy impact Specimens Fracture Surfaces for Prairie Island Unit 1 A533 Grade B Class 1 Correlation Monitor Material S-18 mm.

g.

-g+

m

('C) 0 50 100 150 200 250 300 110 i

e i

i i

100

~

AN 90 A

TENSILE STRENGTH g

w 600 80 A-i m

7 500g 9

~

-8 8

400 60 0.2% YlELD STRENGTH 50 300 40 LEGEND OPEN POINTS - UNIRRADIATED CLOSED POINTS -IRRADIATED AT 1.25 x 10n/cm8 80 REDUCTION IN AREA 70 0f=

mg 60

[ 50 C

2 40 o

30 a

b TOTAL ELONGATION

-a -

a g

20 A

l e

e l

10 UNIFORM ELONGATION 0

O 100 200-300 400 500 600 TEMPERATURE (* F)

Figure 5-11.

Irradiated Tensile Properties for Prairie Island Ihit 1 Reactor Pressure Vessel Shell Forging C

( A vial Orientation) 5-19

('C) 0 50 100 150 200 250 300 110 i

i a

i 6

700 100 1

A\\_

TENSILE STRENGTH 90

}

A 3

N

-b

^

- 80 7

- 5*

3 e 70

~

E 9

~

-9 j2_ a 80 0.2% YlELD STRENGTH 50 300 40 LEGEND OPEN PCINTS - UNIRRACIATED CLOSED POINTS -IRRADIATED AT 1.25 x 10'Wem8 80 -

i 9

9

~

o 70 e-REDUCTION IN AREA 60 E s0 Cs 40 b

~

g TOTAL ELONGATION 20 2

G-f g

-;o e

UNIFORM ELONGA(ION t

1 t

t t

O 100 200 300 400 500 600 TEMPERATURE (* F)

Figure 5-12.

Irradiated Tensile Properties for Prairie Island Unit 1 Reactor Pressure Vessel Shell Forging C (Tangential Orientation) 5-20

9

(*C) 0 50 100 150 200 250 300 110 i

i i

i 700 100 90 A

TENSlLE STRENGTH N

W 600

--e 5 80 g

500 E 70 h--

E 60 0.2% YlELD STRENGTH O

- 400 50 300 40 LEGEND OPEN POINTS - UNIRRADIATED CLOSED POINTS -IRRADIATED AT 1.25 x 10n/cm2 80

=

0 O

REDUCTION IN AREA 60

$ 50 C

2 40 o

30 [

p_

TOTAL ELONGATION 20

~A g~

Q

~.-

o

~

UNIFORM ELONGATION t

t I

f f

f 0

100 200 300 400 500 600 TEMPERATURE ('F)

Figure 5-13.

Irradiated Tensile Properties for Prairie Island Unit 1 Reactor Pressure Vessel Weld Metal i

5-21

8 3

~

.c e

y

-.9

-t 2

e w

c 2

-.m n

E ow u

n.

e m

c.

o r,r,

..n c

O

~e C

o e

e H

9 C

e

? t t

3 5

o e m

.c N

a2 m.

e o

e g

3 as c

a~

e o

u

'E.>H i-g

.w v=

tc 4

ao a

w3

.9 E

o e

o o

o 8

.e e

N e

c v

CO N

(!sM) ssagis s

e

~

5-22

4 1

9

~, -

y

- >ae

~

pm -

A 1 2 3 4 6 7 8 9 5

0 1

INCHES SPECIMEN S-16 78*F 1 2 3 4 6 7 8 9 5

0 1

INCHES 1

i SPECIMEN S-18 200*F l

l l

...e..

..;_ ;...., - i; h ',..

, ;.g.

. 7c

^-....

- 7n. -

,;a ?-

[

a,,(,. v r.~.,'

'; i ; ;T 4' f'.

2

' :I^H -),.. f : l l' ;. '-l}

' ~ Y '. '. l ' l *,.!r,. -*s4.

s

. y.

......,, c; 1 2 3 4 6 7 3 9 5

0 1

INCHES SPECIMEN S-17 500*F g.

Figure 5-15.

Fractured Tensile Specimens from Prairie Island Unit 1 Pressure Vessel Shell Forging C (Axial Orientation) 5-23 m

6 7,

8 9

~

1 2 3 4 5

l 0

INCHES SPECIMEN N-16 78*F 6 7 8 9 1 2 3 4 5

1 1

0 INCHES SPECIMEN N-18 200*F 6 7 8 9 1 2 3 4 5

0 1

INCHES SPECIMEN N-17 500* F Figure 5-16.

Fracturad Tensile Specimens from Prairie Island Unit 1 Pressure Vessel Shell Forging C (Tangential Orientation) 5-24

e 6 7 8 9 1 2 3 4 5

l 0

INCHES SPECIMEN W-18 78'F S

6 7 8 9 1 2 3 4 5

0 I

INCHES SPECIMEN W-17 200* F i

em

-s

-m 1

6 7 8 9 1 2 3 4 5

0 INCHES SPECIMEN W-16 500*F Figure 5-17.

Fractured Tensile Specimens from Prair!e Island Unit 1 Pressure Vessel Weld Metal 5-25

s SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1.

INTRODUCTION Knowledge of the neutron environment within the pressure vessel-surveillance cap-sule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons. First,in the interpretation of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, Gux) to which the test specimens were exposed must be known. Second, in relating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship between the environment at various positions within the reactor vessel and that experienced by the test specimens must be estab-lished. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information, on the other hand, is derived solely from analysis.

i This section describes a discrete ordinates Sn transpon analysis performed for the Prairie Island Unit 1 reactor to determine the fast neutron (E > 1.0 Mev) flux and l

fluence as well as the neutron energy spectra within the reactor vessel and surveil-lance capsules; and, in turn, to develop lead factors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. Based on i

spectrum-averaged reaction cross sections derived from this calculation, the analy-sis of the neutron dosimetry contained in Capsule P is discussed and updated evaluations of dosimetry from Capsule V are presented.

6-2.

DISCRETE ORDINATES ANALYSIS A plan view of the Prairie Island reactor geometry at the core midplane is shown in Figure 6-1. Since the reactor exhibi:s 1/8th core symmetry, only a O'-45' sector is depicted. Six irradiation capsules attached to the thermal shield are included in 6-1 c

w

PRESSURE VESSEL o

SURVEILLANCE CAPSULE 13' (CAPSULES V. R)

/

23' (CAPSULES T. P)

THERMAL 33' (CAPSULES S. N)

SHIELD 7

ma p

wimmi,m

.S.

r if ffli ll!! !Ill f f '

l,,uunu/

nnunuu!

/

/H/uHuf l

/

//

/

//

/

REACTOR CORE I/ /

II/

///

Figure 6-1.

Prairie Island Unit 1 Reactor Geometry 6-2

~~

I 1

1 the design to constitute the reactor vessel surveillance program. Two capsules are located symmetrically at 13*,23* and 33' from the cardinal axis as shown in Figure 6-1.

A plan view of a single surveillance capsule attached to the thermal shield is shown in Figure 6-2. The stainless steel specimen container is 1-inch square and approxi-mately 63 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central S.25 feet of the 12-foot-high reactor core.

From a neutronic standpoint, the surveillance capsule structures are significant. In fact, as will be shown later, they have a marked impact on the distributions of neutron flux and energy spectra in the water annulus between the therr,11 shield and the reactor vessel. Thus,in order to properly ascertain the neutron r <ironment at the f

test specimen locations, the capsules themselves must be incluced in the analytical model. Use of at least a two-dimensional computation is, therefore, mandatory.

In the analysis of the neutron environment within the Prairie Island Unit 1 reactor geometry, predictions of neutron flux magnitude and energy spectra were made with the DOTDI two-dimensional discrete ordinates code. The radial and azimuthal distri-butions were obtained from an R, O computation wherein the geometry shown in Figures 6-1 and 6-2 was described in the analytical model. In addition to the R, O computation, a second calculation in R, Z geometry was also carried out to obtain relative axial variations of neutron flux throughout the geometry of interest. In the R, Z analysis the reactor core was treated as an equivalent volume cylinder and, of course, the surveillance capsules were not included in the model.

Both the R, O and the R, Z analyses employed 21 neutron energy groups, an Se angular quadrature, and a P cross-section expansion. The cross sections were i

generated via the Westinghouse GAMB1Tiel code system with broad group proces-sing by the APPROPOSI73 and ANISN 'l codes. The energy group structure used in I

the analysis is listed in Table 6-1.

A key input parameterin the analysis of the integrated fast neutron exposure of the reactor vessel is the core power distribution. For the analysis, power distributions representative of time-averaged conditions derived from statistical studies of long-l l

6-3

i o

term operation of Westinghouse two-loop plants were employed. These input distri-butions include rod-by-rod spatial variations for all peripheral fuel assemblies.

It should be noted that this particular power distribution is intended to produce accurate end-of-life neutron exposure levels for the pressure vessel. As such, the calculation is indeed representative of an average neutron flux and small (= 15-20%)

deviations from cycle to cycle are to be expected.

e t

m i

l i

i i

TABLE 6-1.

21 GHOUP ENERGY STRUCTURE Group Lower Energy (Mev) 1 7.79 al t

2 6.07 3

4.72 4

3.68 5

2.87 6

2.23 7

1.74 8

1.35 9

1.05 10 0.821 11 0.388 12 0.111 13 4.09 x 10-'

14 1.50 x 10-'

15 5.53 x 10-8 16 5.83 x 10-*

17 7.89 x 10~'

18 1.07 x 10-8 19 1.86 x 10-*

20 3.00 x 10

21 0.0

a. Upper energy of group 1 is 10.0 Mev.

l l

6-5 l

I o

(13', 23*, 33' )

l r fyECtMEN (12*. 22'. 32')

r j

j l

/ A

^

/

177)/n77 THERMAL SHIELD

~

Figure 6-2.

Plan View of a Reactor Vessel Surveillance Capsule 6-6 f

l\\.-

Having the results of the R,0 and R,Z calculations, three-dimensional variations of neutron flux may be approximated by assuming that the following relation holds for the applicable regions of the reactor.

$(R,Z,0,E ) = & (R,Z,0,E ) F(Z,E )

(6-1) g g

g where:

$(R,Z,0,E ) = neutron flux at point R,Z,0 within energy group g g

$(R,0,E ) = neutron flux at point R.O within energy group g obtained from g

the R,0 calculation F(Z,E ) = relative axial distribution of neutron flux within energy group g g'

obtained from the R,Z calculation 6-3, NEUTRON DOSIMETRY The passive neutron flux monitors included in the Prairie Island Unit 1 surveillance program are listed in Table 6-2. The first five reactions in Table 6-2 are used as fast neutron monitors to relate neutron fluence (E > 1.0 Mev) to measured materials properties changes. To properly account for burnout of the product isotope gener-ated by fast neutron reactions,it is necessary to also determine the magnitude of the thermal neutron flux at the monitor location. Therefore, bare and cadmium-covered cobalt-aluminum monitors were also included.

The relative locations of the various monitors within the surveillance capsules are shown in Figure 4-2. The nickel, copper, iron, and cobalt-aluminum monitors,in wire form, are placed in holes orilled in spacers at several axiallevels within the capsules.

I The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the center of the capsule.

The use of passive monitors such as those listed in Table 6-2 does not yield a direct measure of the energy-dependent flux level at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux levelincident on the various monitors may be derived from the activation measurements only if the

-irradiation parameters are well known. In particular, the following variables are of interest:

6-7

'l i

TABLE 6-2 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS 4

Target Fission Monitor Reaction Weight

Response

Product '

Yield 1

Material of Interest Fraction Range Half-Life

(%)

Copper Cu'*(n u)Co" 0.6917 E>4.7 Mev 5.27 years fron Fe"(n p)Mn" 0.0585 E>1.0 Mev 314 days Nickel Ni"(n.p)Co" 0.6777 E>1.0 Mev 71.4 days Uranium-238 *3 U*(n,f)Cs'3' 1.0 E>0.4 Mev 30.2 years 6.3 l

Neptunium-237 *l Np (n.f)Cs'2' 1.0 E>0.08 Mev 30.2 years 6.5 1

i Cobalt-Aluminum *l i '"(n,y)Co" 0.0015 0.4eV<0.015 Mev 5.27 years l

Cobalt-Aluminum t,o"(n y)Co" 0 0015 E<0.0015 Mev 5.27 years

a. Denotes that monitor is cadmium shielded.

9

o i

e The operating history of the reactor a The energy response of the monitor e The neutron energy spectrum at the monitor location a The physical characteristics of the monitor i

The analysis of the passive monitors and subsequent derivation of the average l

neutron flux requires completion of two procedures. First, the disintegration rate of l

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

The specific activity of each of the monitors is determined using established ASTM procedures.N '0 " ' "I Following sample preparation, the activity of each monitor is determined by means of a lithium-drifted germanium. Ge(Li). gamma spectrometer.

I The overall standard deviation of the measured data is a function of the precision of j

sample weighing, the uncertainty in counting, and the acceptable error in detector calibration. For the samples removed from Prairie Island Unit 1, the overall 2a deviation in the measured data is determined to be 10 percent.The neutron energy spectra are determined analytically using the method described in Section 6-1.

Having the measured activity of the monitors and the neutron energy spectra at the locations of interest, the calculation of the neutron flux proceeds as follows.

The reaction product activity in the monitor is expressed as t

R=

f; Y a(E)$(E)

Il~'

i),

d 6-2 P

s 6-9 5

+

-Y er 9-

-4 % fi9wybM+^

>s_-etamu--*wNT--"'eir**wp-F&-3ewwa h-g 9 Twiee wr -crot-T w-yP'9^w-m*e7+ -ry t w w wm p-

re-p&==*

e 9-mi: e m

4 UN i

where.

R = induced product activity f

N o = Avagadro's number A = atomic weight of the target isotope ti = weight fraction of the target isotope in the target material Y = number of product atoms produced per reaction a(E) = energy-dependent reaction cross section

$(E) = energy-dependent neutron flux at the monitor location with the reactor at full power PJ = average core power level during irradiation period j Pmax = maximum or reference core power level A = decay constant of the product isotope tj = length of irradiation period j td = decay time following irradiation period j Since neutron flux distributions are calculated using multigroup transport methods and, further, since the prime interest is in the fast ' neutron flux above 1.0 Mev, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-2) is replaced by the following relation.

FW(E> 1.0 Mev) a(E) 4(E)dE

=

.E 6-10

1 l

1 s

where:

N

9#9 a(E) $ (E)dE o

G*1 pm N

$ (E)dE s i o u..

,9 o o u,,

Thus, equation (6-2) is rewritten N

N P

o i

-At )e-At i

d R=

fi y a & (E > 1.0 Mev)

(1-e A

P max jet or, solving for the neutron flux, R

c (E > 1.0 Mev) =

N N

P

-At

-At fya (1-0

)e 6-3 j

l=1 The total fluence above 1.0 Mev is then given by N

~

p.

$ (E > 1.0 Mev) = $ (E > 1.0 Mev){

max tj 6-4 I

3.,

(

6 l-

e where:

N tj = total effective full power seconds of reactor operation Pmax up to the time of capsule removal An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered Co"(n.6)Co" data by means of cadmium ratios and the use of a 37-barn 2200 m/sec cross section. Thus, hD-1 J Rbare h

D i

@Th =

6-5 N

No Pj

-At ) e-A t j

d f ya (1-e i

P A

max j=1 where:

are D is defined as rcd covered 6-4.

TRANSPORT ANALYSIS RESULTS Results of the So transport calculations for the Prairie Island Unit 1 reactor are summarized in Figures 6-3 through 6-8 and in Tables 6-3 through 6-5. In Figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vessel inner radius,1/4 thickness location, and 3/4 thickness location are presented as a function of azimuthal angle. The influence of the surveillance capsules on the fast neutron flux distribution is clearly evident. In Figure 6-4, the radial distribution of maximum fast neutron flux (E > 1.0 Mev) through the thickness of the reactor pressure vessel is shown. The relative axial variation of neutron flux within the vessel is given in' Figure B-5. Absolute axial variations of fast neutron flux may be obtair.ed by multiplying the levels given in Figures 6-3 or 6-4 by the appropriate values from Figure 6-5.

In Figure 6-6 the radial variations of fast neutron flux within surveillance capsules V and P are presented. These data, in conjunction with the maximum vessel flux, are 6-12

3 used to develop lead factors for each of the capsules. Here the lead factor is defined as the ratio of the fast neutron flux (E > 1.0 Mev) at the dosimeter block location (capsule center) to the maximum fast neutron flux at the pressure vessel inner radius.

Updated lead factors forall of the Prairie Island Unit 1 surveillance capsules are listed in Table 6-3.

Since the neutron flux monitors contained with the surveillance capsules are not all located at the same radial location, the measured disintegration rates are analytically adjusted for the gradients that exist within the capsules so that flux and fluence levels may be derived on a common basis at a common location. This point of comparison was chosen to be the capsule center. Analytically determined reaction rate gradients for use in the adjustment procedures are shown in Figures 6-7 and 6-8 for Capsules V and P. All of the applicable fast neutron reactions are included.

In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. The neutron energy spectrum calculated to exist at the center of each of the Prairie Island Unit 1 surveillance capsules is given in Table 6-4. The associated spectrum-averaged cross sections for each of the five fast neutron reactions are given in Table 6-5.

6-5.

DOSIMETRY RESULTS The irradiation history of the Prairie island Unit 1 reactor is given in Table 6-6.

Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsules V and P are listed in Tables 6-7, and 6-8, respectively. The data are presented as measured at the actual monitor locations as well as adjusted to the capsule center. The measured results for both Capsules V and P were obtained by Westinghouse. All adjustments to the capsule centers were based on the data pre-sented on Figures 6-7 and 6-8.

j The fast neutron (E > 1.0 Mev) flux and fluence levels derived for Capsules V and P are j

presented in Table 6-9. The thermal neutron flux obtained f rom the cobalt-aluminum monitors is summarized in Table 6-10. Due to the relatively low thermal neutron flux at the capsule locations, no burnup correction was made to any of the measured activities. The maximum error introduced by this assumption is estimated to be less than 1 percent for the Nisa (n p) Cosa reaction and even less significant for all of the other fast neutron reactions.

6-13 i

..;. ? y '.. n. -

n,.

I

' L. ::,..

. 1

_.y

j 2;, '

' - ~. ; y : p-

,a.-a n.c :

y' Using the iron data presented in Table 6-9, along with the lead f actors given in Table 6-3, the fast neutron fluence (E > 1.0 Mev) for Capsules V and P as well as for the reactor vessel inner diameter are summarized in Table 6-11 and Figure 6-9. The agreement between calculation and measurement is excellent, with measured flu-ence levels of 1.25 x 10 and 5.46 x 10 compared to calculated values of 1.21 x 10

and 6.14 x 10 n/cm2 for Capsules P and V, respectively. Further, the graphical representation in Figure 6-9 indicates the accuracy of the transport analysis for Prairie Island Unit 1 and supports the use of the analytically determined fluence trend curve for predicting vessel toughness at times in the future. Projecting to end-of-life, a summary of peak fast neutron exposure of the Prairie island Unit 1 reactor as derived from both calculation and measurement may be made as fo; lows.

Fast Neutron Fluence (n/cm*)

Surface 1/4 T 3/4 T Capsule P 4.48 x 10

2.94 x 10

8.67 x 10's Capsule V 3.87 x 10

2.62 x 10" 7.72 x 10

8.19 x 10

Average measurement 4.17 x 105 2.78 x 10

Calculation 4.30 x 10

2.87 x 10'8 8.46 x 10

These data are based on 32 full-power years of rperation at 1650 MWt.

Based on the new capsule to vessel inner wall lead f actors identified in Table 6-3 and the new capsule withdrawal schedule identified in ASTM E185-79, it is recommended that future capsules be removed from the reactor per the following schedule.

Capsule Capsule Vessel Lead Removal Fluence identity Location Factor Time (n/cm8)

V 77*

3.37 1.34 EFPY *l 5.46 x 10

l P

247' 1.94 4.6 EFPY 'l 1.25 x 10

I R

257' 3.37 6.0 EFPY 2.72 x 10

T 67' 1.94 15.0 EFPY 3.91 x 10'8 N

237' 1.79 32.0 EFPY 7.70 x 10

S 57' 1.79 Standby

a. These Capsules have been Removed.

6-14

)

g e4

. ~.

l

,f i

TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 MEV) AND LEAD FACTORS FOR PRAIRIE ISLAND UNIT 1 SURVEILLANCE CAPSULES C'apsule A::Imuthal

$ (E>1.0 Mov)

Lead identification Location (n/cm'-sec)

Factor V

13*

1.45 x 10" 3.37 R

13*

1.45 x 10" 3.37 T

23' 8.33 x 10'o 1.94 P

23' 8.33 x 10'o 1.94 S

33' 7.67 x 10'o 1.79 N

33' 7.67 x 10'o 1.79 6-15

o

~'

t e

t i

l

(~

TABLE 6-4 l

CALCULATED NEUTRON ENERGY SPECTRA ATTHE DOSIMETER BLOCK LOCATION FOR PRAIRIE ISLAND UNIT 1 SURVEILLANCE CAPSULES i

l Gg Neutron Flux (n/cm8-sec)

No.

eg-E!:: V&R Capsules T & P Capsules S & N 1

8.17 x 108 5.99 x 10' 5.26 x 10*

2 2.68 x 10' 1.99 x 108 1.75 x 10' l

3 4.43 x 108 3.08 x 10' 2.73 x 10' 4

4.98 x 108 3.18 x 10' 2.88 x 10' 5

8.66 x 10' 5.20 x 10' 5.14 x 10' I

6 1.70 x 10'o 1.01 x 10'o 9.26 x 10' 7

2.46 x 10

1.41 x 10'o 1.30 x 10

8 3.53 x 10'o 1.97 x 10'o 1.39 x 10'o 9

4.67 x 10'o 2.53 x 10'o 2.35 x 10'o 10 5.04 x 10'o 2.67 x 10

2.48 x 10'o 11 1.67 x 10" 8.66 x 10'o 8.03 x 10'o 12 2.11 x 10" 1.05 x 10" 9.76 x 10'o 13 9.42 x 10'o 4.65 x 10'o 4.34 x 10

14 7.11 x 10ia 3.52 x 10'o 3.28 x 10'o 15 5.67 x 10'o 2.80 x 10'o 2.62 x 10'o 16 1.32 x 10" 6.41 x 10'o 5.99 x 10'o 17 1.03 x 10" 5.07 x 10'o 4.73 x 10'o 18 1.06 x 10" 5.14 x 10'o 4.82 x 10

19 8.41 x 10'o 4.09 x 10'o 3.83 x 10'o 20 9.34 x 10'o 4.52 x 10'o 4.23 x 10'o 21 2.97 x 10" 1.51 x 10" 1.36 x 10" l

e 6-16

f i

l TABLE 6-5 j

SPECTRUM AVERAGED REACTION CROSS SECTIONS AT THE i

DOSIMETER BLOCK LOCATION FOR PRAIRIE ISLAND UNIT 1 SURVEILLANCE CAPSULES F (barns)

Reaction Capsules V & R Capsules T & P Capsules S & N Fe5d(n,p)Mn5' O.0595 0.0683 0.0666 Nise(n.p)Co58 0.0811 0.0912 0.0893 Cu'3(n a) Coa 0.000404 0.000517 0.000494 U238(n f)F.P.

0.333 0.345 0.344 Np237(n.f)F.P.

2.93 2.80 2.82

=

a(E)&(E)dE g=

o

"&(E)dE 1 Mew 1

I 6-17

b TABLE 6-6 IRRADIATION HISTORY OF PRAIRIE ISLAND UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE P l

Pg irradiation Decay *3 PJ Prnax Tiene Tiene P

Month (MW)

(MW) anas (days)

(days) 12R3 107 1650

.065 31 2905 1#4 0

1650

.000 31 2874 294 480 1650

.291 28 2846 3/74 118 1650

.072 31 2815 4/74 480 1650

.291 30 2785 5/74 0

1650

.000 31 2754 664 0

1650

.000 30 2724 i

7#4 837 1650

.507 31 2693 864 1025 1650

.621 31 2662 964 163 1650

.099 30 2632 10/74 255 1650.

.155 31 2601 1194 1515 1650

.918 30 -

2571 12H4 1566 1850

.949 31 2540 in5 1296 1850

.787 31 2509 265 1277 1650

.774 28 2481 3/75 1605 1850.

.972 31 2450 465 1263 1850

.765 30 2420 565 903 1650

.547 31 2389 6/75 1286 1650

.779 30 2359 7n5 1231 1650

.746 31 232(I 8/75 1044 1650

.633 31 2297 9/75 1097 1650

.665 30 2267 10R5 1370 1850

.830 31 2236 11R5 1534 1650

.930 30 2206 12H5 1545 1650

.936 31 2175 1R6 1534-1650

.930 31 2144 2R6 1140 1650

.691 29 2115 l

a. Decay times are referenced to December 14.1981 and August 23.1976 for monitors of Capsules P i

and V. respectively. The Np and U monitors of Capsule V were counted on August 25.1976. Capsule V was removed in March.1976.

l 6-18 m

-.. ~

l TABLE 6-6 CONTINUED IRRADIATION HISTORY OF PRAIRIE ISLAND UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE P Py irradiation DecayN PJ Pmax Time Time P

Month (MW)

(MW) max (days)

(days) 366 114 1650

.069 31 2084 4n6 0

1650

.000 30 2054 Sn6 940 1650

.570 31 2023 6/76 1490 1650

.903 30 1993 7/76 1442 1650

.874 31 1962 8/76 1523 1650

.923 31 1931 9/7d 1033 1650

.626 30 1901 10n6 1612 1650

.977 31 1870 1196 1580 1650

.958 30 1840 12/76 1589 1650

.963 31 1809 1n7 1566 1'650

.949 31 1778 2/77 1501 1650

.910 28 1750 3/77 770 1650

.467 31 1719 4n7 0

1650

.000 30 1689 Sn7 1338 1650

.811 31 1658 i

6n7 1462 1G50

.886 30 1628 7R7 1551 1650

.940 31 1597 867 1559 1650

.945 31 1566 9n7 1523 1650

.923 30 1o36 10/77 1509 1650

.914 31 1505 1167 1573 1650

.953 30 1475 1267 160w 1650

.975 31 1444 in8 1561 1650

.946 31 1413 2n8 1508 1650

.914 28 1385 3n8 1238 1650

.751 31 1354 4R8 516 1650

.313 30 1324 Sn8 1573 1650

.953 31 1293

a. Decay times are referenced to December 14.1981 and August 23.1976 for monitors of Capsules P arid V, respectively. The Np and U monitors of Capsule V were counted on August 25.1976. Capsule V was removed (n March,1976.

6-19

TABLE 6-6 CONTINUED IRRADIATION HISTORY OF PRAIRIE ISLAND UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE P tel P

Irradiation Decay P

P J

max Time Time P

Month (MW)-

(MW) max (days)

(days) 6/78 1496 1650

.907 30 1263 7/78 1156 1650

.700 31 1232 868 1349 1650

.818 31 1201 9R8 1353 1650

.820 30 1171 10n8 1577 1650

.956 31 1140 1168 1485 1650

.900 30 1110 1298 1582 1650

.959 31 1079 in9 1588 1650

.963 31 1048 2R9 1600 1650

.970 28 1020

~

3/79 1516 1650

.919 31 989 469 247 1650

.150 30 959 Sn9 982 1650

.595 31 -

928 6/79 1334 1650

.809 30 898 7R9 150 1650

.091 31 867 8n9 1263 1650

.766 31 836 969 1381 1650

.837 30 806 10R9 449 1650

.272 31 775 1169 666 1650

.404 30 745 12R9 1482 1650

.886 31 714 1/80 1524 1650

.924 31 683 2/80 1206 1650

.731 29 654 3/80 1500 1650

.909 31 623 4/80 1482 1650

.898 30 593 5/80 1437 1650

.871 31 562 6/80 1395 1650

.845 30 532 7/80 474 1650

.288 31 501 8/80 1028 1650

.623 31 470

a. Decay times are referenced to December 14.1931 and August 23.1976 for monitorspf Capsules P and V, respectively. The Np and U monitors of Capsule V were counted on August 25,1976. Capsule V was removed in March,1976.

6-20

7

~

TABLE 6-7 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE P I

Saturated Activity Adjusted Saturated Activity Reaction Radial (DPS/gm)

(DPS/gm) and Axial Location Location (cm)

Capsule P Calculated Capsule P Calculated Fe"(n p)Mn" Top 157.87 4.25 x 105 3.83 x 105 4.10 x 105 Mid-top 157.87 3.31 x 108 3.83 x 10' 3.19 x los Mid-bottom 157.87 3.93 x 105 3.83 x 10*

3.80 x 108

{

Bottom 157.87 4.41 x 108 3.83 x 108 4.25 x 105 Average 3.84 x 108 3.70 x 105

^

Cu'2(n a)Co**

Mid-top 158.87 3.63 x 105 2.45 x 105 4.32 x 105 Mid-bottom 158.87 3.38 x 105 2.45 x 105 4.01 x 105 Average 4.17 x 10

2.91 x 10'-

NiS*(n p)CoS8 Middle 158.87 5.41 x 10' 4.51 x 108 6.45 x 10' 5.38 x 10' i

Np22'(n,f)Cs'2' I

Middle 158.10 4.48 x 10' 3.69 x 10' 4.48 x 10' 3.69 x 10' U22*(n.f)Cs'2' Middle 158.10 5.81 x 105 4.59 x 105 5.81 x 105 4.59 x 105

TABLE 6-8 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE V Saturated Activity Adlusted Saturated Activity Reaction Radial (DPS/gm)

(DPS/gm) and Aulaf Location Location (cm)

Capsule V Calculated Capsule V Calculated l

Fe"(n.p)Mn" Top 157.87 5.43 x 105 5.98 x 10' 5.16 x 108 Mid-top 157.87 4.99 x 10' 5.98 x 108 4.74 x 108 Middle 157.87 5.09 x 108 5.98 x 108 4.84 x 10*

Mid-bottom 157.87 5.28 x 108 5.98 x 10*

5.02 x 10'

{

80 Bottom 157.87 5.64 x 105 5.98 x 105 5.36 x 10*

5.02 x 10' 5.68 x 108 Average I

l Cu*'(n.a)Co" Mid-top 158.87 3.77 x 105 3.29 x 105 4.48 x 105 Mid-bottom 158.87 4.26 x 105 3.29 x 105 5.07 x 105 4.77 x 105 3.91 x 105 Average l

NiS*(n.p)Co**

Middle 158.87 6.29 x 10' 7.17 x 10' 7.44 x 10' 8.48 x 10' Np22'(n.f)Cs

Middle 158.10 6.11 x 10' 7.01 x 10' 6.11 x 10' 7.01 x 10' US(n.f)Cs'2' Middle 158.10 7.93 x,105 7.64 x 105 7.93 x 105 7.64 x 10*

-a k

TABLE 6-9 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULES P AND V Adjusted Saturated Activity p(E > 1.0 Mev)

$(E > 1.0 Mev)

(DPS/gm)

(n/cm*-sec)

(n/cm8)

Capsule Reaction Measured Calculated Measured Calculated Measured Calculated E!

P Fe5d(n.p)Mn5* '

3.84 x 108 3.70 x 105 8.61 x 10

8.33 x 10

1.25 x 10

1.21 x 10

l Cu*'(n,a)Co" 4.17 x 105 2.91 x 105 1.22 x 10" 1.77 x 10

NiS*(n,p)Co58 6.45 x 10' 5.38 x 10' 1.01 x 10" 1.46 x 10

Np23'(n,f)Cs'*'

4.48 x 10' 3.69 x 10' 9.68 x 10'o 1.40 x 10

U228(n,f)Cs'3' 5.81 x 105 4.59 x 105 1.06 x 10" 1.53 x 10

V Fe5*(n p)Mn54 5.02 x 10*

5.68 x 10*

1.29 x 10" 1.45 x 10" 5.46 x 10

6.14 x 10

Cu'3(n a)Co" 4.77 x 105 3.91 x 105 1.79 x 10" 7.58 x 10

Ni"(n.p)Co" 7.44 x 10' 8.48 x 10' 1.30 x 10" 5.50 x 10'a Np22'(n,f)Cs'2' 6.11 x 10' 7.01 x 10' 1.37 x 10" 5.80 x 10

U2'*(n.f)Cs'8' 7.93 x 10*

7.64 x 105 1.50 x 10" 6.35 x 10

m

~

TABLE 610 RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULES P AND V Saturated Activity (dps/gm) pg Capsule Axial Location Bare Cd-covered (n/W-sec)

P Top 7.56 x 10' 2.67 x 10' 8.58 x 10'8 Bottom 9.02 x 107 2.84 x 107 1.09 x 10" V

Top 1.90 x 10' 8.25 x 108 1.28 x 10" Bottom 2.23 x 10' 7.68 x 108 1.73 x 10" w

9 6-24

v t

TABLE 6-11

SUMMARY

OF NEUTRON DOSIMETRY RESULTS FOR CAPSULES P AND V l

Irradiation Vessel Calculated Time p(E > 1.0 Mov)

((E > 1.0 Mew)

Lead Fluence,

~ Vessel Fluence m

h Capsule (EFPS)

(n/cm*-sec)

(n/cm')

Factor (n/cm2) '

(n/cm')

P 1.45 x 10*

8.61 x 10

1.25 x 10

1.94 6.44 x 10 '

6.23 x 10

V 4.23 x 10' 1.29 x 10" 5.46 x 10

3.37 1.62 x 10's 1.82 x 10

=

6

3 2

10" 8

6

- SURVEILLANCE CAPSULES 4

i 1

i iiE2 3

d 2

____ PRESSURE VESSEL IR lE b

10 Z

1/4 T LOCATION 8

6 4

3/4 T LOCATION 2

10' 0

10 20

' 30 40 50 AZIMUTHAL ANGLE (deg)

Figure 6-3.

Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mov) within the Pressure Vessel Surveillance Capsule Geometry 6-26 r

i k

l-l l

l l

I 10" 8

3 L

i-5 6

l 4

l IR l-I E

8 m

1/4 T O

g 2

I E

=

l m

u o

sS 1/2 T g

a 104 w

$8 E

[

3 m

m 6

3/4 T

~

z g

4 f

I

~

OR 2

I I

E VESSEL HO 2

l e

2 10' l

160 162 164 166 168 170 172 174 176 178 180 182 184 186 188 RADIUS (cm)

Figure 6-4.

Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mov) within the Pressure Vessel l

i l

6-27 l

1

s O

~

8 6

4 2

~

10" 8

x 3

6 2

O 4

2 z

w2 2

E d

e 10-8 2 8

6 4 2 3

8 5

~

O 2

~

To VESSEL CLOSURE HEAD t o-a I

I l

l l

-300 200

-100 0

100 200 300 DISTANCE FROM CORE MIDPLANE (cm)

Figure 6-5.

Relative Axisi Variation of Fast Neutron Flux (E > 1.0 Mov) within the Pressure Vessel 6-28

/

9' NEUTRON FLUX (n/cm8-sec) s s

8 e

n a

a n

a a

2 I

I l

l l l Ill i lIll l

I I

I I Ill l Ill

.ms c

xm e

55 3 I~

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to

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ff h

O i

=*1 s\\\\\\\\\\\\\\

~

,s CAN O

o A\\\\\\\\\\\\\\

s :D h 3 $.

2RE a

c^

N

"O

?

o m

i m

"V4 m

e Ta g H

3

$'E O

I aEo 9

ea C

o_

FRONT MONITORS -

- 157.87

<So g$ ~%

DOSIMETER BLOCK -

- 158.10

!E a

5 4

gn U

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l SECTION 7 REFERENCES f

1. Yanichko, S. E., Lege, D.J.," Northern States Power Company Prairie Island Unit No.1 Reactor Reactor Vessel Radiation Surveillance Program." WCAP-8086, June 1973.
2. ASTM Designation E-185-73," Surveillance Tests for Nuclear ReactorVessels"in ASTM Standards (1974), Part 10, pp. 314-320, Am. Soc. for Testing and Materials.
Philadelphia, Pa.,1974.
3. Davidson, J. A., Anderson, S. L. and Scott, K. V., " Analysis of Capsule V from Northern States Power Company Prairie Island Unit No.1 Reactor Vessel Radia-tion Surveillance Program," WCAP-8916, August 1977.
4. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Standard Review Plan, NUREG-75/087 Sect. 5.3.2, " Pressure-Temperature Limits," November 1975.
5. Soltesz, R. G., Disney, R. K., Jedruch, J. and Ziegler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation, Volume 5
- Two-Dimension Discrete Ordinates Transport Technique " WAN L-PR(LL)034, Volume 5, August 1970.
6. Collier, G., et. al., "Second Version of the GAMB1T Code," WANL-TME-1969, November 1979.
7. Soltesz, R. G., et. al.," Nuclear Rocket Shielding Methods. Modification, Updating and input Data Preparation - Volume 3 Cross-Section Generation and Data j
Processing Techniques," WANL-PR(LL)034, August 1970.
I l
7-1
37 1,. -
N-
--a
-V
8. Soltesz, R. G., et. al.," Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation - Volume 4 - One-Dimensional Discrete Ordinates Transport Technique " WANL-PR(LL)034, August 1970.
9. ASTM Designation E261-70," Standard Method for Measuring Neutron Flux by Radioactivation Techniques,"in ASTM Standards (1975), Part 45, Nuclear Stand-ards, pp. 745-755, Am. Society for Testing and Materials, Philadelphia, Pa.,1975.
10. ASTM Designation E262-70," Standard Method for IVsasuring Thermal Neutron Flux by Radioactivation Techniques,"in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 756-763, Am. Society for Testing and Materials, Philadelphia, Pa.,
1975.
11. ASTM Designation E263-70," Standard Method for Meeuring Fast-Neutron Flux by Radioactivation of iron,"in ASTM Standards (1975), Part 45, Nuclear Stand-ards, pp. 764-769, Am. Society for Testing and Materials, Philadelphia, Pa.,1975.
. 12. ASTM Designation E481-73T, " Tentative Method of Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver,"in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 887-894, Am. Socioty for Testing and Materials, Pfiiladelphia, Pa.,1975:
13. ASTM Designation E264-70," Standard Method for Measuring Fast-Neutron Flux by Radioactivation of Nickei," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 770-774, Am. Society for Testing and Materials, Philadelphia, Pa.,
1975.
l l
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APPENDIX A HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A-1.
INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature). The most limiting RTNDT of the mate-rial in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ARTN DT-RT i
NDT s designated as the higher of either the drop weight nil-ductility transition temperature (TNDT) or the temperature at which the material exhibits at least 50 ft Ib of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.
RT i
NDT ncreases as the material is exposed to fast-neutron radiation. Thus, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT ue d
to the radiation exposure associated with tha,t time period must be added to
' the original unirradiated RTNDT. The extent of the shift in RT i
NOT s enhanced by certain chemical elements (such as copper and phosphorus) present in reactor vessel steels. The Regulatory Guide 1.99 trend curves which show the effect of fluence and copper and phosphorus contents on ART f
NDT orreactorvesselsteels are shown in Figure A-1.
Given the copper and phosphorus contents of the most limiting material, the radiation induced ARTWT can ba estimated from Figure A-1. Fast-neutron fluence (E > 1 Mev) at the 1/4 T (wall thickness) and 3/4 T (wall thickness) vessel locations are
=
given as a function of full-power service life in Figure A-2. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RTN OT-A-2.
FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture-toughness properties of the Prairie Island Unit 1 reactor vessel materials are presented in Table A-1. The fracture toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accord-ance with the NRC Regulatory Standard Review Plan.I4 The postirradiation fracture A-1
___-___m_
_ _ _ - - - - - - - - - - - - - - - - - - - - ' - - - - - - - - - - - - ' - - - ^ - ~
A = [40 + 1000 (% Cu - 0.08) + 5000 (% P - 0.008)l [f/10] '
ppgg LigtT 300 200 i
gd\\
g l
F 5
l 9 im y
g m
I A
/.
0.35 0.30 0.25 0.20% Cu 0.15% Cu 0.10% Cu
/
/
/
/
LOWER LIMIT 40
%P = 0.012
% Cu = 0.08 j
/
/
% P = 0.008 A
30 d WELD METAL
/
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1
$ SHELL FORGING C I
I I
I I
I I
I I
I I
^
20 2x 10
4 6
8 10'*
2 4
6 8
10
2 4
6 I
FLUENCE. n/cm' (E > 1MeV)
Figure A-1.
Effect of Fluence, Copper Content, and 4
NDT w Reactw f
l Phosphorus Content on ART Vessel Steels per Regulatory Gu,ide 1.99 O
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1/4 T 10" 2
6 3/4 T 4
l
~
I UZ 2
Sw z
O E 10
m Ea 6
4 2
I I
I I
I I
t o
O 5
10 15 20 25 30 i
SERVICE LIFE (EFFECTIVE FULL POWER YEARS)
Figure A-2.
Fast Neutron Fluence (E > 1.0 Mov) as a Function of Full Power Service Life I
A-3 L
~
4 toughness properties of the reactor vessel beltline material were obtained directly from the Prairie Island Unit 1 Vessel Material Surveillance Program.
A 3.
CRITERIA POR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves forvarious heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined g
thermal and pressure stresses at any time during heatup and cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at IR s obtained from the reference fracture toughness curve, defined in i
that time. K Appendix G to the ASME Code.I'3 The KIR curve is given by the equation:
KIR = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160H
( A-1)
IR s the reference stress intensity factor as a function of the metal tempera-i where K ture T and the metal reference nil-ductility temperature RTNOT.Thus,the governing equation for the heatup-cooldown analysis is defined in Appendix G to the ASME Code!81 as follows':
gy + K :S K g (A-2)
CK it i
A-4
e TABLE A-1 i
REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)
NMWD 50 ft Ib/35 mils NMWD Lateral Expansion Upper Shelf RT Material Cu P
NDTT Temperature NDT Energy Component Type
(%)
(%)
(* F)
(* F)
(* F)
(ft Ib)
Closure Head Dome A533 Gr. B, Cl.1
-4 64 *3 1
4*l 75 *l l
1 i
I Head Flange A508 Cl. 3
-4 12 'l
-4 *I 84 *3 I
I 1
?
Vessel Flange A508 Cl. 3
-4 41 al 1
t l
4 al 77.5 *3 l
Injection Nozzles A508 Cl. 3
-22
-114'l
-22I'l 97*l I
1
+5 39*3 t
S'l 92 *3 I
l inlet and Outlet Nozzle A508 Cl. 3 I
l l
Upper Shell A508 Cl. 3
-4 39 'l
-4 'I 85 *3 Inter. Shell A508 Cl. 3 0.06 0.013
+14 14 14 143 Lower Shell A508 Cl. 3 0.07 0.014
-4 45
-4 134 Trans. Ring A508 Cl. 3
+5 63 'l I
1 I
S'l 79 *1 Bottom Head A533 Gr. 8, Cl.1
-4 57 *l
-3 al 68.5 'l 1
t I
Inter. to Lower Shell Girth Weld Sub-Arc Weld 0.13 0.017 0
10 0
78.5 l
Weld HAZ HAZ O*l
<-100 0
211 l
I
a. Estimated using the NRC Standard Review Plan.
4 i
4 4
1 e
3 i, i,ii<,,- -,
e O
1 where gy s the stress intensity factor caused by membrane (pressure) stress K
i it s the stress intensity factor caused by the thermal gradients K
i IR s a function of temperature to the RTNDT of the material K
i C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical IR s determined by the metal At any time during the heatup or cooldown transient, K i
temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from tempera-ture gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K, for the reference flaw are computed. From it equation (A-2), the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation cf the allowable pressure-versus-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall.
During cooldown,thecontrollinglocation of theflawisalwaysattheinsideof thewall,
because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are gen-erated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, wh'ereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of KIR at the 1/4 T location for finite cooldown rates than for steady-state operation.
A-6
/
D s-l Furthermore, if conditions exist such that the increase in KIR exceeds K, the it calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature i
at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal
temperature at the crack tip lags the coolant temperature; therefore, the K f
IR orthe 1/4 T crack during heatup is lower than the K f
IR orthe1/4Tcrackduringsteady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower KIR's do not offset each other, and the pressure-temperature ct. / based on steady-state conditions no longer represents a lower bound of all simnar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to insure that at any coolant temperature the i
lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup an.d the time (or coolant tempera-ture) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
l 4
A-7 1
,---,.-.--n--,,...--.w,---.,.-,-~,,,n,,,-.
.--,-$------.---,_n--w rn,,,,,,,,---
_. _.... ~.....
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
Then, composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instroments by 5
the values indicated on the respective curves.
+
t A-4.
HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System
~
have been calculated using the methods discussed in paragraph A-3. The derivation of the lirnit curves is presented in the NRC Regulatory Standard Review Plan.W
[
Tran'sition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program.
f Charpy test specimens from Capsule P irradiated to 1.25 x 10" n/cm8 indicate that the mu s in R NDTof Fr p vely sh n by ure
.The s r
ell within the appropriate design curve (Figure A-1) prediction. Heatup and cooldown i
limit curves for normal operation of Prairie Island Unit 2 for up to 20 effective-full-power years (EFPY) were presented in the Capsule radiation surveillance program report.Pl These heatup and cooldown curves are applicable for Prairie Island Unit 1 up to 7.4 EFPY, and they are shown in Figures A-3 and A-4.
Allowable combinations of temperature and pressure for specific temperature e
change rates are below and to the right of the limit lines shown on the heatup and j
cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line, shown in Figure A-3. This is in addition to other criteria which must be met before the reactor is made critical.
t A-8 1 -. _ _...
c o
l The' leak ' test limit curve shown in Figure A-3 represents minimum temperature
- requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of reference.[2.si Figures A-3 and A-4 define limits for insuring prevention of nonductile failure.
J A-9 i
'o 3000 MATERIAL PROPERTY BASIS WELD METAL Cu = 0.13%
LEAK TEST LIMIT f
INITIAL RTN OT = 0* F 2600 AT 7.4 EFFECTIVE FULL POWER YEARS 2W RTNDTAT 1/4 T THICKNESS = 110* F l
2200 =
RT AT 3/4 T THICKNESS = 74*F l
NDT 3
b g 2000 =
l W 1800 b l
I E
=
1 g1600 g i
e I
a.
1400 =
0
$ 1200 =b UNACCEPTABLE ACCEPTABLE OPERATION OPERATION o
5 1000 C 2
Z C
/,RITICALITY LIMIT 800 =
Z HEATUP RATES UP 600 C TO 100* F/HR 400 3 200 3.
0 O
50 100 150 200 250 300 350 INDICATED TEMPERATURE ('F)
Figure A-3.
Prairie ledand Unit 1 Reactor Coolant System Hestup Umitations Applicable for Periods up to 7.4 Effective Full Power Years.
Margins of 60 peig kn 10'F are included for Possible instrument Error A-10 N
. i
.a t by l-1
\\
}l l
l l
l 2600 C MATERI AL PROPERTY BASIS 2400 2 WELD METAL Cu = 0.13%
INITI AL RTN DT = 0* F 2200 Z AT 7.4 EFFECTIVE FULL POWER YEARS 2000 3 RTNDT AT 1/4 T THICKNESS = 110*F z RTNDTAT 3/4 T THICKNESS = 74* F
(
y 1800 e
2 y 1600 t
2
$ 1400 C w
=
1200 C q 1000 =_
9
=
800 z
COOLDOWN RATES
'F/HR 600 0
I 20 60
~
400 100 200 2 0
I I
l 0
50 100 150 200 250 300 INDICATED TEMPERATURE (* F)
Figure A-4.
Prairie Island Unit 1 Reactor Coolant System Cooldown Limitations Applicable for Periods up to 7.4 Effective Full Power Years.
Margins of 60 psig and 10* F are included for Possible instrument Error j
1 A-11 1
r 1
,*,s APPENDIX A REFERENCES
1. " Fracture Toughness Requirements," Branch Technical Position MTEB No. 5-2.
Section 5.3.2-14 in Standard Review Plan, NUREG-75/087,1975.
2. ASME Boiler and Pressure Vessel Code, Section ill, Division 1 - Appendices,
" Rules for Construction of Nuclear Vessels," Appendix G, " Protection Against Nonductile Failure," pp. 481-489,1980 Edition, American Society of Mechanical Engineers, New York,1980.
3. Pressure-Temperature Limits," Section 5.3.2 in Standard Review Plan, NUREG-75/987,1975.
4. Yanichko,S. E.,and Anderson,S.L " Analysis of Capsule Tfro.m Northern Stat 6s Power Company Prairie Island Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9877, March 1981.
5. " Pressure-Temperature Limits," Section 5.3.2 in Stande.rd Review Plan, NUREG-75/087,1975.
A-12
<

v t e Letter Sequence Other
TAC:49027, (Open) TAC:59974, (Approved, Closed) TAC:59975, (Approved, Closed)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML20069H205, ML20069H222
MONTHYEAR1981-09-23 [Table View]

ML20069H222

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