ML18348A744

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Response to Request for Additional Information Reactor Vessel Material Surveillance
ML18348A744
Person / Time
Site: Palisades Entergy icon.png
Issue date: 05/23/1978
From: Hoffman D
Consumers Power Co
To: Ziemann D
Office of Nuclear Reactor Regulation
References
Download: ML18348A744 (60)
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Text

consumers

  • Power company General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201
  • Area Code 517 788-0550

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May 23, 1978 co *:-)

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u Director, Nuclear Reactor Regulation < -i*=

Att: Mr Dennis L Ziemann, Chief CJ 5 --1 JTl 0 ..

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Operating Reactors Branch No 2 (_,')

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US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR P.l\LISADES PLANT - REACTOR VESSEL MAT~~IAL SURVEILLANCE Consumers Power provided a partial response to the NRC request dated May 20, 1977 for information concerning reactor vessel materials and associated sur-veillance program on July 29, 1977 .

The attachment responds to NRC staff ~uestions and is as complete as Consun.ers Power or Combustion Engineering records allow.

Four heats of Type Mil B-4 weld wire and five heats of E 8018 electrodes were used to fabricate the weld seaIUs of the beltline regions. lfo as welded 11 chemical composition data is available; Further documentation of 11 bare wire 11 chemistries could possibly be obtained from CE's suppliers' quality control records but the data obtained would be of no practical significance. This is because the nbare wire" chemistry of the T-,tpe Mil B-4 wire is modified by the type of flux used in welding and the copper coating applied to the wire. The information desired by the NRC ~uestionnaire is the composition of the as-deposited welds. in the reactor vessel. Therefore, no further useful data on the composition of materials employed in the Palisades reactor vessel welds is available.

The surveillance program weldment was made with the same type of filler wire and flux as the longitudinal weld seams in the reactor vessel beltline but not the same heat of wire or batch of flux.

In general, the Palisades sii:rveillance program complies with 10 CFR 50, Ap-pendices G and H and with the i*ecommendations of ASTM E-185-66, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. 11 The one difference between the plant surveillance prog:ram a.nd the require-ments of Appendix H is that the surveill e.nce :program allows for the attaclmie!lt REGULATORY DOCI<fT .fltE COPY ,,-- - - - - - - - - - -

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  • of surveillance capsules to the inside wall of the beltline region of the reactor vessel. This modification to 10 CFR 50, Appendix H, Section C.2-II was .described in topical report CENPD-155P, "C-E Procedure for Design, Fabri-cation, Installation and Inspection of Surveillance Specimen Holder Assemblies."

The NRC has previously reviewed CENPD-155P and found all the procedures de~

scribed therein as acceptable.

Fracture toughness and tensile property data for the surveillance materials was obtained from the "Final Report on Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Properties." This report, prepared by Battelle Columbus Laboratories, describes the result of the baseline sur-veillance testing and provides mechanical property data consistent with the latest edition of the ASME code and 10 CFR 50, Appendix G.

David P Hoffman Assistant Nuclear Licensing Administrator CC: JGKeppler, USNRC

,

  • 295 I
  • REPLY TO NRC QUESTIONNAIRE ON REACTOR VESSEL MATERIALS OF CONSTRUCTION AND SURVEILLANCE PROGRAMS .*

Question l: Provide the estimated maximum fl uence (E> l MEY) at the inner surface of the reactor vessel wall, as of March 31, 1977.

Response: The inner surface of the vessel at the point of maximum fluence received 2.23 x 10 18 n/cm 2 (E>MeV) as of March 31, 1977 Question 2: Provide the effective full power years (EFPY) of operation accumulated as of March 31, 1977.

Response: The effect full power years (EFPY) accumulated as of

    • Question 3:

March 31, 1977 are 1.9~ .

Identify the firm or firms that fabricated yo.ur reactor vessel.

Response: Combustion Engineering, Inc. fabricatr-d the reactor vessel for Consumers Power Co.'s Palisades Nuclear Power Station.

Quest ion 4( a): Provide a sketch of the reactor vessel showing all materials, including welds, in the beltline region and provide an identification number for each material.

Response: Figure 1 is provided to identify all beltline region materials for the Palisades reactor vessel. The beltline region was determined in accordance with 10CFRSO, Appendix G, using the peak axial fluence

  • profile on the inner surface wall and an uncontrolled weld chemistry.

l

295 The beltline region welds are the three longitudinal welds of the lower shell (Weld Seam No's 3-112A, 3-1128, and 3-112C); the lower to intermediate shell girth seam (Weld Seam No. 9-112); the three longitudinal welds of the intermediate shell (Weld Seam No's 2-112A, 2-1128 and 2-112C); the intermediate to upper shell girth seam (Weld Seam No. 8-112); and the Tower 14" of the three longitudinal welds of the upper shell (Weld Seam No. 1-112A, 1-1128, and 1-112C) . . The plate and forging materials in the reactor vessel are identified by piece number and code number. The beltline region plates are the three plates of the lower shell:

Piece No. 112-0SA Code No. D-3804-1 Piece No. 112-058 Code No. D-3804-2 Piece No. 112-0SC Code t~o. D-3804-3 and the three plates of the intermediate shell:

Piece No. 112-04A Piece No. 112-048 Code No. D-3803-1

. Code Na. D-3803-2 Piece No. l12-04C Code No. D-3803-3 Question 4(b): Provide the .following information for each of the welds in the beltline region:

(1) Shop control number or procedure qualification number; (2) Filler metal and heat number; (3) Type of flux and batch number; (4) Welding process (sub arc, electroslag, manual metal arc, etc.);

2

295

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(5) Post-wel.d heat treatment; (6) Chemical composition (particularly Cu, P and S content);

(7) Drop weight TNDT (8) RTNOT, (9) Charpy upper shelf energy (unirradiated);

(10) Tensile properties (unirradiated);

(11) Firm performing weld if more than one firm parti-cipated in welding; (12) The maximum end-of-life fluence at the vessel inner wall.

Response: The information for Question 4(b) Items 1-4 and 6-10 is in Tables 1 .1 through 1.12*. The submerged are welds for the Palisades reactor vessel used a dual wire feed consisting of the Mil .B-4,:wire and a 1/16 0 Nickel 200 11 additive. Root welds were chipped back and manually

  • welded with* the E-8018 electrode. The difference in as-welded chemical composition may be seen in Table 3.2 where the F~ce weld contains the nickel additive.

11 11 Response 4(b)5: Post weld heat treatment is carried out at 1150 +25°F with heat up and cooldown rates from Qnd to 600°F of l00°F per hour. The time at temperature for a specific weld depends on its sequence in the vessels' fabrication.

Intermediate and final stress relief times are chosen so as to provide 40.hrs. at 1150°F for the plates in the vessel. Thus the longitudinal seams in the inter-mediate and lower shell assemblies would have 40 hr.

stress relief treatments while the closing girth seam would see only the final vessel assembly stress relief treatment of 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> at 1150 +25°F .

  • 3

295 Response 4(b)ll: All fabrication of the Palisades reactor vessel was perfonned by Combustion Engineering, Inc.

Response 4(b)l2: The maximum end-of-life exposure at the point of maximum fluence on the inner surface wall is predicted to be no greater than 3.84 x lo 19 n/cm 2 for the Palisades reactor vessel.

Question 4(c): Provide the following information for each of the plates or forgings in the beltline region:

(1) Plate or forging serial number; (2) Plate or forging heat number;

  • (3) Plate or forging material speciftcation number; (4) Plate or forging supplier; (5) Plate or forging heat treatment; (6) Chemical composition (particularly Cu, P and S content);

(7) Drop weight TNOT (8) RT~mT (unirradiated);

(9) Charpy upper shelf energy (unirradiated);

(10) Tensile properties (unirradiated);

(11) The maximum end-of-life fluence at the vessel inner wall.

  • Response: The infonnation for Question 4(c) Items 1-4 and 6-10 is provided in Tables 2.l through 2.6 for the Palisades reactor vessel.

Response 4( c) 5: ASTM A-302 (modified) Grade B plate is a quenched and tempered product. The heat treatment performed by Combustion Engineering, Inc. consisted of: Austenization 4

295

  • at 1575 +25°F for four (4) hours, water quenched then tempered at 1225 :_25°F for four (4) hours. The interme-diate and final vessel assembly stress relief treatments, (described in response to Question 4(b)5) consists of forty (40) hours at 1150 +25°F, furnace cooled to 600°F.

Response 4(c)ll: The maximum end-of-life exposure at the point of maximum fluence on the inner wall surface is predicted to be no greater than 3.84 X lo 19 n/cm 2.

Question 5(a): List the weld, plate, and forging materials included in the vessel material surveillance program.

Response: The base material for the Palisades reactor vessel materials surveillance program is intermediate shell plate no. D-3803-3, piece no. 112-048. The weld material was fabricated by welding plate D-3803-1 to plate D-3803-2 using the detailed -weld procedure used for the longitudinal weldseams. The same type of filler wire and flux as were*us~d in the fabrication of the longitudianl welds were used in the fabrica-tion of the surveillance material but not the same heat of wire or batch of flux.

Question 5(b): For each weld listed in 5(a), provide the information requested in items 1 through 11 of Question 4(b).

Response: The information for Items 1 through 4 and Items 6 through 10 is provided in Table 3. 1. The fracture toughness and tensile property data are taken from "Final Report; Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Properties. 111 5

295 The post weld heat treatment (Item 5) consisted of a forty-hour stress-relief treatment at 1125 +25°F with heat-up and cooldown rates from and to 600°F of 100°F per hour. The weld for the surveillance material program was made by Combustion Engineering, Inc. (Item 11).

Question S(c): For each plate or forging specimen listed in 5(a) provide the information listed in Items 1 through 10 of Question 4(c).

Response:* The information for Items 1 through 4 and Items 6 through 10 is provided in Table 3.2. Item 5, Question S(c) is answered in Item 5 of Question 4(c). The fracture toughness and tensile property data are taken from* "Final Report; Palisades Pressure Vessel Irradia-tion Capsule Program: Unirradiated Mechanical Pro-perties.111

  • Question 5(d): Provide a copy of the *report which describes the sur-veillance program for your reactor vessel(s), if available.

Response: A copy of "Recommended Program for Irradiation Surveil-lance of Palisades Reactor Vessel Materials," dated August, 1968 is attached.

References:

1) J. S. Perrin and E. 0. Fromm, "Final Report on Palisades Pressure Vessel Irradiation Capsule Program: Unir-radiated Mechanical Properties 11 , (Battelle Columbus Laboratories, August 25, 1977).

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Figure 1 PALISADES REACTOR VESSEL

~ACTOn vt:ss::L BEL Tlll\!E MATERIALS NOT SHO\VN INTEl\MEDIATE SHELL WELD SEAM ~Jo. 2-112C LOWER SHELL

. WELD SEAM No. 3-112B WELD SEAM No. 3-112C PLATE No. D-3804-3 UPPER SHELL .

WELD SEAM N0.1-112A*

WELD SEAM N0.1-112B

  • WELD SEAM NO .1-112 ~.~-..!--------------'1-'t'::f;::I 42" ID OUTLET NOZZLE .....--........_ Q)

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30" !D.

INLET NOZZLE UPPER TO INTERMEDIATE INTERfi1!:D!ATE SHELL SHELL GI HTH SEAM-.-~>-t=v l:ONGiTUD!NALWELD WELD No: 8-112 SE.-~i\1 No. 2-112B INTt=RiviC:DIATE SHELL---1-~ I I _--ll'*JTEiir*.iEDIATE SHELL PLATE No. D-3803-*1 ~ PLATE rJo. D-3803-3 lf\JTERMEDIATE SHELL l!'JTERMEDit\TE-TO*LOWER LONGITUDINAL WELD SHELL GIF1TH SE.~M SEAr*!i No. 2 -11 ZA V'*JE LD f'*Jo. 9 -112 INTERMEDIATE SHELL PLATE r-Jo. D-3803-2 --;-r--LOWEn SHELL PLATE No~ D-3804-2 LOWER SHELL PLATE-----+-+--'="

No. D-3804-1 LOWER SHELL LONGITUDl!\IAL WELD SEAM No. 3-112A .

  • REACTOR VESSEL TABLE 1.1 Palisades Beltline Region.Material Weld Seam No. 2-112 A/C Reference Drawing E 232-112-07 Filler Metal ____ ___._

Raco 3 3/16 ___ Meat No. W5214

. Flux Type _ _ _L_in_d_e_l_0_9_2_ __ Batch No. 3617 Process ____S_u_bm-'-e-"-r.._ge_d'--A'--r_c__ Weld Procedure No. ' WB-2966A-l l 2-0 Composition (Filler wire. w/o) c Mti p s Si Mo Ni Cu

.13 l.90 .010 .009 .04 *51 NA NA Composition (As welded, w/o) c Mn p s Si Mo Ni Cu

.077 1.05 .021 *.012 .26 .50 1.20 NA CX>

Fracture Toughness:

Drop .weight TNDT(°F) --~NA~-- Charpy 1 V1 Notch {ft-lb)@ 10 °F, 35, 39, 48 RTrmT(0F) _ _ _ _ _ _3_0_ _ MTEB Position 5-2 Paragraph 1.1(4)

Upper Shelf Energy (ft-1 b) "-'-tlA-'----

Tensile Properties:

Yield Strength (Ksi) - -65.3  % Elongation {2 11 ) ___ 2_5_.o_ __

Tensile Strength (Ksi) 88.4 Reduction of Area (%} _*_6_9._6_ __

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

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TABLE 1.2 \.0 '

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Palisades Beltline Region Material 1 9-112 Reference Orawi ng -~~E~2;...;.3~2-_l;._;1_;;;2_-0;;_:7_ __

Weld ~earn No.

Filler Metal E-8018 1/411 Heat No. LODG Batch No.

Flux T y p e - - - - - - - - - -

Process Manual Arc Weld Procedure No. MA-33-8(4)

Composition (Filler wire, w/o) Si s p Mn C Ni Mo Cu

.49 . 011 . 009 1.02 .081 .93 . 21 NA Composition (As welded, w/o)

Fracture Toughness:

Charpy 'V' Notch (ft-lb)@ 10 °F, 62, 83, 99 Drop Weight TNDT(°F) -----"N~A'-----

RT NDT(°F) ___________1_0___ MTEB Position 5-2 Paragraph 1. 1(4)

Upper Shelf Energy (ft-lb) -'-NA_ __

Tensile Properties:

Yield Strength (Ksi) 72.9  % Elongation (2 11 ) 26.0 Tensile Strength (Ksi) 87.9 Reduction of Area (%) 71.3 L'

TABLE 1.3 /'", ~

"'-J Palisades Beltline Region Material Weld Seam No. 9~112 Reference Drawing** E 232-112-07

--~--~-~~~

Fi 11 er Metal 84 >3/16 11

-~--"-------

Heat No. 27204 Flux Type Linde 124 Batch No. 3687 Process _ _ _ _s_u_bm_e_r_..g_ed_A_r_c_ Weld Procedure No. WG- 2966A- l .l 2-2 Composition (Filler wire, w/o) C Mn P S Si Ni Cr Mo Cu

.21 1.51 *.014 .010 .08 1.07 .07 *.54 . 12 Composition (As welded, w/o) 0 Fracture Toughness:

Drop Weight TNDT(°F) _ _ _N_A_ __ Charpy 'V' Notch (ft-lb)@ 10 °F, 35, 48 1 42 RTNDT(of) _ _ _ _ _ _ _3_0_ __ MTEB Position 5-2 Paragraph 1.1(4)

Upper Shelf Energy (ft-lb) NA Tensile Properties:

Yield Strength (Ksi) 74.7  % Elongation (2 11 ) 26.0 Reduction of Area(%)

Tensile Strength (Ksi) 90.9

-64.7

. TABLE 1.4 N

\D

-....J Palisades Beltline Region Material Weld Seam No. 9-112 Reference Drawing** E- 232-112-07

~~~~~~~---

Filler Metal E-8018 1/4 Heat No. J BF G Batch No.

Manual Arc Weld Procedure No. MA-33-8(4)

Composition (Filler wire, w/o) Si s p Mn C Ni Mo Cu

.43 .010 .005 l. 25 . 081 1.29 .30 NA Composition (As welded, w/o)

Fracture Toughness:

Drop Weight TNDT(°F) NA Charpy 1 V1 Notch (ft-lb)@ 10 °F; .40, 44, 31 RTNDT(°F). 30 MTEB Position 5-2 Paragraph l. 1(4)

Upper Shelf Energy (ft-lb)

-NA- - -

Tensile Properties:

Yield Strength (Ksi) 89.4  % Elongation (2 11 ) 25.0 Tensile Strength (Ksi) 100.0 Reduction of Area (%) 68.5

--~---

TABLE 1.5 N

lO

""'-J Palisades Beltline Region Material Weld Seam No. 9-112 . Reference Drawing _ _ _..;..E_2_3_2_-_l_l2_-_0_7_ __

Filler Metal B4 3/16 Heat No. 27204 Flux Type Linde 1092 Batch No. 3714

'---~-'----

Process Submerged Arc Weld Procedure No. WG-2966A-ll2 Composition (Filler wire, w/o) Mn p ~ s Si Ni Cr Mo Cu

.. 21 1.81 .014 .010 .08 1. 07 . 07 .54 . 12 Composition (As welded, w/o)

Fracture Toughness:

Drop Weight TNDT(°F) _ _ _N_A_ __.. Charpy 'V' Notch (ft-lb)@ 10 °F, 71, 57, 42 RTNOT(oF) _ _ _ _ _ _ _ _3_0_ __ MTEB Position 5-2 Paragraph l. 1(4)

Upper Shelf Energy (ft-lb) NA Tensile Properties:

Yield Strength (Ksi) 62.2  % Elongation (2 11 ) 27.0 Tensile Strength {Ksi) 80.7 Reduction of Area(%) 67.7

N>

TABLE 1.6 l.O Palisades Beltline Region Material Weld Seam No. 8-112 Ref ere nee Drawing ____E_2_3_2_-_1_l2_-_0_7_ __

Filler Metal E-8018 1/4 11 Heat No. COFC Flux T y p e - - - - - - - - - - Batch No.

Process Manual Arc Weld Procedure No. WG-2966A-112-l Composition (Filler wire, w/o) NA Composition (As welded, w/o) NA w

Fracture Toughness:

Drop Weight TNOT( 0 F) _ _ _N_A_ __ Charpy 1 V1 Notch (ft-lb)@_*_ °F, - - - - - - -

RTNOT(of) _ _ _ _ _ _ _N_A_ _ MTEB Position 5-2 Paragraph Upper Shelf Energy (ft-lb) NA Tensile Properties:

. Yield Strength (Ksi)

- -NA- - -  % Elongation (2")

- - -NA- - - -

Tensile Strength (Ksi) NA

Reduction of Area (%) NA

TABLE 1.7

  • N

.... lO

-....J Palisades Beltline Region Material Weld Seam No. 8-112 Reference Drawing E 232-112-07

---~-------

Filler Metal E-8018 1/411 Heat No.78-478

-~-~-----

Batch No.

Process Manual Arc Weld Procedure No. WG-2966A-112-l

-~-~-------

Composition (Filler wire, w/o) NA Composition (As welded, w/o) NA

+::- Fracture Toughness:

Drop Weight TNDT( 0 F) _ _ _N_A_ __ Charpy 1 V1 Notch (ft-lb)@ __ °F, RT NDT(oF) ________N_A_ __ MTEB Position 5-2 Paragraph

~

Upper Shelf Energy (ft-lb) NA Tensile Properties:

Yield Strength (Ksi) NA  % Elongation (2 11 ) - - -NA- - - -

Tensile Strength (Ksi) Reduction of Area (%) NA

-NA

  • -***.............. --~ . -~ .............. _.~****- **~*--**

TABLE 1.8 Palisades Beltline Region Material Weld Seam No. - ~ ~ 8-112

; . _ _ ____ Reference Drawing _ _ _E;;c_2_3.....;.2-_1_1_2_-0_7_ __

Filler Metal Raco 3 3/16 11

~_...;,;~;;.,,_,;;~~--- Heat No. 348009 Flux Type ____L1_*n_d_e_l0_9_2_ __ Batch No. 3692 Process _ _ ___;;S=u=bm=e:. :.r_,.,g_;;_ed,;.;,_;_A;. ;_r_;;_c__ Weld Procedure No. __ WG-2966A-112-l

__;.,;_~....:.....;;__.;._;,..;.;_...;..__

Composition (Filler wire, w/o) c Mn p s Si Mo Cu

.14 2.01 .010 .017 .040 . 51 NA Composition (As welded, w/o) NA 01 Fracture Toughness:

Drop Weight TNDT( 0 F) _ ___;_;N,;. .;_A_ _ Charpy 1 V1 Notch (ft-lb)@ 10 °F, 46, 56, 59 RTNDT(oF) - - - - - - - = l _ O_ _ MTEB Position 5-2 Paragraph 1. 1(4)

Upper Shelf Energy (ft-lb) NA Tensile Properties:

Yield Strength (Ksi) 65.4  % Elongation (2 11 ) 29.0 Tensile Strength (Ksi) 81.0 Reduction of Area (%) 66.9

-~----

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

TABLE l. 9 Palisades Beltline Region Material Weld Seam No. 3-112 Re~air Reference Drawing ---~E_.2;;..;3:....;;2;_-..-1...;..;12;;;...-_.0-7_ __

Filler Metal E-8018 3/16 11 Heat No. CBBF Flux Type Batch No.

Process Manual Arc Weld Procedure No. RWG-2966A-O Composition (Filler*wire, w/o) NA Composition (As welded, w/o) NA Fracture Toughness:

Drop Weight TNDT(°F) _ ____;N_A_ __ Charpy 'V' Notch (ft-lb)@ 10 °F, 135, 128, 172 RTNDT(oF) _ _ _ _ _ _ _ _l_O__ MTEB Position 5:..2 Paragraph l. 1(4)

Upper Shelf Energy (ft-lb) NA Tensile Properties:

Yield Strength (Ksi) 74.6  % Elongation (2 11 ) 27.0 Tensile Strength (Ksi) 87.8 Reduction of Area (%) 72.3

_,;..=.;...;.._._ __

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

TABLE 1.10 Palisades Beltline Region Material Weld Seam No. 3-112 B/C Re~air Reference Drawing _ _ _E;;;;..._2~3=2-_l~l;.,;;;;2'--0_7'----

Filler Metal E-8018 3/16 11 Heat No. CBBF Flux Type Batch No.

Process Manual Arc Weld Procedure No. RWC-2966-A(l)

Composition (Filler wire, w/o) NA Composition (As welded, w/o) NA Fracture Toughness:

Drop Weight TNDT( 0 F) _ _ _N_A_ __ Charpy 1 V1 Notch (ft-lb)@ 10 °F, 135, 128, 172 RTNDT(oF) _ _ _ _ _ _ _ _l_O_ _ MTEB Position 5-2 Paragraph '* l. 1( 4) *

~~-~~-'-----

Upper Shelf Energy (ft-lb) _NA..;.___

Tensile Properties:

Yield Strength (Ksi) 74.6  % Elongation (2") 27.0 Tensile Strength (Ksi) 87~8 Reduction of Area (%) 72.3

..... - ........... **~- .-* **--** ---- - . *-* .... - -** ...... - ***-*---*--**** . ----** - *- __ .,_, ____.. _----

  • -----~-- --~- *~*-** . --~-~-.. ~.............
  • TABLE 1. 11 Palisades Beltline Region Material Weld Seam No. 3-ll2A/C Reference Drawing _ _ _E_2_3_2-_l_l_2_-_07_ __

Filler Metal Raco 3 3/16 11 Heat No. 34B009 Flux Type Linde 1092 Batch No. 3692 Process _ _ _ _S_,;.u_bm_e__.;.r_...g_ed....;_;_A.;.;...r~c-- Weld Procedure No. WC - 2966 A- l l 2 Composition (Filler wire, w/o) c Mn p s Si Mo Cu

.14 2.01 . 010 . 017 . 040 . 51 NA Composition (As welded, w/o) NA Fracture Toughness:

Drop Weight TNDT( 0

~) ___ N_A_ __ Charpy 1 V1 Notch (ft-lb)@ 10 °F, 46, 56, 59 RTNDT(oF) _ _ _ _ _ _ _ _1_0_ _ MTEB Position 5-2 Paragraph 1. 1(4)

Upper Shelf Energy (ft-lb) NA Tensile Properties:

Yield Strength (Ksi) 65.4  % Elongation (2 11 ) 29.0 Tensile Strength (Ksi) 81.0 Reduction of Area (%) -66i9 ----

. . . . . . . . . . . . *** * ........ _ ........ *,* * ' . . . . . . . . . . . . _._ ...1.. _ _ _ ....... _.

TABLE 1.12 Palisades Beltline Region Material Weld Seam No. 3""'.112 A/C Reference Drawing E 232-112-07 Filler Metal Raco 3 3/16" Heat No. W5214

~

Flux Type Linde 1092 Process Submerged Arc Batch No.

Weld Procedure No.

3692 WC-2966A-112 e*

Composition (Filler wire, w/o) NA Composition (As welded, w/o) NA l.O Fracture Toughness:

Drop Weight TNDT( 0 F) ____N_A_ __ 1 1 Charpy V Notch (ft-lb)@ °F, - - - - - - -

RT NDT(°F) -----------N_A_ __ MTEB Position 5-2 Paragraph---------------

Upper Shelf Energy (ft-lb} _N_A_ __

Tensile Properties:

Yield Strength (Ksi} NA  % Elongation (2") NA Tensile Strength (Ksi)

-NA- - - Reduction of Area (%) NA

  • *
  • _j

TABLE l.13 Palisades Beltline Region Material Weld Seam No. 1-112 A/C Reference Drawing E 232-112-07 Filler Metal Raco 3 3/16 Meat No. \~5214 Flux Type Linde 1092 Batch No. 3617 Process Submerged Arc Weld Procedure No. WB-2966A-112-0 1-Compos it ion (Filler wire, w/o) c Mn p s Si Mo Ni Cu

.13 l.90 .010 .009 .04 . 51 NA NA Composition (As welded, w/o) c Mn p s Si Mo Ni Cu

.077 l.05 .021 .012 .26 .50 1.20 NA

~ Fracture Toughness:

Drop Weight TNDT(°F) NA Charpy 1 V1 Notch (ft-lb)@ _1_0_°F, _3_5~,_3_9~,_4_8 __

RTNDT(oF} 30 MTEB Position 5-2 Paragraph _____l_.1_;(.._4_._)_ __

Upper Shelf Energy (ft-lb} _~_IA_ __

Tensile Properties:

Yield Strength (Ksi} 65.3  % Elongation (2") 25.0 Tensile Strength (Ksi} 88.4 Reduction of Area (%) -69.6 ----

TABLE 1.14 Palisades Beltline Region Material lJe 1d Seam No. 1-112 A/C Reference Drawing E 232-112-07

~----

Fi 11 er Meta 1 E-8018 3/16 11 H~at No. CB BF Flux Type Process Manual Arc l~e 1d *Procedure No. ___R_W_C_-_29_6_6_-A_(~l~)_ _

Composition (Filler wire, w/o) NA Composition (As welded, w/o) NA Fracture Toughness:

N Drop Weight TNDT(°F) NA Charpy V1 Notch (ft-lb)@ 10 °F, 135, 128, 172 in NDT( OF) 10 MTEB Position 5-2 Paragraph 1.1(4)

Upper Shelf Energy (ft-lb) -NA- - -

Tensile Properties:

Yield Strength (Ksi) 74.6  % Elongation (2 11 ) 27.0 Tensile Strength (Ksi) 87.8 Reduction of Area (%) -72.3

TABLE 2.1 Palisades Beltline Region Material Piece No. 112-04-A Code No. D-3803-1 Heat No. C-1279-3 Reference Drawing No. E 232-138-01

~

Specification SA302-B Hod.

Supplier .Lukens Steel Co.

Composition (w/o) C Mn P S Si Ni Mo Cu Al V

.20 1.41 .013 .021 .23 .48 .46 ~25 .037' .003 N

N Fracture Toughness:

Drop Weight TNDT(°F) -30 RTNDT(oF) .0 MTEB Position 5.2 Paragraph ___1_._l(._3...._)__,(._b~)_ _

Upper Shelf Energy (ft-lb) -89.7 MTEB Position 5.2 Paragraph ___1_._2_ _ _ __

Tensile Properties:

Yield Strength (Ksi) 64.4 %Elongation (2") - - -28.5 Tensile Strength (Ksi) 85.5 Reduction of Area (%) - -69.3

      • ***-* - * * - * - - - - - - - - - - - - - - - - - - - - - - - * * * - - - - - * * - * - * * - - - - * - - * * - - - - - * * - * - - * * - - - - * - * * * - - * * * * - ** * * - * * * * - - - -...- - - - - - - - - - - , . - - * * * - * - * -
  • h . . . . - * - * - _ _ ..........__.. ...... * ..... **- ...... ~ -*

TABLE 2.2 N U)

Palisades Beltline Region Material Piece No. 112-04 B Code No. D-3803-2

--"""-'--'--~--~---~

Heat No. A-0313-2 Reference Draw1ng No. E 232-138-01

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

Specification -~---"'----'-'--=------

SA302-B Mod.

Lukens Steel Co.

Compos it ion {w/o) C Mn P S Si Ni Mo Cu Al V

.19 l.35 .011 .026 .27 .50 .49 . 25 .022 .003 w Fracture Toughness:

Drop Weight TNDT{°F) -30 RTNDT{°F) -30 MTEB Position 5.2 Paragraph ___l_._l(._3_._)_.(._b....._)_ _

Upper Shelf Energy (ft-lb) -86.0 ---- MTEB Position 5.2 Paragraph 1.2 Tensile Properties:

Yield Strength (Ksi) _ 65.6  % Elongation (2 11 ) ___ 2~9_.8.;..____

Tensile Strength (Ksi) 87.8 Reduction of Area (%) __7_;_0_.7_ __

TABLE 2.3 N

u:

Palisades Beltline Region Material Piece No. 122-04 c Code No. D-3803-3 Heat No. C-1279-1 Reference Drawing No. E 232-.138-01 Specification SA302-B Mod.

Supplier Composition (w/o)

Lukens Steel Co.

c

.22 Mn 1.44 p s

.013 .021 Si

  • 21 Ni

.48 Mo

.46 Cu

.25 Al

.037 v

.003 N

.p. Fracture Toughness:

Drop Weight TNDT(°F) -30 RT NOT( of) 0 MTEB Position 5.2 Paragraph ___l_._l(._3_._)__,(.__b_._)_ _

Upper Shelf Energy (ft-lb) 91.2 MTEB Position 5.2 Paragraph --- 1.2- - - - -

Tensile Properties:

Yield Strength (Ksi) - -65.5

---  % Elongation (2") ~------

30.0 Tensile Strength (Ksi) 86.3 Reduction of Area (%) __7_0_.o_ __

TABLE 2.-4 N

lD Palisades Beltline Region Material Piece No. 112-05 A Code No. D-3804-1 Heat No. C-1308-1 Reference Drawing No. E 232-138-01

~

Specification SA302-B Mod.

Supplier Lukens Steel Co.

Composition (w/o) C Mn P S Si Ni Mo Cu

.23 1.22 .017 .024 .24 .45 .47 NA N Fracture Toughness:

U1 Drop Weight TNDT(°F) -30 RTNDT(oF) 0 MTEB Position 5.2 Paragraph ___1_.1_(~3......)_,('-b.._)_ _

Upper Shelf Energy (ft-lb) MTEB Position 5.2 ___1_._2_ _ _ __

-70.9

--- Paragrap~

Tensile Properties:

% Elongation (2 11 )

Yield Strength (Ksi) 70.2 - - -26.0 Tensile Strength (Ksi) 93.5 *Reduction of Area (%). - - 67.5

TABLE 2.5 Palisades Beltline Region Material Piece No. ll2-05B Code No. D-3804-2

--.:....:..::~=--------

Heat No. C-1308-3 Reference Drawing No. E 232-138-01

~

Specification SA302-B Mod.

Suppl i er ___L..::...:u.:.:.ck~en.:.:.cs:.....--=.S~te.:....:e:...:l~Co:.....:._ _ __

Composition (w/o) c Mn p s Si Ni Mo Cu

.23 1.25 .018 .022 .24 .50 .50 NA rv O"I Fracture Toughness:

Drop Weight TNDT(°F) -40 RTNDT(°F} -30 MTEB Position 5.2 Paragraph _ __;_1..:. . .1;__,(>. :;.3_,_)_,(-"'-b.L.-)_ _

Upper Shelf Energy (ft-lb) 76.0 MTEB Position 5.2 Paragraph 1.2 Tensile Properties:

Yield Strength (Ksi) _ _6_6._6_ _  % Elongation (2 11

) --~2~7-'-.0'----

Tensile Strength (Ksi) 89.5 Reduction of Area (%) -*-6-'-8_.;....l""-----

TABLE 2.6 Palisades Beltline Region Material Piece No. ll2-05C Code No. D-3804-3 Heat No. ___B_-_5_..;,.29_4'----"-2_ _ _ _ _ __ Reference Drawing No~ E 232-138-01 Specification SA302-B Mod.

Supplier Lukens Steel Co.

Composition (w/o) C Mn P S Si .Ni Mo Cu

.22 1.27 .010 .020 .25 .* 54 .48 NA Fracture Toughness:

Drop Weight TNDT(°F) -30 RTNDT(°F) -25 MTEB Position 5.2 Paragraph ___l_.1_(._3..._)_,(.__b.._)_ _

Upper Shelf Energy (ft-lb) 75.0 MTEB Position 5.2 Paragraph ___l~.-"2_ _ _ __

Tensile Properties:

Yi.eld Strength (Ksi) _ _6_3._7_ _  % E1ong at ion ( 211 ) __ __;;:2:..:. 7. :. ;8::__

Tensile Strength (Ksi) _ 85.6 Reduction of Area (%) _ 68.. 3

TABLE 3.1 Palisades Surveillance Program Material Weld Seam No. 9-112 Reference Drawing NA Fi 11 e*r Metal Mil B-4 Heat No. 27204 Flux Type Linde 124 Batch No. 3687 t **:

Process Submerged Arc Weld Procedure No. WG-2966A-ll2-2 Composition (As welded, Root, w/o) c Mn p s Si Ni Cr Mo Cu

.088 1*01 .011 .010 .25 .63 .05 .55 .26

  • Composition {As welded, Face w/o) c Mn p s Si Ni Cr Mo '. Cu

.086 l.02 .011 .010 .22 1.27 .03 .52 .22

~ *Fracture Toughness:

Drop Weight TNDT(°F) _____N_A_ __

RT NDT(oF) _ _ _ _ _ _ _ _0_ __ MTEB Position 5-2 Paragraph ___l_.l___..(_l)..___

Upper Shelf Energy ______ 12_0_ __

  • Tensile Properties:

Yield Strength (Ksi)  % Elongation {l 11) _ _ _ _ _ 32_._3_ _ __

- - - - -64.3 Tensile Strength (Ksi) - - - -82.0

--- Reduction of Area {%) - - - - 70.l---

TABLE 3'.2 Palisades Surveillance Program Material Piece No. 112-04R Code No. D-3803-1

~

Heat No. C-1279-3 Reference Drawing No. NA Specific~tion SA 302-B Mod.

Supplier ___L_u_k_e_ns_S_t_ee_l_C_o_.__

  • Composition (w/o} c Mn p s Si Ni Cr Mo Cu Cb

.22 1.55 .011 .019 .23 .53 .73 .58 .25 NA B Co Al Ti As Sn Zr v NA NA .037 NA NA NA NA .003

  • Fracture Toughness: (Transverse}

Drop Weight TNDT(°F} -10 RTNDT(oF) -5 Upper Shelf Energy (ft-lb) - 105

  • Tensile Properties:

Yield Strength (Ksi) - - - - 65.2

--  % Elongati6n (1") 29

~-----

Tensile Strength (Ksi) - - - 86.9 Reduction of Area (%) . 68.3

  • "Final Report: Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Properties" 1

.,j:I

),

  • l i i!

RECOMMENDED PROGRAM FOR IRRADIATION SURVEILLANCE OF PPLISADES REACTOR VESSEL MATERIALS Combustion Engineering, Inc.

Utility Division

  • Nuclear Power Department Windsor, Connecticut Prepared by

" Approved by

~-

Approved by i.

ne

\

\

I

  • J, I

)

~ .

  • I OBJECTIVE The irradiation surveillance program for the Palisades reactor vessel is designed to provide the capability of determining the radiation-induced changes in the mechanical and impact properties of the reactor vessel materials.

By conducting such determinations periodically, the continuing serviceability of the Palisades reactor vessel can be established.

_,I

~

- II INTRODUCTION Irradiation surveillance tests are conducted to obtain data on changes in the l nil-ductility transition temperature (NDTT) and the mechanical properties of

  • ! the reactor vessel materials. These data permit an evaluation of the integrity I! of the reactor vessel over its design lifetime. It 'is, therefore, essential l

that the capsules containing surveillance test specimens for monitoring the neutron-induced property changes in the ~eactor vessel materials be irradiated under conditions which represent the irradiation conditions of the reactor vessel as closely as practically possible.

ASTM E 185-66, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactor::;:., presents certain criteria which should be considered for conducting surveillance tests and outlines the procedures necessary for obtaining representative exposures and meaningful data. The irradiation surveillance program for the Palisades reactor vessel materials is based on this ASTM recommended practice and the recommendations presented therein are adhered to as closely as possible.

.1l

  • 1 4

"l 1

..-1

III TEST MATERIALS The intermediate and lower shell course regions of the reactor vessel are nearest to' the active core, and, therefore, will sustain the greatest neutron damage.

Surveillance test specimens will be fabricated from sections of the plates used to form the intermediate shell course of the reactor vessel.

The test materials will be representative of the materials in the fully fabri-cated reactor vessel. A complete record of chemical analyses, fabrication history and mechanical properties of the surveillance test materials will be maintained.

Three metallurgically different materials representative of the reactor vessel will be used for making test specimens. These include base metal, weld metal, and weld heat-affected zone materia1s. The procedures for preparation of the surveillance test materials from reactor vessel plate sections are established in CE NPD Specification No. 2966-165-1, "Process Specification for Preparation of Materials for Reactor Vessel Surveillance Testing."

A. BASE METAL (Figure 1)

Base metal test specimens will be fabricated from sections of the intermediate shell course plate which exhibits the highest unirradiated NDTT. The NDTT of each plate in the intermediate and lower shell courses will be determined from drop weight tests.

The material used for the base metal test specimens will be adjacent to the test material used for ASME Code Section III tests and will be at least one plate thickness away from any quenched edge. This material will be heat-treated to a condition which is representative of the final heat-treated condition of the base metal in the completed reactor vessel.

B. WELDED PLATES (Figure 2)

Weld heat-affected zone (HAZ) material' and weld metal will be produced by welding together two plate sections from the intermediate shell course of the vessel.

Material used for HAZ and weld metal test specimens will be at least one plate

. ~

thickness away from any quenched edge .

The procedures used for making the longitudinal butt welds in the reactor vessel will be followed, whenever applicable, in the preparation of the HAZ and weld metal test materials. These materials will be heat-treated to a condition which is representative of the final heat-treated condition of the welds in the.

completed reactor vessel. The procedures for inspection of the reactor vessel welds will be followed, whenever applicable, for inspection of the materials for HAZ and weld metal tests.

2 *

  • t t

II ll.

I A. TYPE IV TEST SPECIMENS The extent of the property changes in the reactor vessel materials can be de-termined by comparing the results of postirradiation tensile and impact tests to the similar pre*- irradiation tests. The changes in the NDTT of the vessel

... can be determined by adding the temperature change for a given Charpy impact energy value to the pre-irradiation NDTT determined from drop weight tests.

  • Tensile, Charpy impact, and drop weight test specimens (figures 3, 4, and 5) will be provided for conducting both the pre-irradiation and postirradiation tests. The procedures for fabrication of these test specimens from the test materials are established in CE NPD.Specification No. 2966-165-3, "Process Specification for Fabrication of Specimens for Reactor Vessel Surveillance Testing." All specimens will be marked to identify the parent materials, and their orientation and location within the parent materials. Drawings will be provided to show the specimen orientation and location with respect to the origional product form.
1. Base Metal Base metal test specimens will be cut from either side of the 1/4T (thickness) positions as shown in figures 6, 7, and 8. Longitudinal specimens will be oriented with the major axis of the specimens parallel to the principal rol-ling direction and to the surface of the plate. Transverse specimens will be oriented with the major axis of the specimens perpendicular to the principal rolling direction and parallel to the surface of the plate. The direction of the notch of the Charpy impact specimens will be perpendicular to the surface of the plate. The crack starter weld on the drop weight specimens will be applied to the surfaces adjacent to the l/4T position.
2. Weld Metal Charpy impact and tensile specimens will be cut from the deposited weld metal as shown in figures 9 and 10. Longitudinal specimens will be oriented with

.. the major a.~is of the specimen parallel to the direction of the weld and to the surface of the weld. Transverse specimens will be oriented with the major axis of the specimen perpendicular to the direction of the weld and parallel to the surface of the weld. The direction of the notches of the Charpy impact specimens will be perpendicular to the surface of the weld.

3. Heat-Affected Zone Charpy impact and tensile specimens will be cut from the weld heat-affected zone as shown in figures 11 and 12. The orientation of HAZ transverse speci-mens will be the same as the orientation of the weld metal transverse specimens.

The notch of the Charpy impact specimens will be parallel to the surface of the weld and the root of the notch will be centered as closely as possible on the weld/base metal fusion line. The F.AZ will be centered as closely as possible within the gage length of the tensile specimens.

3

B. UNIRRADIATED SPECIMENS The importance of establishing accurate mechanical and impact properties and NDT*temperatures for the unirradiated material cannot be overemphasized.

Taole I lists the quantity and type of specimens which will be provided for establishing the unirradiated properties of the reactor vessel materials.

Table I. Quantity of Specimens Provided for Establishing the

  • . Unirradiated Tensile and Impact Properties and the NDT Temperature of the Reactor Vessel ~..a.terials
  • Number of Specimens Material pact Tensile Drop Weight Base Metal Longitudinal 30* 18 16 Transverse 30 18 Weld Metal 30 18 Weld RAZ TOTAL:

30 120 18 72 16 *

1. Charpy Impact Specimens Thirty specimens each of base metal (longitudinal and transverse), weld metal, and heat-affected zone materials will be provided. This quantity exceeds the minimum recommended by ASTM E 185-66 for a Charpy transition curve and is intended to provide a sufficient number of specimens for establishing accurate Charpy transition temperatures for these materials.
2. Tensile Specimens Eighteen tensile specimens of the base metal, weld metal and heat-affected zone materials will be provided. This quantity also exceeds the minimum recommended by ASTM E 185-66 and is intended to provide a sufficient number of specimens for accurately establishing the tensile properties for these materials.
3. Drop Weight Suecimens The sixteen drop weight specimens will be provided for establishing the unir-radiated NDTT for the base metal at the 1/4 thickness position. These data can then be correlated with the wiirradiated Charpy transition temperatures for the
  • base metal in order to obtain an accurate basis for the determination of subsequent shifts in NDTT due to irradiation damage.

4

  • c.

Bo~h IRRADIATED SPECIMENS tensile and impact specimens will be used for determining changes in the mechanical and impact properties of the materials due to neutron irradiation.

A total of 480 Charpy and 90 tensile specimens will be provided.' Table II lists the number and type of specimens which will be provided for establishing the properties of the irradiated materials over the lifetime of the vessel.

  • . Table II. Quantity of Specimens Provided for Determining the Irradiated*

Mechanical and Impact Properties of Reactor Vessel Materials Number of Specimens Material Impact Tensile Base Metal Longitudinal 120 30 Transverse 72 Weld Metal 120 30 Weld HAZ 120 30 Standard Reference Material 48 TOTAL: 480 90 V SPECIMEN IRRADIATION

  • A. IRRADIATION CONDITIONS All test specimens should be irradiated at positions in the reactor that duplicate as closely*as possible the neutron flux, flux spectra, and temp-erature history experienced by the reactor vessel. Differences between the respective irradiation and temperature conditions to which the surveillance

.. specimens and the reactor vessel are exposed should be minimized .

B. IRRADIATION LOCATIONS

.. The majority of surveillance specimens will be irradiated at two radial positions about the midplane of the core. A third series of specimens will be exposed to the thermal environment in a low flux region above the core.

The exposure locations are presented in figure 13 and a summary of the specimens at each location is presented in table III .

  • 5
  • * - - - - - - - - - * - - . . , - - - - - * * - " * - * * * * - , . * - - -..- .~ ** ,..... _, ___ .',~',.i.,.__._.......,,_ .. l*.l."-L'" u.~-.....--_,.r. ....... _,Jo-*&......;...__::..._._,_..___ *<>*.*"*"*'"**-
  • TABLE III.

SUmmary of §pecimens Provided !2E_ ~ E?q>osure Location TYPE AND QUANTITY OF SAMPLES

                                      • 0 ****************** 0 *******************************************************

Base Metal Total Capsule Capsule Impact Heat-Affected Zone

...*.....*........ ............... . standard Impact Weld e * * * * * * * * * * * * * * * *

  • Designation Location i!!.2. (T) Tensile Impact Tensile Impact Tensile Impact ~ .{.!1 Tensile T-330 Thermal (a) 12 12 3 12 3 12 3 36 12 9 T-150 Thermal 12 12 3 12 3 12 3 36 12 9 Ch A-60 Accelerated(b) 12 3 12 3 12 3 12 li8 9 A-240 Accelerated 12 12 *3 12 3 12 3 36 12 9 W-290 Vessel(c) 12 12 3 12 3 i2 3 36 12 9 W-110 Vessel 12 3 12 3 12 3 12 48 9 W-100 Vessel 12 12 3 12 3 12 3 36 12 9 w-280 Vessel 12 12 3 12 3 12 3 36 12 9 w-260 vessel 12 3 12 3 12 3 12 48 9 w-80 vessel 12 120 72 30 3

120 12 30 3 12 120 30 3 12 48 48 420 72 90 9

e

  • * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *
  • Ci * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * ., * * * * * * *

(L) = Longitudinal (T) = Transverse Thermal Capsules Located above the core.

Accelerated Capsules - Located between thermal shield and core support barrel.

vessel Capsules - Located between thermal shield and reactor vessel.

  • 1. Accelerated Exposure One series of specimens will be placed on the core side of the thermal shield to' obtain an accelerated exposure. These specimens will receive the design lifetime fluence in a relatively short time and will provide advance information for predictions of the extent of the NDTT changes in the reactor vessel materials.
2. Vessel Wall A second series of specimens will.be placed on the vessel side of the thermal shield near the reactor vessel. The NDTT shifts resulting from the irradiation of these specimens will closely approximate the NDTT shifts of the vessel materials and will serve as a means of validating the data from the accelerated-exposure specimens ..
3. Thermal Control A third series of specimens will be placed in a low flux region above the core. These specimens will be exposed to all reactor temperature cycles but will receive an insignificant neutron exposure. Changes in the mechanical and impact properties of the vessel materials due to thermal exposure only can, therefore, be monitored by changes in the properti~s of these specimens.

C. IRRADIATION CAPSULES The func~ions of the irradiation capsule are (i) to facilitate the*removal of all test specimens from the cansule holder when the specified fluence levels have been attained, and (2) to prevent deterioration of the test ~pecimens .

during irradiation. In addition, the capsule design must eliminate excessive temperature differentials between the specimens and the reactor environment.

The capsule shape and size must minimize the number of flange penetrations; thermal, flux and hydraulic perturbations within the reactor; and temperature and flux variations among the specimens within the capsule.

The final capsule design should ensure that inclusions of such features as are required to obtain meaningful and comprehensive data from the irradiation program will not interfere with the overall operation of the reactor. In addition, the problems associated with postirradiation disassembly of the capsule and specimen removal should be minimized.

All of these basic requirements are adequately satisfied by the Palisades irradiation capsule.

1. Capsule Assembly A typical capsule assembly illustrated in figure 14 consists of a series of specimen compartments connected by wedge couplings, extension assemblies and a lock assembly. The wedge couplings serve as end caps for the specimen
  • 7

compartments and pos1tion the compartments within the capsule holders. The extension assemblies are provided to position the lock assemblies to a com-mon elevation. The lock assemblies locate the capsules within the holders exerting an axial fore~ on the wedge coupling assemblies positioning these assemblies against the sides of the holders. They also serve as a point of attachment for the tooling used for removal of the capsules from the reactor.

2. Specimen Compartments Each capsule assembly is made up of four Charpy impact compartments and three tensile-monitor compartments .

. a. Charpy Impact Specimen Compartments Each Charpy impact specimen compartrri.ent (figure 15) contains 12 impact specimens. This quantity of specimens will provide adequate data for the construction of Charpy transition curves for the respective materials. Com-parison of the unirradiated and irradiated Charpy transition curves will then permit the determination of the NDTT shift for the various materials.

The specimens are arranged vertically in four 1 x 3 arrays and are oriented with the notch toward the core. The temperature differential between the specimens and the reactor coolant is minimized by using spacers between the specimens and the compartment and by sealing the ~pt~§.§~~9ly in an at.mos~

~-hel'.~_.of be] irn..-A. stunmary of the type and quantities of the impact test specimens in each of the capsules is presented in table III.

b. Tensile-Monitor Compartments Each tensile-monitor compartment (figure 16) contains three tensile specimens, a set of threshold detectors for determinipg the flux spectrum, a set of threshold detectors for determining the flux atte:iuation through the specimen, and a set of temperature mo~itors for measuring the maximum temperature to which the specimens have been exposed. The entire tensile monitor compart-ment is sealed within an atmosphere of helium.

The tensile specimens are placed in a housing machined to fit the compartment.

Split spacers are placed around the gage length of the specimen to minimize the temperature differential between the specimen gage length and the coolant. A summary of the type and quantity of the tensile specimens in each of the capsules is also presented in table III.

In order to obtain precise predictions of vessel NDTT changes from postirradi-ation test data which are derived from specimens irradiated to different fluence levels and in different neutron energy spectra, accurate and com~lete in-formation on the neutron flux, neutron energy spectra, and irradiation temperatures of the surveillance specimens must be available.

8

  • Flux measurements will be obtained by insertion of fission threshold de-tectors into each capsule of surveillance specimens. Such detectors are p~rticularly suited for the proposed application because their effective threshold energies lie in the low Mev range.

Various threshold detectors, presented in table IV, have been selected to monitor the fast neutron spectrum (>1 Mev) incident on the reactor vessel.

These detectors were selected on the basis of a long half-life and an activation cross section covering the desired neutron energy range.

Table IV. Threshold Detectors Flux Spectrum Set Thre~hnlo Energy Material Reaction (Mev) Half-Life Uranium u238(n,f) 0.7 28y Sulfur s32(n,p)P32 2.9 14.3d Iron Fe54(n,p)Mn54 4.0 314.0d Nickel Ni 58 (n,p)Co58 5.0 71.0d 63 60 Copper Cu (n,CY)Co 7.0 5.3y Titanium T.46(

i n,p.)S c 46 8.0 84.0d The 5.3y Co formed by the Cu-63 (n,CY) reaction and the.28y Sr-90 formed by fast fission of U-238 will provide a means of automatically integrating the fast neutron flux. The Co-60 will provide integrated fast neutron flux for the first 10 to 15 years of plant operation while the Sr-90 will provide integrated flux for the entire life of the plant owing to its 28 Yf7ar half-life.

All the other reactions become saturated quickly and are only useful in de-terming the fast neutron spectrum over the desired energy range .

Two sets of flµx monitors will be used. One set of flux monitors, consisting of six different materials, will be used to determine the neutron spectrum.

Each material will be placed inside a stainless steel sheath which will be used for identification and to facilitate handling. Cadmium covers will be used for those materials which have competing thermal activities.

The second set of monitors is composed of iron wires in stainless steel sheaths and will be used to measure the flux attenuation through the thickness of the Charpy specimen. In order to determine the fast neutron attenuation at various positions in the compartments, the sheath will be used to identify the position of each wire .

The flux monitors for both sets will be placed in holes drilled in stainless steel housings as shown in figure 16. Both the flux spectrum set and the flux attenuation set of flux monitors will be placed at three axial locations in' each capsule assembly to provide an axial profile of the fluence level to which the specimens are exposed .

~ In addition to threshold detectors, the program will use correlation monitors made from a standard reference heat of A533 material which will be irradiated along with all the specimens of the vessel materials. The reference material has been obtained from the Heavy Section Steel Technology (HSST) Program.

The changes in impact properties of the reference material will provide a cross-check on the dosimetry in the Palisades surveillance program. In addition, these changes will provide data for correlating the results from this program with the results from 9ther experimental irradiations and reactor surveillance programs using specimens of the same reference material.

Because the changes in mechanical and impact properties or irradiated specimens are highly dependent on the irradiation temperature, it is necessary to have accurate knowledge of specimens as well as pressure vessel temperatures during irradiation. Instrumented capsules are not practical for a surveillance program extending over* the design lifetime of a power reactor, but the maximum temperature of the irradiated specimens can be monitored with reasonable accuracy by including in the specimen capsules small pieces of low melting point alloys or pure metals individually sealed in separate containers. The compositions of alloys with melting points in the operating range of this reactor which will be used in this program are listed in table V.

The temperature monitors consist of a helix of low melting alloy wire inside a sealed quartz tube. A stainless steel weight is provided to destroy the integrity of the wire when the melting point is reached. The melting temper-ature of the temperature monitor will be indicate1 by the physical length of the monitor.

The temperature moni~ors will be placed in holes drilled in stainless steel housing; as shown in figure 16. The temperature monitors will also be placed at three axial locations in each capsule assembly to provide an axial profile of the maximum temperature to which the specL~ens .were exposed.

Table V. Composition and Melting Points of Temperature Monitor Materials Comnosition Melting Temperature

(~t %) (OF) 92,5 Pb, 50.0 Sn, 2.5 Ag 536 90.0 Pb, 5.0 Sn, 5.0 Ag 558.

97. 5 Pb, 2. 5 Ag 580 97,5 Pb, 0.75 Sn, 1.75 Ag 590 10

=--*

.(

  • ~ .
  • f:, W * .Hewitt K~viaed Pa lh~dea Snrvcil lancc-Capsule J.emoval Schedul~

D. W, Stephen P-PH-232 i

I*

February 20, 1970

  • The SUl'Vcillance caosule removal achedul~ a(:l shown in Table 4-17 I

1"< the Pa !iQaccs FSA '1 ha.a been l"eviscd to l'eilect the the,.ma l shield t"C- I rnoval a.:id a Wfi.ll capsule rem.oval with a fluence of approximately 1.xl 019 Thie t'.;ovh:ed schedulu is as shown in Tnble l. The {luence levels given.

n/cm~

.I I a.,.e nominal valuea. The ap;:>?"cximafo refueling schedule is based on reactor o:H~r1'.tion o! Z540 mwt and a O. 8 load !attar.

I

.I I

I Surveillance C?.psule aernova.l Schedule b

I Iutegr~ted Approximate Capsule Target

  • Pow.:-1° ~MW DJ Refueling .Removed Flueuce (nlcm2)
  • l.73 x 106.
  • 3. 97 x 1 o'1 3.9?xlo6 z

5 5

A-6Q A-240 W-Z90

1. 1 x io-19 3.S.x.Iol9
3. 7 x 10 1 3 II I

I I

3.<n x io 6

  • 5 T~330 R. 44 x 106 11 W-!10 9.3xlo 18
1. 51 x 107 zo W-100 I.7.x:lol9 l, 51 x ioi zo T-150 I. EVi x io 7 ZS W-280 z. 1 x 10 1q Z.63 x 107 35 W-260 3.0xlo 1 9 3.0ci.:c 10 7 4-0 W-80 3. 4 x io 1 9

_DWS:*-'?

Dibti-ibl1tion:

A. J. .Blatter V. C. H<'.11

{, J. h :otz

. jJ. J. Koziol S. Visoer

  • .?v:. F. V3 lerino

':l. E. Wolf

D. CAPSULE REMOVAL SCHEDULE All surveillance capsules will be inserted into their designated holders dilring the final reactor assembly operation. The capsules will remain in the reactor until the desired fluence.level has been attained by the speci-mens. Table VI presents the target fluence for each of the capsules and a tentative schedule for removal .

Table VI. Capsule Removal Schedule Target Refueling Capsules Fluence Schedule Removed (n/cm2) 1 A-60 7.55 x 1018 4 A-240 2.46 x 10 19 4 W-290 1.48 x 1018 4 T-330 18 18 W-110 6.27 x 10 20 18 W-100 8.21 x 10 20 T-150 25 W-280 1.02 x 10 19 35 W-260 1.42 x 1019 40 W-80 1.63 x 1019 The first refueling is assumed to occur after 16 months of operation and each succeeding refuelling occurs every 12 months. The calculational uncertainty for all fluence values in this table is +20 percent. Additionally, there can be a 30 percent variation in fluence-over the length of the capsule due to variations in the axial power.

11

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  • /

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View*

'1 I Figure 13 I

Enlarged Plan View .I

....... --~- ---- LOCATION OF SURVEILLANCE CAPSULE ASSEMBLIES

1--- Lock Assembly Extension Assembly Tensile -Monitor----..1....1 Compartment Charpy Impact Compartments Tensile - Monitor Compartment Charpy Impact Compartments Figure 14 TYPICAL SURVEILLANCE Tensile -Monitor-----1 Compartment CAPSULE ASSEMBLY

~

Charpy Impact Specimens

  • i Rectangular Tubing Q

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  • Wedge Coupling - End Cap Figure 15 TYPICAL CHARPY IMPACT COMPARTMENT ASSEMBLY

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Split Spacer 1 I

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Q"

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  • TYPICAL TENSILE - MONITOR COMPARTMENT ASSEMBLY
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