Text

Response to NRC 800825 & 0910 Ltrs Re TMI-related Questions on RCS Vents.Supporting Oversize Drawings Encl
	ML19345A202
Person / Time
Site:	Farley
Issue date:	09/22/1980
From:	; ALABAMA POWER CO.
To:	;
Shared Package
ML19338E087	List: ML19338E086; ML19345A202;
References
	NUDOCS 8009240522
	Download: ML19345A202 (165)
	v • d • e

{{#Wiki_filter:r-C =1. TMI RELATED QUESTION ON SECTION II.B.1, REACTOR COOLANT SYSTEM VENTS REQUEST a. Provide your design of the pressurizer vent system per our November 9,1979, letter. Assure that all requirements and recommendations of that letter are satisfied. b. If a _ break in the vent line is beyond that defined as the smallest LOCA by 10 CFR 50, Appendix A, provide an ECCS performance analysis for a complete spectrum of breaks in the vent line. Assure that the thermal-hydraulic modeling is consistent with break location. c. Demonstrate that the exhaust from the vent. system will not impinge on other equipment, that the vent system will vent RCS hot legs, and that the vent exhaust is to a portion of the containment with maximum ventilation and cooling. d. Demonstrate that the vent system is qualified to RPS safety grade standards. Include seismic design, IEEE-279 requirements, minimize inadvertent actuations, vent valve position indication in the control room, qualification to pass non-condensibles, steam, water and combinations, thereof, etc. e. Provide procedural guidelines and analytical bases (preferably, generically developed by owners group) for vent operation and termination as related to plant performance. The procedures thould be based on the following criteria: p (J) the plant must meet the requirements of 10 CFR 50.46 and 10 CFR 50.44 for DBA's; and (2) the plants ability to maintain core cooling and containment integrity for events beyond DBA's must be increased. Procedures should also address methods to (1) assure natural circulation through the U-tube portions of the steam generator with the potential accumulation of gases in this region, and (2) assure that combustible limits are not exceeded.

RESPONSE

a. The ability to vent the pressurizer is provided by two installed Power Operated Relief Valves (PORV's) which vent to the pressurizer relief tank. Their associated circuits meet the separation requirements for redundant systems with one PORV Powered from train "A" and the other powered from train "B". Each PORV is air-operated and equipped with two solenoid-valves in their respective air lines. The solenoids are powered from train A and B 125Vdc buses, respectively. These buses are powered through associated battery chargers which can also be supplied from the train A or B diesel upon' loss of offsite power or from 125Vdc batteries. A system will be installed as a backup system to the instrument air system for operating the~PORV's. Each PORV has a motor-operated block valve located in its respective piping. These valves' associated circuits meet the separation requirements for redundant systems. The operator and its associated control circuits for -each block valve are. powered from its associated train A or B emergency bus. A (

The motive and control power interfaces (for the PORV's and associated block valves) with the emergency buses are ace.omplished through safety grade devices. These PORV's are operated from the main control room and have positive position indication provided to the operator via stem-mounted limit switches. The PORV's are seismically qualified for static loads. Alabama Power Company is participating in the EPRI Relief and Safety Valve Test Program and will upgrade the PORV's as necessary following the com-pletion of the test program and plant specific analysis, b. The Reactor Vessel Head Vent System (RVHVS) is orificed to limit the blowdown from a break or inadvertently opened flow control valve downstream of the orifice to the capacity of one charging pump. A postulated break of the reactor vessel head vent line upstream of the orifice would result in a small LOCA of not greater than one inch diameter. Similarly, for pressurizer venting using the PORV's a postulated break would result in a small LOCA of approximately 2.5 inches diameter. Such breaks have been analyzed in WCAP-9600, Sections 3.2 and 3.3. This WCAP contains extensive discussions regarding the applicability of the break flow models employed as well as other specific modeling features employed for small break LOCA analysis. The analysis in WCAP-9600 provides ECCS performance analysis and shows that no calculated core uncovery would occur for postulated breaks of sizes corresponding to the RVHVS and PORV vent systems. c. The RVHVS will discharge into an open area between the reactor and the CRDM platform. As shown by the attached drawings, this discharge is directed towards the reactor cavity walls and does not impinge on any safety-related equipment. The discharge rate for hydrogen is 5600 ft3 per minute. Dis-charging into the reactor cavity provides sufficient space to prevent any gas collecting. Ventilation is provided by the CRDM cooling fans to ensure optimum dilution of combustible gases. The exhaust from the pressurizer vent system is directed to the pressurizer relief tank (PRT) which prevents impingement on other safety-related equip-ment. The post-LOCA mixing fans provide ventilation for the entire con-tainment in an accident situation. Three sets of fans will provide sufficient ventilation in case of a PRT disc rupture. For a Westinghouse-designed plant, the RCS hot legs are not a high point in the system..Therefore venting of the hot legs is not required. j

d. :The Reactor Vessel Head Vent System ~(RVHVS) provides for venting the reactor vessel and head to remove noncondensible gases or steam using only safety-grade equipment.

The.RVHVS consists of two paralled flow paths with redundant ~ isolation valves in each flow path. A schematic diagram of the RVHVS is shown in Figure 1. Th6 solenoid operated valves in each flow path are powered by opposite vital buses. If one single failure prevents a venting operation through one flow path, the second flow path 'provides a redundant backup. 1

( The two isolation valves in each flow path provide a single failure method of isolating the venting system. Each isolation valve is a fail closed, nonnally closed, active valve. With two valves in series, the failure of any one valve or power supply will not inadvertently open a vent path. Thus, the combination of safety-grade train assignments and valve failure modes will not prevent vessel head venting or prevent venting isolation with any single active failure. The RVHVS has two nonnally deenergized valves in series in each flow path. The two series valve arrangements eliminate the possibility of a spuriously opened flow path due to the spurious movement of one valve. As such, power lockout to any valve is not considered necessary. The system is operated from the control room. The isolation valves have stem position switches which are monitored in the control room by status lights. The RVHVS is seismically designed and supported for seismic, thermal, pressure, deadweight, and dynamic loads resulting from the flow of gases, steam, water, or two-phase flow. The isolation valves will be qualified to IEEE-323-1974, IEEE-382-1972, and meet Regulatory Guide 1.48 (active valves). In addition, this system meets IEEE-279-1971 requirements. e. The Westinghouse Owner's Group has approved a program to develop appropriate instruction regarding the conditions under which reactor vessel head vent operations should be conducted, and the manner in which head vent operations would be made. This program is scheduled to be compelted prior to the end [ of this year so as'to be implemented in plant specific procedures to make the reactor vent systems operational by January 1,1981. It should be noted that the Farley Nuclear Plant already meets the require-ments of 10 CFR 50.46 and 10 CFR 50.44. The addition of the vent systems and its operation will not compromise this ability. The instructions developed as part of the above referenced Westinghouse Owner's Group effort will be developed specifically to enhance the safe recovery of the plant for events beyond the design base to assure that combustible gas limits are not exceeded and also to assure natural circulation through the U-tube portions of the steam generators. The procedural guidelines developed by the Westinghouse Owner's Group will be incorporated into the Farley procedures. 4 .-..U .w. m_

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a g l'J L SU,E_STION 122.10: The containment pressure boundary is constructed using materials meeting the requirements of ASME Section III or various B 31.X piping codes; however, our licensing reviews, including that for Farley-2 do not include evaluation of explicit compliance with these codes. Consequently, to complete our final SER for Farley in response to the July 18, 1980 memorandum, D. G. Eisenhut to R. H. Vollmer, we require the following information: 1. Identification of the fabrication codes (edition and addenda) and the specific paragraphs in these codes that specify the fracture toughness requirements and acceptance criteria (for weldments and base metals). Codes and code paragraphs should be identified for all materials which constitute part of the containment pressure boundary (e.g., piping penetrations, personnel airlocks, equipment hatch). 2. The materials test data that certify that the fracture toughness acceptance standards have been met for each of the identified materials in the containment pressure boundary. RE_Sp0NSE: 'O l'.' The Farley-Unit 2 containment penetrations are identified on drawing ( D-206157 (enclosed). Also enclosed are drawings D-206151, D-206152 and D-207411 which provide detailed information concerning the contain-t ment penetrations. A documentation package is enclosed which provides: i (a) a penetration summary that accounts for all penetrations by ~ number and explains which numbers required fracture toughness tes ts; and, (b) a table of general requirements for containment penetrations which delineates the fabrication codes and other applicable requirements. The table separates penetrations into three categories determined generally by the design specification which governed fabrication. 4 2. The enclosed documentation package also includes material test data which demonstrates that the fracture toughness acceptance standards required by code have been met for the containment pressure boundary ma terials. All carbon steel penetration materials which serve as containment pressure boundaries, and which exceed 1/2 inch thickness, have been satisfactorily impact tested. ~ Exclusion of carbon steel ! n 1/2 inch thick or less and stainless steel is in conformance with V - ASME Section III, Paragraph NE 2300. In the case of forged carbon i steel flued heads, successful impact tests were performed for all penetrations. e

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to 1: INDEA l. Penetration Summary 2.- General. Requirements for Containment Pcnetrations 3. Mechanical Penetration Listing i a. Sleeve mill test reports b. Cap mill test reports J. c. Flued head mill test reports d. Penetration No. ' 84 Equipment Hatch e. Code reports 1. Penetration #86* Personnel Hatch {) 2. Penetration #87* Auxiliary Hatch t 3. Penetration #84 - Equipment Hatch I 4. Electrical Penetration Listing 4 a. Sleeve, cap and adapter mill. test reports I r 5. Drawings j' I 4x )

MTRs on file at Chicago Bridge and Iron I

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V O I PENETRATION 'JIf tMARY l PENETRATION NUMBERS (Drawing D-206157) COMilENTS I thru 9, 51, 52, 53 Carbon steel flued heads. Material rest Reports attached. 15 thru 18,-23, 28

'inless steel flued heads.

Fracture 56, 57, 58 ' toughness test not required. MTR's a ttached. 10,,11, 92, 94 Stainless steel containment sump ' () lines. Fracture toughness test j not required. 12, 13, 14, 19, 20, 24 Carbon steel penetration sleeves and 25, 26, 27, 29, 30, 31, nozzles (pipe caps) with wall thick-33, 43, 47, 48, 49, 50, nesses of 1" or less. Fracture . 54, 55, 59, 61, 62, 63, toughness tests not required by 64, 66,.67, 68, 70, 73 Paragraph NE2300 of the Code thru 82, 85, 88 21, 22, 32, 34 thru 42, Carbon steel penetrations with at 44, 45, 46, 60, 65, 69, least some material greater than }" 71._72, 83, 84, 86, 87, thick: Sleeves, nozzles, (pipe caps), 89 thru 92, 95 thru 104 equipment hatch, personnel access lock and auxiliary access lock. Material test reports showing accept-able fracture toughness tests are included. l^ h u-Lj

1 O O ~ 1 O GENERAL REQUIREMENTS FOR CONTAINMENT PENETRATIONS FRACTURE FABRICATION TOUGHNESS TEST ACCEPTANCE ' PENETRATION CATEGORIES NATERIAL~ SPECIFICATION CODES REQUIREMENTS TEMPERATURE CRIIERIA COMENTS Flued Head SA105 (as modified by ASME BPVC ASME BPVC 30 F. AFa5 BPVC. Only' carbon steel ~ Section III Section III-Section III flued heads are codeCase1519}$$$$~. !N!!_M.. _ 1971 Edition Para. NC2300 Appendix I-considered since Table I-7.1 fracture toughness SA182 F304 Susser 1972 tests are not Addenda Subsection NC required for stain-less steel. ASME BPVC - Elsctrical. G.E. Section III ASME BPVC 0F ASME.BPVC G.E. penetrations did PInetrctions SA182 - F304 1971 Editiou Section III Section III not exceed %" thick-SA240 - 304 Summer 1972 Para. NE2300 Appendix I ness for carbon SA105 - GR11 Addenda and Table I-7.1 steel. SA312 - 304 1977 Edition Subsection NE 4 - Panatration Sleeses, Seamless Pipe ASME BPVC ASME BPVC -30 F for ASME BPVC !M33 _ Grade,6,, Section III Section III Equipment Section III N:nzles (Capa), a Equipment Hatch, Fittings Subsection NE Para. NE2300 Hatch. Appendix I - Parconnel Lock and. gg34, Grade,g B 1971 Edition O'F for Table I-7.1 3 . Auxilicry Access Lock Forgings Summer 1971 other items E#q}0,,(,gade,g Addenda Steel Plate SA516, Grade 70, e 4 A e-

T HECHANICAL PENETRATIONS MILL HEAT NUMBERS NETRATION SLEEVE CAP FLUEC HEAD 1 80lJ25770 (A) 4-2676 2 80lJ25770 (A) 4-2684 3 801J25770 (A) 4-2708 4 802J13199 (A 1-3440 5 802J13199 (A 1-3486 6 802J13199 1-3504 7 113379 4-2532 8 113379 (A 4-2586 9 313379 4-2586 NA 10 .itainless Steel 11 Stainless Steel (A) NA 12 80lJ25820(D) A2225 NA 13 80lJ25820 A2225 NA 14 80lJ25320 (B) NA 15 802J15000 (J10713-2) A) 1-3632(D 16 802J15000 (J10713-1) A) 1-3591 (D 17 802J13460 (A) 1-3668 0 18 80lJ28569 (A) 1-4674 D) 19 113379 645J483 iM 20 113379 645J483 NA 21 62320 FT4HH NA 22 62320 IT4HH NA ('_') 23 113379 (C) 1-3629 (D) 24 113379 645J483 NA 25 224337 94624-26 NA 26 224337 94624-26 tM 2/ 224337 94624-26 NA 28 113379 (C) 1-3629 (D) 29 224337 C-E(D) fM 30 224337 94624-26 NA 31 224337 94624-26 NA 32 121035 IT4HJ1 NA 33 223610 GA4BH (D) NA 34 42794 IT4E NA 35 60217 IT4m IM 36 42794 FT4E NA 37 42794 FT4m IM NA 38 62320 IT4HH1 E 39 62320 ET4HH1 NA 40 62320 ET4HH1 NA 41 62320 FT4HH1 00 42 A45256 FT4HJ1 NA 43 224337 94624-26 44 A45256 FT4HJ1 45 A45256 FT4HJ1 g 46 A45256 FT4HJ1 N4 47 223610 GA4BH (D) ('# g 48 223610 CA4BH g4 49 213748 CA4BH gg 50 213748 GA4BH(D) pg 51 224337 4-2532 52 2243:7 2L-2574 t

e SLEEVE, CAP FLUED HEAD PENETRATION 53 224337 (A) 4-2532 54' 213748 GA4BH (D) ~ NA 55 213748 GA4BH (D) NA 56 H3379 (A) 1-3591 (D) 57 H3379 (A) 1-3090 (D) 58 113379 (A) 1-3632 (D) 59 224337-94624 -26 NA 60 62320 M4HH1 NA 61 H3379 645J483 NA 62 213748 GA4BH(D) NA. 63 213748 CA4BH (D) NA 64 224337 FN4WH(D) NA-NA 65 802J13460 FT4JHL 66 '213748 GA4BH(D) NA 67 213748 GA4BH (D) NA NA 68 64P148 GA4Bd (D) NA 69 60277 FI4KH

E 70 223610 CA4BH (D)

M 71 121035 FT4HJ1 NA 72 12'1035 FT4HJ1 M .73 213748 GA4BH (D) (D)) (D 74 213748 CA4BH 75 213748 GA4BH gp ^ 76 224337 94624 -26 f 77 224337 94620-26 g 78 224337 94624 -26 g O-79 224337 94620-26 NA 80 113379 645J483 g 81 H3379 645J483 g 82 113379 645J483 g 83 121035 FT4HJ1 g 84 Equipment Hatch (See Attached) g 85

Elae"4 e =1 Penetrations (See Attached) g 86 802K78860 80lK08329 (D)

Personnel Lock gg 87 80lK08329 802K78860lD') Auxiliary Access Lock NA 88 Electrical Ground (B) NA 89 A45256 FT4HJ1 NA 90 A45256 FT4HJ1 N4 91 A45256 FI4HJ1 NA 92 A45256 FI4HJ1 NA 93 Stainless Steel (C) NA 94 Stainless Steel (C) NA 95 60277 FI4KH NA 96 42794 FI4KH NA 97 802J13460 FI4JH1 NA 98 802J13460 FI4JH1 NA 99 A45256 FT4HJ1 NA 100 A45256 FI4HJ1 NA 101 802J13460 ET4JH1 NA 802J13460 FI4JH1 NA O 102 103 145256 rr4ua1 "^ 104 A45256 FT4HJ1 NA w_

g 7 O-NOTES ~ (A) Cap not required'with flued head. (B) Carbon steel material less than 1/2" thick. (C) Stainless steel material. t i (D) Charpy impact test not noted and not required because of material thickness or because material is stainless i -- steel. i 4 e i i O ~ l ]_ 4 i~ ? ) +h t i- { 4._ I-l ' I i y y g-7 w -- v-..ar,. .-,,em, y.-

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jt M LEH bRPORATION,:..;.'.,. :., strwIs2sh t-D - f w vinrusa............ :.:. REPORT O 'AND ANALYSIS W sin smns g f cea os viruca No.. u we sets no. RURNS HARBOR 803-21023 12-24-72 PC LOUISVILLE 1.N PC 577403 PAGE ~ CHICAGO BRIDGE C IRON CO CHICAGO BRIDGE C' IRON CO. BOYLES.Ah-I BOX 277- ,5 BIRMINGHAM AL 3S202-E .- %,.: c. .. a. - s. noas i.f,2, ", l"u ""'e'fs',7j' e'E'No**" 5" '**'"' veo vtusu sr= war-
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ly 0:.9, /.C 65-2 SHEET 43' GH O25-2179A -CHARgY V NOTCH OF 20 FT LB' AT HINUS 30 DEG.F PER NB-2300,- N ! 7 k E .c h 1/ 2 0l7 0' o c0 % 68 13 .,3a72 ~ 51 J 33200-01801J25770 J 33200-02801J25770 1: 1-1/ 2-68 134 3872T 51 ' 74 OK ^ ~ ~ l. 510p / '0 8"3)-(. EEE I i i i i-i J 33200-03801J25770' l' 7 0 OK 1/ 2 68 1 3872 i l-i cgg [ 11 c==m I i i i i l c=-=s 2 555 I l l l l l i. i i ',,31., ace t e I I l g-m_i u. o!. he specif ation numbers shoten here have been cot. PLATE N0RMALIZED AND STAMPED MT 7 M 9 - d o d # curxcit uutvsis A / p'. o ct - . ui c, ,j 5 7 3, f c, u; e, me y n 01J25770 .2d 1.63 . rA5 074 /25 p SUDSCRITED h.:D SWO3M TO ;EFORE ME 20!TR5770 N nn3 h DAY CF,.. 1973 f. ^- .N u.1:/ i UOL1C f ' ( ~ PORTF.it COUNTY INDIANA MY COMpg1SSION EXPilt!;iSg MAY 10, 1975 ,. %Id W
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c.rm q'e ..=.s I se sifg the above seivtts to be correct es contained in the eecords of the Bethlehem Steel Corporollon. .e
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to Dy o. INPACT PROPERTIES CUSTOMER: Chicago Bridge and Iron Company PAGE 3 ATTACHMENT hilIPl!ENT NO. 803-21023 D. ATE SH.I.P. PED12/24/72 LONGITUDINAL CHARPY V-NOTCH. TESTED @ C-F SERIAL NUMBER HEAT NUMBER CHARPY SIZE FT. LBS.,- % DUCTILE FRACTURE AREA LATERAL EXPANSION (TN) I 1 34-2 .028.0,3) 038 3 J33200 (1-3) hJ257h Full .<32 t I I m C3 p F~ 3 J w 3 W %S W p ,/' r, b 3 E 'i son ~ose w e n g g_ g y
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U n L. ty,,. - -( r,. g .7 v y RECORD OF HEAT TREATMENT l PAGE 3 ATTACIDfENT CUSTOMERt Chicago Bridge and Iron' Company. , ;,..,,. !p SHIPMENT NO. 803-21023 DATE SHIPPED 12/24/72 '.4 ~_.. \\ PLATENORMALIZINGCYCLE;2f -::4'. ; FURNACE TIME s n N CdOLING ~ HEAT NUMBER TEMP OF ' (MIN) Q^ SERIAL NUMBER QO1J25770 V,,- 1645/1655 ~ 51: .;=
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' &)/ AliCoo'Q_ J33200 (1-3) %y,. Q 93 Cf. .= q ? II. 9 C-W FM JEE3 e s u (. [, i((f,h.'),h, ',- 4 ? ...: t,. ' * : ' _q. PLATES AND TEST SPECIMENS NORMALIZED PER PROCEDURE PM300. TEST SPECIMENS iNOT REMOVED U1TiL AFTER PLATE N0y I ZED. ',9 +.. ?ob-o-o oorg;l f y n.s 1 ,7 p q..;, Q
mIp7((So r.* '~ / BETHLEH CORPORATION .a*- Q (\\ urr . t otrAntutur - .q, ~ (e s REPORT O .fS AND ANAL.YSIS s cAtt vt.no cat ci vuncu wo. g s giur saiuwt so. BURNS HARROR 803-21023 12-24-72 PC LOUISVILLE LN PC 577403 P A'GE ,~ CHICAGO BRIDGE C IRON CD. CH'ICAGO BRIDGE C IRON CO S B O Y L E S A I,.f., -) BOX.277 l ,y s BIRMINGHAM At'35202.;- c.. h .... R.-i - ~ ..,.p -;&c. "T,,',L'.o.=/=Ja sra - mm um ,,~,o ,co ,. i i w i,..,.. i m,. i... im c=
== cra -er. s y ,. i, ~~a ~~a T EEL PL AT E S-MS-6 52 REV. O & QAS 318 REV O ASME S A516 GR.70 PVQ & ASt4E SEC.T 3 ,.::7 4 'y .l COS2065'-2 SHEET 43 C'! 025-2279 J 3047-1.C80lJ25820 1.. 6 1/ i. 87_. s. , 783 5160 -0 8-5160 i 76'O.Or,8.? K ~ Q J133047-ID801J25820.' I' ' I/ 52 16 1245; E 3 J133047-1 E801J25 820.' l' 1/ 2[ 52 169 124d 5160p7'6D 0 8 2.6dK/ we) i i ? ~ - 1 3 28) l l l l s e) I i i. i e M / l i I i N l l i i m .u,. u_ _yu..__ _._ ho specification nu::)bers shown hereon bavo been cot., PLATES NORI%LIZED $NJVSTAMPED MT ooca-gsoovy p/f.o/ / / / / curarcat ex^tysts a c/ u./ r/ s / u /c. uvunta n; cr u. v 1; 901J25820.. v25 1.p9 .010. 02f .26/ j.. SUBSCEI.'ED P.I:D S~.7"'I TO EETORE ME 5 TIII3 .D.~.Y O. Q._ 1973 Q; 7 .~ l'A1:f 1 U1:1.!c y ". P.^.::TP.R C INDIANA l ur conypstoN 1:xr:n.rg')UNTY n.w no, im 4
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i-RECORD OF HEAT TREATMENT \\ t 1 '- h J PAdE.E. TTACIDINNT . ;[,',p,M.gh., yh.fy.,,gs.; SHIPMENT NO 803-21023 CUSTOMER: Chicago Bridge and Iron Company ' 'N.J y, de ...*fS..! DATE SHIPPED 12/24/72 ' ' ~ ~' PLATE NORMALI2INO CYCLE .y, FURNACE. TIME 18f/ SERIAL NUMBER HEAT NUMBER TEMP OP
(HIN)_
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RECORD OF HEAT TREATMENT h CUSTOMER: Chicago Bridge and Iron Company $i *:r: PAGE. 3 ATTACIDfENT 'l 'I SHIPMENT NO. 8031023 4 DATE SHIPPED 1/05'/.7) M ik m . PLATE NORMALIZING CYCLE q Q '_ SERIAL NUMBER HEAT NUMBER TEMP OF (MIN) ' g' FURNACE T HIE i C00$ING h U o ,.gl@ Aik:CooG_. J10750 (1) 8569 1650/1670 / ~ 4 0 -- p 4 ) e o e PLATES AND TEST SPECIMENS NORMA ZED < PROCEDURE PM300. TESTSPECIMENSNOTREMOVEDUNTIL,{.fTERPLATENORMALIZED. 9 on-o- soor.r p g, 7s .Q M o/ ...Q
~ A-IMPACT PROPERTIES Si CUSTGMR: Chicago Bridge and Iron Company V PAGE 3 ATTACIDENT 803-02314;;- '4 SHIP)ENT No. 1/05/73%.(h DATE SHIPPED . M' ,e .x LONGITUDINAL'CHARPY V-NOTCH TESTED G 5 ,Q g SERIAL NUMBER HEAT NUMBER CHARPY SIZE FT. L'ES. % DUCTILE FRACTURE AREA LATERAL ANSION (I J10750 (1) 801J28569 Full ~ 4[6 30- M 3 .040be p r4 6 9 ass 6 ) > red C (5E) i W g .: +;. q hff 'fhhh;j$:'4;. .at - 3.' ;t'.'.O.... f4 ' C, kly ,f ,h' $;I .][I 7M ndst$ g iodofYj l__
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.;..Nk. nownon .m .mmo. v / REPORT OF T .ND ANALYSIS -q CA1 C3 VEMick! HO. DATE SMWFLD PAG / 4 sawmW4 HO. .4 06-3d-73 PC l_CUTSVTELE LN PRR 4RO367 ,, BtJR N S H A RBOR 803 54917 ~ CHICAGO BRIDGE C' IRON CD. CHICAGO BRIDGE C IRON CD' * ' ' E BOYLES AtJ.. s.~' BCX 277 S' BIRMINGHAM AL 352C2... 5 ~,.. . 4, ..c.. ,Ja, nusu aO,e. =^=er t Su a ovemrr Scg' s',gt,'gy,, ^* samt mir $== mmm n5 1,,. I v.i I wu==. wuam
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FLATE NORFALIZIED AND STMPED MT t -)OcM 6. CO WY s-A1/t.oc/ 6 enunen ntws / / /- / W >n c, = v. ti A / si wu.9:s cj u/ s 31 Kj]_8_321).fi 1.JS.%.p .[ SUBSCRIDED AND WORN..-TO.BEFORE Mr. ] gg g. T1H3f'2DAYcy.- 7 :, 19 3 f. ff/s, % M:M M %. Q My coststigston-EXPIRES,;. GAY, 30,1975 ,f. cs.. . w: 4 - -., mp ... c....
l.. C,.. .Q,;;.-:.. '. . ', _ ~ ', ' ( h. l.:... ', ......,...,;. m s' .,I, p ..a 7 RECORD OF HEAT TRE$1. MENT ~' ,,.1 ... :\\ ~ .r ~.. ~
2. I PAGE' 4' ATTACHMENT.
' l, i CUSIOMEF.: Chica'go'. Bridge and Iron Company SHIPMENT NO. 803-14917 .,V 1 ... I) ATE SHIPPED 6./30/7.3.1 a . c. .~ .l. 9.. ,.e. ..s. .,.,. s a ... J. *- .s'" PLATE NORMALIZING' CYCLE g '. FURNACE. .~ TIME r.... ' - - M,.,S. - ] 'H TEMP OF j-e' _(HIN) ;.: .; COOLING m SERI.AL NUMBER'. NE'AT NUMBER TET6.. U ~ l .) K 10547-01 ?' ~*
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chornhereonhajv ioencot. A // o / PLATES NORMAll ZED AND STAMPED MT f entment uutms e j,, /y / s s, . c. u; e, y ,s nunun 0_2Jl5000 .2F L,. O V yO O 6'.01Y f. 2 3' / 3 02J131p. 23/ .99/.006/.023/ '24f j,, SUE 007.II pITI G't.*f'91 To sgroar y.r; j 907'It' alt 1 Ii' ? 'f E~ E I. 5:~. Tid:i.i' ~ c '...:RT n: :T*J.s:77 3xmaxA MY COMMig-10N r.';;In2.$ YAY 10,1075 r w,. Guv {4 4

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I' M E.INAy g..,. g -O D O. ~ IMPACT PROPERTIES CUSTOMER: Chicago Bridge and Iron Comp'eny
PAGE 7 ATTAC10ENT SHIPMENT NO.
803-21023,' i DATE SHIPPED 12/24/72..'. .g/ LONGITUDINAL CHARPY V-NOTCH TESTED @ .f 57 e Q m F5UCTILE FRACTURE AREA LATERAL EXfANSIO (IN) SERIAL NUMBER HEAT NUMBER CHARPY SIZE PT. LBS. J10713 (2) 802J15_000 'Jull -29 f 44-33 3 .05 45-J 5 3p'839 65-2 .03 p'.055 2 f J10710 (1-3) (80'2J13199) Full m I il l
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a spag.,4. ;+% [. .y.7 y. RECORD OF HEAT TREATMENT [S PAC'E7 IiTACRIENT CUSTOMER: Chicago Bridge and Iron Company 1'i SilIPMENT No. 803-21023 DATE SHIPPED 12/24/72 ' 7
.W;:',Q 4
PLA'TE KbRMAi,IZI'NG CYCLE e .7, FURNACE TIME bLIN SERIAL NUMBER IIEAT NUMBER. TEMP OF -,{MJpQ J10713 (2) 80"J15000 1650 V 36 ' A13 Cok O' J10710 (1-3) (802J13T9', 1650/1665 / 29 Cool 11 il l .?.))B,(N'..~,i:..o?- T st.t I PLATES AND TEST SPECIMENS NORMA ZED PER PROCEDURE PM300. ,d. TEST SPECIMENS NOT REMOVED VNT AFTE LATE NORMALIZED. 's h QOV9- $ 66;eap) n,,y,
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srm e4. red .,l.,., /. I BETH CORPORATION '.'y ".,.. ,,-(..' c rAuen y / REPORT lesTS AND ANALYSIS t A ru : 5,.,-im a. r oats s,.ero ca ca vuncu wo. q RURNS HARBOR 803-21023 12-24-72 PC LOUTSVILLE LN PC 577403 PAGE I i CHICAGO BRIDGE C IRON CD CHICAGO BRIDGE'C'IROW'C0
;n BOYLES AL.-
BOX 277- ~ j y BIRMINGHAM AL.35202' 3 1 e. saat near vm tmsu sud' 1 sus a cuumtv 8 h "$' ',Co',5',',$ ','.^o' *" umu wwa simcm sincm - 9,g isco e,ou wo ,,,,,,, l g ww., ox .l t,,,,. l w,,,,, ,i STEEL PL ATES-HS-652 REV O 6 QAS 318; REV, O ASHE i SAS16 GR; 70 PVQ C jASHE SECT 3 LONG .tk. % @k 6 A w CHA".P V NOTCH OF 20 FTJLB'AT HINUS' '30 DEG F PER NB-2300~ ~ ~"' -P. ~ '.% ( C0v2065-2 SHEET 4 Git 025-22796 -~- m 7 J A0713-02l8b2.115000g (,1 J ~ SS .I. S.-.81Q 5 g 1 m-g 1 3/ 4 65 l '80 1105' 491,0b'l 71 0.23 V J10710-01'802J13199l 0 J 10710-02802J13199 l' 3/ 4 '65 00 1105 49 71p WOK C5) J 10710-03802J13199 1 3/ 4 6 80 '1105 49600Ih100'd8"2 OK /' i l l l l -l i. i i I N,- i. l i 1 7 { g I l M/ l l I l
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,,,t e .he specification numbers //OR7 $-0009/ A // d chown bercort hav,o $cen rect. PLATES NORi%Ll2ED AND STAMPE0 MT I CHEM! CAL ANALYSTS / ,e ggg /y / s ss- / c. w: cu u. v n / ""u' m e u ],4fl2 J 15 0j00 ' . 2 3' L. 0 V yO O; '.019' y'.23' / SUESCMJD I..: s :;.*;;..n go 3gro g y13 802J13199 . 2J/ .99/. 0 06,/.023/ .24f v l r, msMn.rcce 103 h Y+h ?
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..,l.. -[t IMPACT PROPERTIES %, b CUSTOMER: Chicago Bridge and Iron Company PAGE 7 ATTACH' MENT SHIPMENT NO. 803-21023.I DATE SilIPPED 12/24/72.p Ef/ LONGITUDINAL CHARPY V-NOTCH TESTED @ f e' fj SERIAL NUMBER HEAT NUMBER CHARPY SIZE PT. LES. _%'bkJCTILEFRACTUREAREA LATERAL EXPANSIO (inh ~ 44-33 .05 45-is O / J10713 (2) Full 0 J10710 (1-3) 802J13199 Full 3ffI 9 65-2 ~ .03 [.055 i O r, .} g
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.f .s. RECORD OF REAT TREATMENT PAG'E7 ifTACHMENT. CUSTOMER: Chicago Bridge and Iron Company. Si!IJMENT NO. 803-21023 ' ~ DATE SilIPPED 12/24/72 9 S' -RJ' I PLAYE NbRMAi,i' ZING CYCLE ,,).[ .7 FURNACE TIME SERIAL NUMBER NEAT NUMBER TEMP OF THIN) .C0bLINGr-J10713 (2) h03.i1300) 165 36 . Cook: i J10710 (1-3) 802J1319 1650/1665 / 29 (Air) Cool M r N. p'l l J l0 \\ t i l i. j;. . ) ". 4 2 . 'i. - . l.' li..R.,. A9:*. . ; p:[. s. PLATES AND TEST SPECIMENS NORMA ZED PER PROCEDURE PM300. .J .< j ' ' u ,,,._p=h 1.
. ey' 7(.;l LATE NORMALIZED.
TEST SPECIMENS NOT REMOVED UNTIL AFTElR s O 9 0 6 6 - s oo 9.s ai Q w J
3 8 c c, 9 .* BFTHLEH t CORPORATION;. 3FF31 fee 3 F I-- At DEPARTMENT . + - < h [- REPORT .TS AND ANAI.YSIS PLM1 &ntatNT NO. DAlt snetto,.... CAA Ca vin 6CLE NO. c DURNS HARBOR 803-21023 12-24 PC LOUISVILLE LN PC 577403 YAG ~ HIC AGO BRIDGE C IRON CO CHICAGO BRIDGE & IRON.C0 C g 2 BOYLES AL,' BOX 277 t k BIRMINGHAM'AL 35202..,. g 4 ~ 9 IE E 0" M d $ DIAL pftAT httD TTNSLI ttCNC. g g ggg ti St=cm Steca n 5. r,,. avan = "= I t... I w.:.. eco e.oo wo. w.. l u I w. ~ ' 5T EEL PL AT ES-MS-652 R EY 0 T. Q AS 318. R EV O ASME ! SA516 GR' 70 PVQ C 'ASME ' SECT 3 LONG CO).-2065-2 SHEET +3.GH 025 22798. i ,. I p CHARPY V NOTCH.0F 20 FT LB. AT MINUS 30 DEG E PER NB-2300 N.T ' ' j 'T.t j 7 E8d -5 #.J 321.8-02802.i13460. i. I-1/ d. 51 'u'2' 6'O ). ',f985 49C J 33218-038 2J13460; 1 1-1/.8 51 1/ 2 60 905 49 7 8" ,.t O[ 2 J 33218-04802J13460 1 1-1/ 8 1/ 2, 60 98 ie9 7 8 74 00 8.' 2/d J 33218-05802J13460 1 1-1/ 8 $ - 1/ 2 60
985 49 J
J 33218-06802JI3460-1 1-1/ 8 51-1/ 2 60 98 49 4 8"2 J 33218-07802J13460; 1 1-1 8 51-1/.2 60 989 49 74 42[0K l l l,74 f 1153 52 0 7 00 8"2 OK J 0713-01802J15000 1 '--flor @ duw the requireseuw c i odIhospecification t1=bers chorin hereon.havo on c:et. 1 PLATES NORP IZED AND STAMPED MT / j / cut >ccat mAtssns moy0-4-coe W / j 7,, f,' f 3 7 si , e, un e, y, .y vi ff 77 o ef 1 uvuum e '({ 60 2J 13 460 .2/1.06/.006/ 019/ 26' .2Y 1 1 7/ GOV 26/ SUBSCPJrED M:D _SV[OR ! T0 DEFOR 23 1073 T1HS DAY OF. ' &?]A j No..'.RY I C2;t.tc ( i POICE3 COUN'lY JNDIANA ( 31Y CO,p!!SSION ENI'Q}'S SIAY 10, 1975 e @ h. I i .....,,................ _...... _,.. ~...............,......, '..., _ c. gu. }, e
IMPACT PROPERTIES i CUSTOMER: Chicago Bridge and Iron Company PAGE 6 ATTACHMENT SHIP 1ENT NO. 803-21023 r j DATE SHIPPED 12/24/72 1 LONGITUDINAL CHARPY V-NOTCH TESTED G - m.- ,9 'O "'"'^' ""*"'" "'^' """""" "^" "'2"
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'^'""^' '*'*" " ('") ' 6.1440N62[ ,( 037pfSC / J33218 (2-7) 80_2J13_460 Fuli 52 4 4045 0$5 ,2g.050 44- -33 J10713 (1) dO2J15000] Full M 11 1 O 1 e f D.3 i f s .r-i 9 6 t/ Q & -2 2 C / N jyf.of h-
RECORD OF HEAT TRF.ATMENT CUSTOMER: Chicago Bridge and Iron Company PAGE 6 ATTACIDENT .~SilIPMENT' NO. 803-21023 '44TE5111PPED 12/24/72 PLATE WORMALIZING CYCLE "" O,3 ..g%# FURNACE-THIE _ SERIAL NUMBER HEAT 1RTMBER _ TEMP OF_ .(MIN). COOLING J33218 (2-7) 802J13460 1655/167 / 39 Airboo1D J10713 (1) @02f1500 1650 # 36 Air:CooC { 03 4' G-s } 1 1 mU t }. - ft PLATES AND TEST SPECIMENS NORMALIZ D PER PROCEDURE PM300.. TEST SPECIMENS NOT REMOVED UNTIL AFTER [ TE NORMALIZED. ,B, 4,y 70, now-4-ooou Q .A/f Aff
w EE-p b.? o W) BETHLEHEM b RPORATION \\ .2rb I... / ms umttua Aitrar.nr / i U REPORT OF T .ND ANALYSIS \\- 7 .q CA cavemcts no. so,e no. our Sn:ro PC LOUISVILLE LN PC 5774033 PAGE BURNS HARBOR S03-21023 12-24-72 CHIC AGO ' BRIDGE t IRON CO CHICAGO BRIDGE C IRON CO ~~ S BOYLES AL BOX 277 j BIRMINGHAM AL; 35202 ' -s,
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/ / // /[. C / I MLAf / 51 . C. Na Cs V 11 -) C / faa / _P [ 5 QJ_13460)j .2Y 1.12/.2) CU p*26// 26< [ SUBSCREED ED S[OI"' TO EEFOSE ME E D2J 15000 . 2?/ 1.0//.006/.019 23 5 ims MD.w cF. to 73 ... 46., b 50.*.4Y t CSLIC PORTE!L COUS'lY INDIANA /,[ sty contatgstox Ext'luts;.gtAv 10,.1975 .l, c;e p.Q d t e.et.fr th..bo@f'n crmrocn con eined in the records of the Bethlehem Steel Corporotron., s
IMPACT PROPERTIES CUSTOMER: Chicago Bridge and Iron Company PAGE 6 ATTACINENT - SHIP 1ENT NO. 803-21023 DATE SHIPPED 12/24/72 LONGITUDINAL CHARPY V-NOTCH TESTED G' - 4'
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SERIAL NUMBER HEAT NUMBER CHARPY SIZE FT. LBS. % DUCTILE FRACTURE AREA LATERAL EXPANSION (IN) J33218 (2-7) dO2J1356D yu11 2 52.[ 037);.F / ~ e,.050j045-(0$5 J10713 (1) 802J15000 full 6.1"40f2 [ '44- -33 O' m 'c ii 11 O l ,l'.' i \\ lb 'e , gg, o/ e
(v) ~ RECORD OF HEAT TREATMENT PAGE 6 ATTACHMENT CUSTOMER: Chicago Bridge and Iron Company SHIPMENT'NO.' '803-21023 44TE SHIPPED 12/24/72 .55-/ PLATE NORMALIZING CYCLE FURNACE TIME. Q,. ' ' -g COOLING _'(MIN) HEAT NUMBER TEMP OF q SERIAL NUMBER ,i< Ai bool g Ain Coo,h J33218 (2-7) 62J13460 1655/167 39 J10713 (1) 802J1500 V,.. 165 V, 36 /.
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,...; 4;.: <. s:,. -. a w., . n. RECORD OF HEAT TREATMENT 1 ~ i. CUSTOMEk: Chicago Bridge and Iron Company. PAGE 3 ATTACNMENT SHIPMENT NO. 803-14917. - '*~ . DATE SHIPPED 6/30/73. ' ..y. ... ~ PiATENORMAt,IZINGChCLE'.Th'.@"- / . FURNACE. ... TIME . .J.. ,5, (HIN) ..y. COOLING m . TEMP oF_- A SERIAL NUMBER . HEAT NUMBER V
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g. We hecoby c ortify the t the mater $ al herewith conforma to thc ASME 3A-516 CBI Spec. MS-920 Rev. 0-4 S-31d Rev. O. 3553 E; b Mark: 8120306-2)065-2 ~ 7 3 :., l .l' - 4 t \\ {0 SUE 5CRIDED AND sworn TO BEFORE ME THIS.. O I HEREDY CE8'elFY THAT TO THE DEST OF MY KNOWLEDG 6- 'I, th January - _ y _7 y AND DELIE THE ADOVE R - RT 15 TRUE AND CORREC1 33y og _ ...,.a....* h_,,, _ g.. ei: '. '1 r&tm euauc s_.y., u.:,... .. : :n ...;,.,:,;;.;-M.w..a w....:..,.,.u . d. ~- T.~.. e ( nr Couul n exrtREs August 22., 1996'..n.-
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tRANff0 AD, OUT., CANADA
.Jk.:I.9 ' . lAETALLURGICA!./3ATERIAL ANALYSl5 REPORT O I As
p. g i
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_PIPII:3 & EOtIT Mearr rono. p n-2191 ' ' ~ g F71200A c l
215 3RD AVE. NORTH... 'i *f ; g....r!;' 3'
,.o me.,m.s - u.ecaciasy q 9.== f..p. - P.O. BOX 370 ~ j.V '. Cl!ICACO LRIDGE & IRON c i B120306-2065-2 iBESSEMER,'ALA '35020 ^ :b s. i> ew .m we i-.<,is
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I" 12/20% ;.
r. . ;j ' CY:frHIANA, KY. 9 N' ^ f:,i. $'*1( h ~ ta w su---.,. ] 6.* 7 { a, 3. IfM NO.,,, SPECil kATICN 3l DESCR; Pilo 8.:> ! >...i '.T C0 01 ' '.
p.....
Sy . *ASME SA516 Gr. 70 12 PCS.10 S/SO CAP i h( BEAT 780 j 85149 h; r3l4 C H E M I C A L C O M P O S I I I O N g, y.,d. ) f P HT5'TW L P R O P E R I I E.5 f C,'
; pan P
5 53 Ni, b ] 1 . Mo Cu .L YleId.K51 Ultimate K58 Elorg.". M.~. [h/. l j.[57 q 72,5 N '30 D-gQgc.90 - .01,2 .023 .22 $s h- ]} - 1 El ~ 01.1 v .s . g.. o.: -~ _ _M 90sM2't g IEM tio.: * < ' 6 '# ' ', SPkC'i8CAllGN .et t.* CODE.. , 3. s,,.y. NEATNO
*; i:' t DESCRIPitON g.3, ic.
b 3N
ASME SA516 G-. 70 5 PCS.12 S/80 CAP R3M9 i
.p CHE48 CAL C O M P O SiilO N. e -ia v; P iby-s+(,-A3 P R O P E R T I E 5 iCb. l. fb,Mn. .I.Ni l'j Ci [.' ' f.k Cu 'lV O. Yield K58 ' Ull.Imot$ K53 .Elong.4 P 5 Si Red.*. . r., .11.. i;,! ,A .90 .o12 .o23 .22 g; 'm! .F.! ~ " 20? ! .i.3. 5.7.0 72.5 ; h po.. - 5t toi st. AllON DE SCMIFDON. 39 TCS.12 S/40 CAP ' y CODE -l l ' PEATt:0 10 W.
ASME SA516 Gr. 70 FN4J-H
' 94620 ', C hi' ~., C H E M I C A L C O tA P O 5 8 I S O N. PHYSICALPROPERTIES T.C' I !'d* ', '.'le ' o Cu VP Yield K58 Uli; mote K58 Etong.*. Red.*. Ma P 5 5; ' Hi .13 ~ .80 . 010 .016 .18 48.1 7p. 38h -[I i,.

-.sacM

= x-_ rg
"5EPC#HCAleOr4 - ~
ON Copp , ~~ DESCRIPT HEA1 t:0i I ~ - C,4 E M I C A L C O fA P O 5 t i,80 N : P H Y SIC A L PRO P E R IIE S g. En P 5 53 Ni ~ Cr Mo Cu y, Wid K58 Ultimata M5I Elong.*.' Rod.' .) *. 2 __n-- , EMS.5&6 HAVE BEEN.Atteni th a 1 Ar 1Mn' v Ang, wn t ro-nyog A g i U Tempared at l ')OD k' M5 et in Attogo wgiggoAp AND !O 10 Normalizee. at 16a0 1.6 hours. -RE5ULTS WS AE ACCEPTABLE REstnIS WEHE ACCEPTABLE OTES: 'ASME SECTIO!? III - NUCLEAR CLASS l'C TE HEREBY CERTIFY THAT THE IMTERIAL REFE.2"Jf"/9EI'1 C NFORMS TO" A516'G9, 70 ND CBI MS @0: ' i
EVg O & QAS-318 REV. O
./A.<.Ac4 ', DATIL.d" O'77 d 'ILL' TEST REPORTS ATTACHED-q W' ~f38]ESk fb.. TF215 & 6 L!ARPY V" EDICil IMPACT TEST RESULTS': 32ft./lbs. 60 % Shear 9 30 Lat. Exp.
F:
22 @1 E." Q 60.;.U".
Ni 16 ' " . g "a st size: 10 X 10 X 8-65 ~" ' 27 ' 25 sevo fittings were r:etal stornped using round nose low str ecif pg 7 tem // 5 - 10" S/80 (L) SA516-70 FT41!J1
d.. Item // 10 -i(L).}2". S dj.4J1 4 tem " 6 - 12" S/60 (L) SA516-70 FT/.IlJ1 l:..i I
/ 4 (s () na *woan to aEror4 ME 1H85 j-I ut:Esy CEP IFY THAT IO is. OCSt OF A$y'XNow. o- 'toj(Ano p utr itiE AeovE.RErggiay( g g oAY O, n . :.x =D 7o ad idln Nis 5II@ ruoi 3NN f P"'f [ i^ u is: ' 4 D d# nihSH CO. C D ML .... o.: :
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ECS ANGEtES, Ct.ttf.. CRANTTORD, Otif., CANADA.

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. f.. . t.s D ' '..' IAETAl.LURGICAL /AATERIAL ANAL.YSIS REPORT. ' d.W....' .,.d ]. .. g# ro. c.u. -3 3'. 3 cl.p' w........ o. F71200A'. m b 9 'i9 PIPIT!G F. eof tmorr mp. .:h B-2191 r jM, 215 3RD AVE. UCRTH
t CllICACO ERIDGE E"IRONv..o we 3 ~i,,.a ooie even~o.. aa. p$
[ B120306-20659.2h;-.'
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ii se e no., r T --
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b,,btwJ.! "i c
<.:n e m dgy ? e. - n j 72.5 -e.. 8gi wt ' H, 57.0 ~ 30 ~ ..H..o A.,,, .t i ..; ;-.g CODE i;
4 14EA1 N'). I lk M !!O....
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ASME SA516 Gr. 70 5 KS.12 S/80 CAP
,FT4 M-11 t e31.49 p. i g ;,., C H E MIC A L C O M P O,5 8 i t O te. t, [. 9 P'H-Y.-&+t-A' L P R O P E M I I E 5,* 2 CdU.: ' Ma. P S 53 Ni ' <
i Cr ', ~
Mo
Co
.V j,. Yield K51 pitimate K51 Do*2.% Red.' y M s 411f! 1: .90 ".012 .023 .22 JUM-M;i 57.0 .'72.5 30 .!. h ,i lEM l40. br LQe ls'. AllOf4 DESO(IPilO 4 ,t. ; CODE
s l'E Ar t

0 10 *
ASME SA516 Gr. 70 39 TCS.12 S/40 CAP
~ FN4J-H' 94620 C H E MIC A L C O M P O 5 814 0 td,- 4 l' H Y 5 l C A L P R O P E R i t E S ".3 q' ; .;[, 3-i Me 35 5 53 Hi ! Cr (, 1 Mo Cu ,V Yjold K58 UIsimore K51 Eloog.'. Re d.,'. ;,
m
,l f.[. !,,, 48.1 '70,a 3a ,1 .80 . 010 .016 .10 6 s. -___ _ _ - - - ~ - I u #10. - ". 5EPOHCAI. Ora - .,-- DESCRIPTIOr4
CODE, It E A1 t;C'.
4 C H E MIC A L C O M P O 58 I,l o ti ,p P H Y S I C A L P R O P E. R i l E 5 8 Red.'.. ej M.,, P S Si foi Cr, No Cu .V, Yie'd K51 Ull.imate K*>! E!ong.*". t... ,' j;; ,i, __.3
EMS. J.g.S & 6 HAVE DEEri,fpdml tInd A
I fM y Anpymi" quenchod I Te~m. red nf 1 ')nO* l IEMS *E . d 10 Nornalizee., at. 16 8 1.o hours. tri ACCORD WiftficAP AND .RESULTS WEllE ACCEPIABLE s RE5utTS WERE ACCEPTA8tE V pgss;.
ASME SECTION III - NUCLEAR CLASS FC
~~ 7 3)i.q / l ,E HEREBY CERTIFY THAT THE MATERIAL REFERU""cT*PCi(EIL1 C9NFORMS TO SA516 GR, 70 AND CBI MS-920 9 $EVf~}O,& QAS-318 REV. O w M /A #.z,4'A-j /. l LLiTEST I:Ei0RTS ATTACHED' O-77 .pg,4-d l'.- DAT ?I TEM 5 L 6 WlG U$ $ h ..'.l hl!ARPY "V" NOICH. IMPACT TEST RESULTS: 32,ft./lbs. 60$ Shear i 30 Lat. Exp..,, y 22.N ;"J.il - 60 :- ) F '.h * '.3 8' ,.; 16 bs t ' size: 10 X 10 X 8 - 25 ." b 65[ ' ~" A 27 Lbova fittings were netal stamped using round. nose low stress diest sp/40 S 'q ~ ecifit py)).4J11 fgg7 t t:; M - 10" S/20 (L) SA516-70 FT4HJ1 ' "9 Item # 10 -;({.)}2" 6 Top s [t - 12" S/60 (L) SA516-70 FT41tJ1 !...I ERY CER1tFy relAT to Ikjotsf / g N l [8 ssJ AND 'vh0Rtt 10 BEFORE ME IHIS '" ' ~;l, I Ht: F MY'Kt#0w. 8.tnG8. Af4D pitiLF 1HE ADOVE REPO
IRV' ' ' 'f 7 I
,J CORP'.. L k - O V OF lb ^ ^ ^ Fg L kS 23 m.o we w m, ,_,C MuC _ O,,Ex,.-5 ,.x,-15, - y .pmsg ca. x
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M %. ; $,
$M,,,*. 'Qi "c$'yh..+rniu'Cd ci. i. f l I I E* w 6 .t*- y(,.s ...,MAlH PLAHf AND ONCE. CUDAHP, WIS. * * . EOS ANGELES, CAtlF..* BitAN.UCRD, ONT, C.A.N AD A ~.,..' 1.. .a,."":. ". ". q .5. anAaCa P
3. X.....tAar$ Hou5 ton. TEXAS *: FAEYALLURGICAL fAATERIAL ANALYSIS, l.' \\
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.s y. Cons ,.w. ( . HEAFN'O..f ~ 1.,, ' ..*ASME SA516 GR. 70 11 PCs. 3 XH CAP -', hB-HM,.,, f j.:'T.' 'I.47E024 *, 5PECIFICATI, r4 y deaf # y tDE5CRtPitON .lg;j..y O 4 NOple; ,4 :
i. W
. t." P H.Y'5 TT-A' L P R.O P E.R.1 i E 5..; 5. M P O 5 I T I O N i' Q y .d 66 '- C,H E M lp A 1. C a. MRed.[ T L Cr Cu. t,V N'.8 Yield K58 'ptimets i,dsi'Eio'g.% N.jN'[Mo, E.. @t a
5. 0 'O. ;.I. 75,.Y.../,.

45..,' i... :
Q,f P '. 5 5'..' tJi.. 3C U? .Ib.*l'd1 ,t y,.s ~ . 22. /..;:.,. - l. 03 ' '. 011 .024 .20 /. ,_ j a = e.- ~. CODE M:*. l,..r'. W l.. ,HEAI NO. . DESCRIPfiON .L t .c 't-.a,e
r. m.. '.. - -
.o. s.; a. 4 NO / ~ . SPECH.CAllOre ' D" ' i ~* 67E024. ~ I GA4B-H 2 "..-
ASME SA516 GR. 70 4 PCS. 4 XH CAP P H Y S t.C A L P R O P E RJJ E-5 e,
'.h.'V.$ 'Yield"K54[)e$ireote K54 ','Etorig.3 Re d.%. ..C H E A IC AL CO M P O SITIO N / C..Y. v. .e Aw ' ./ f., NL.. Cr.[=,.*,u[ Mo P S Si Cu .. 7.5.5 [. 241. j .s %t. WD ..l J. 50.0 .l.03 .024 .20 ' .,.y: .. e...,. j .0l.*1 J2 6 s.. ,2 .~., ~ DESCRIPIlO p(..g.., p,, , t.,.. e. j,,. HEAT tio. CODE ra ; JPECiflCAlsON J
ASJG. S,A516 QR. 707 11 KS. 6 XH CAP i H/
FN4T-ID e,. ' i 94624..- j' H Y S l C A L P R,p P lj R T E 5,.. c f. g ; .* y;. H E M I C-A L O M P O 5 1,T I,0 N. Yield K51 l Ultimate K,S't '
Elong.%
Red.%,: Cr . : Mo Cd V', 3 C.. h. ~'P." 5 '- c 5~.
' N. i.
~ s, j s... ( ;'s t ' E;. M '.Y 47,8 ./ri70.*1'fJ32 I's c 09 1.07 / .,0,09 .017 .25 / ,, HEAT,NO. COOg...
, DESCP.tPTION zg.
i4 NO.,.. s SEPCiFiCAllON .g.., ,- t
a
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r.7,
,t..g.f. P H Y S I C A L.P R O P E A T I E 5,'- ..h:,;;". C H E M i C A,L C O M P O S I I I,0 N, Red.% - Cv' V** Yield K58 Ultimo.Ie K51. ;Elong.%. S* Si Ni
'. 'Cr * -
.Au ,, q ?.. y, s + - P
JAtt

C :p) 2 -
w .. :.s n
5. :. s
~ yV; r
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,t -.7 v.* pa .,-l. '...x...., 1650 .1675_,,gg. IMS g=.,11 above..HAVE BEEN,Normali20d y*.- _ g, e , < :...g.:... .e 1.5 hours min. ~ / d - r * *, n'.' -. -
AND i
tra ACCORO WITH IAAP r w$ t..yJ RESULTS WERE ACCEPIABLE. +. y
g 's RESULTS WERE ACCEPTABLE i iit.
f m ':. g 1,,qg . p;..? o ..f iiES: FASME SECTION IJI - NUCLEAR CLASS Ip *FERENCED HEREIN CONFORF)S Tq sA51 GR. -{0 PER CBI. r. T. HE HEREBY CERT FY TilAT Tile MATER ~L RE O. nS.J920 REV. AND QAS-318 REV. .., b l M' j.a.' -.ds :.D. i '
e..;: '
N .-f.r...,y. -*,., v., ,W j 3.:p ,- A. e.:. ..37 IL9 ?.'
.4;T)
HILL TEST REPORTS ATTACHED - ITEMS 1.& 2 CBI SUPP IED fuT'L Aboka f'ittings were metal ' stamped using round: nose loes stres $ {os;'specifb.. hings' ares.*. i.l .y '.E ' 51,pgPAfggg"
- "1'./l:
I/ ' ' 1 - (L) 3" Xil SAS16 GR 70 GA4DH y.d75' '. O.. f g) 2 - (L) 4" XH sA516 GR 70 cyBH3 - (L) 6" XH SA516 GR 70 FNITH - y -c .l [ CY,p f Ite m -g 7 QCDS g_
EREBY CE ItFY H A 10 Ti E T4 $7 Cd MY KNOW-

i..RIBED AtlO SWORN TO BEFORE ME THIS
- o y tEDGE AND EttEF i i ABOV REFORI JIRUE AND CORRECI. .y.., cw [ g 3 17 [ _ nw or.
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.G. f r " *""-""hl.. u..f 4.1.. 7,:s.;,. ::.e!%;1AETALLURGICAL tAATERIAL ANALYSISy.p. .; \\). P.... y,z..,1 J W p,. $m...,.,...... 3.Lq l." A
8.s;'.
f:...n u e. ..... o., rit.1u..*.. - - ;.:. y ..e ygoym 3.q., ,e, i.am.;. i w,a g-mm,7 ao-4. ...I,')if B-2191 F71200A . J.r'ay i pPIPIM P MUIPMENT CORP..
f[li. f..q,:.'.
"'** ' ; "'***' "'"' .*.'" '!y'"'.'; '"'y jf52co / ;- &f> C -/ N [.I ,," j ,d,215,'!Rtl AVE.. NORTH ~ CHICAGO BRIDGE P TRON ' 'B120306-7065-7
P.O. BOX 370. '
~ ). f.B.E..S,SEM, E.R,. AIA.,;.i 35020 .'.6.*.' :., - 'M.
'. '. '"".i". t...e.. g..-
. j:a n ' t's.~'I 9/4 /74 -- j .,_.l :%f.y. CYNTHhna. KL
r.
,y.
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.4. W-m*>. .%. s..g-@.t.M,. $ M.~ 39, a.: ';.c C.. q.4,m.p.. w.- p ~. - M. s.:..c.,,. : m., m - ~~ iv x a...~. +..... . u.q I .5, DESCRIPflON y ~l~ ~ * /. 03DE* " y l *,q.,, psyNo. $
No+ b.@ <
5PECHCAfgN y3 f ~ 1? ./*ASME SA516 GR. 70 ,11 PCS. 3'XH CAP ',?'l GA4B-10.i. d 767E024 ? 6. M P O 5 I T.I O N . ~.. ; b ... P H.Y.5 l C A L F R.O P E,R I l E 5 * 'I -Q p A.,1 03..s d,4 XP i. '... C.H E M spa.L.. C'gj S f Si 5. '. O Ni. [>; J i C.cl$t y i'e.Mo,l Cu. .f. V l, ti..* Yield M51 p)moter ksa.*ESek.% MRed.%. ,Ch Ma. .4. 75;5:"l7.g..g./.. e r .024 1.20 ) - t p...
s y..
-G! ik.-l:-~ y'., ' ",,..~jt '. 4 50.0 ' .u 4 s. .22g. " i l ~ /.l.011 ,6 : j / e' ...m ..a.. w; DESCRIPTION CODE *. J .s, n.: *,,HEA1 NO. .. SPEOFICAllOd W NO. ii. + 2*.-
ASME SA516 GR. 70 4 PCS. 4 XH CAP
.i-fGA48-@ ~.* 67E024-5~- .... C H E A BCAL COMPO5ITION a. P P'.rt-CA L P R O P E R T 1 E 5 N..;V f, - J;Ilimee KS8 ',Elo.ng.*.[ '. Red.%.,' ,fMo*' SP /l5 ~ . 53 7 / "' ~ rii. ~ .. Cd....
Mo' Cu YieldT.51
.C .20' f y.; .22/:.. - t ~ ' N- .s :0,,,. 50. 0
2. :7.s.5j.d.,...24..*. <
.1 . 024 O c-01 c- . s a,..,.;
~. -
i, - 9 4E Sp516 GR. 10 /
11 rCS. 6 xH CAP .T :.' - FN4T-H.* ' {,~; 94624 I]t - M ,g PEOFICAllON DESCRIPilO i.:.,..g..,, CODE.; it.4.t..'( . HEAT NO. i f H Y 5 l-C A L P R,p P E R T E 5.4 Y ,5.~.f*OMPOSt.i8,0N CAL .. y'j..... ,tl E,M g , P 5' .: V -7 Yield K5l f. ' qte l}51 !Elong.*; Uyim , Red.%: W . Ni Cr 1 - -. Mo - Cu . Ma ' C *1
.. /r.
-s,'.' ..h .017 c.25 - [ 7,, 47.B t.r70.l f. [32,, .' 7 s '3 [- . 31.07 /.009 i 09 ,.:, 4,. CODE..;, ; ',. HEAT NO. . DESCRIPriON M NO.{ e- . 3, SEPCFICATION .,, = .r; ';. 1
y.e..
v. ,.., C H E,M i C A L C O,M P O 5,1 I I.O N ; .g..,.- P H Y 51 C A.L P R O P E R T I E 5 - v J. ' (i'aipMn C gy (. P.,, S* Si *
. Ni

.
D Cr ^. 4.',
Mo Cu' 4 V,,
Yield K58 U[limpte K58 Elong.%. Red.% - - l '.O y.e.t.,. f. l:;,n ' f /G ... T.
iQ:
i,.,. ' E'.': .l. 2 .. '.u 4 . ;.'.i!;.
l..c ;t,

? -
s r. x5pFJAll above.. HAVE BEEN. Normalized 1650 -i1675 7433 .1 / At .s
1.5 hours min. /
^ I M ** * '/. IN ACCORD Wi!H LCAP AND M. I, .t. l,, -RE5Ut15 WERE Acer. TABLE, 3 .i RESULTS WERE Aw eBLE -3 ~ * -. .f
.g J.

.,.,;.
>TES: t
ii..
GR.'.70 PER CB[ ?.
'ASME SECTION III - NUCLEAR CIASS MC' ' '
s I WEIHERE5Y CERT rY THAT THE MATER ( REFERENCED HEREI!! CONFORMS TO SASI , ic
4. O.! Kd. 4.s :. s.. v.. :.:
- i. F.,:7 .e.@w.W j.Gj ;o m. t..- 920 REV. AND QAS-318 REV. MS.v.y s v. MIi.LrTEST REPORTS ATT' ACHED - ITEMS 1.8.'.2',CBI1SbPPLIED MAT'L p 7 C"3 } [l S 4 ,... ;/g. 6 ... e .9.* *.,:. .p w g- .s Absvy.-f'ittings were metal ' stamped using round.noseilu.v stres os; specift tings"are(.' '. a T ?! ^~ ' %.J 1I. t ' .1 - (L) 3" XH SA516 GR 70 GA4BH
" *i'^ N '.y 5...
yJ 'l. 2 - (L) 4"' Xil SA516 GR 70 GAdBH l' ", ' d g;, M. It 3 - (L) 6" XH SA516 GR 70 FH4TH 1 /f.o qC C v: . 3' 9<X y Alst.i AND SWORN TO BEFORE ME THISO S " 1 IIEREBY CERIIFY THAT TO THE BEST OF MY KNOW-n, Q, Q, 4 . cJ , J .3 LEDGE AND ptpEF THE AB,OVE PEPORT l$ iltUE AND .p) Co%ECT. fu t 08... 59 /) s =
c. -
? M-. ' iud.Ill.'1GICAl /MilMI!% AI I/d..- i.:,...,. wt
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j! ' m,c.:nn:r.u;m m,imic;; aim: - ~ cac...:,c,:...::v
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^ ~ 2.., :o Me 370m cccu r, su.\\.s., a20 u.....~...m.g.. s
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o. qcyp.t._u.
..w,% .r g~ r.r. .m A.. d.n..q;/ e w t..1 .~ r n v. ~ m . JW4_2 '.. .B. LM 3_e.6 s ig ' ' ' ' *g; , 'a ., i t.. _ ___. _ ....f 4'a,'16 Gr. 70 per* 9.u'.'g','Quo CAP. rQg j ; "p".4: g[*pgw egd _'4_ ..q. 7,,m/ re:ab w
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m..
.9i,n!:id.pj v- - is.s.. i .,f 2% 4..,p.. e 96. 012 023 yll h ~.. .r.v-- CChicego Bridge & Iron cpoc'.HS-920. Ecv 'O'dtd 7.11.712. N. 9 a .,(,; [g,' h.Pct.Un.1 E//Jicu ~d oM Iki'tres.mc:: telog.'o:4:'s0-r$J'fo L -Croorso i:j.. % i .,) 7.N .6;GW 70 90f h @%.1/60!Q S. ~ff?, [.MNTI 4f3.". f ri.'9 L ms.jt h,- VI t K6 .q X,
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!9'@4 K rT .' yC.h t c. l '. I E.ha b -753: j. ns., l dh ',Nu. l. l l.s,
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pt- '.' pd . Gvap.fM v' t' i*. . g W ..s. ' ; 8'.. ...- (g:. a pj; 'p*. ,,y, 5 jpT. s.a,... Q;f,. n' ~ 1J
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u r. .y v., . i u"14,1f.o.!.6. cn 70 per-._ _ ___ _. _7 a o u imn,o etr. ppgm g4 nt:.mpl.m ATE.7. g.g3 Onc rur vio e o t o .f t= c % >>, e. r %,...] .;p. . H ;' ' %. m L D.,
Ce,. p. '
'. u. C.e i %:=f. .1 x$) r.5i e t..- l = Y - ,,c i r P S l 58 3 H. -a 4,.3. j g.s,;9,'e,ya
37..,
p giy4 ... ( .olo
o. ;.30
. s y;, 1 c!:. a J, CCMeuco 'M.GSo a Iron cpec.. INS 9,20 Roy O. k.,t.cb7-l I-
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y m L A t t 't at. t; x f* ts'st's
4 Sff E A7
. l sf* A.37 7
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564MS$ ' '.
o3 s 037..~ , g g.y .ni 7 - ..'!O x 10 x 3 0, ,O' ys..,,c :.3 q .i_.... 1:3. 36 L;h.% G-~910 x 10 x 8 03 f:1.1;6y0 $036,O':0, J,9, 10 x 10 x 8 Go '3 7. h t!. 3.2. _ _.. $0..%.h2% t$ Or 4' J2 g,* _4. iT t.".?I!aC 8t ATEP.I At. COHvC:b.15 70 t. Acts.4 $PE C.3r ll 1C AtlL C $$ PIP F l.HICIj tl&$.1H L A
4. y CS II AVC /. Mr.Yts!U;t H ARO:tESS OF in !!Hil '- - - - ' - - - - - - - - - - - - - - - ' '
2 ._IT E. 84.. _.~.._.T. _E.is.t.i L..E. 5.'.r .h' M l'IT Tit - -- - - - - ' -- - - - EN T S O F HS$.,.1.P..;. 5. !.' h -'- ----, FtT TI'lC5 00::70..:
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s. 'h., *
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3. s. 3,
IUt.T RA!Ot:lC AL1.Y 140PECTED AliD ACCr.PT ED PER L A,015tt PROC. STitEtt 8tFLIC /C M l ' (y '.*,.. 3 j 7,,5. h geogn ;fyg y 8 I ..f..._ .............._' O.fi'JtD PLHETRA:li I:4!PCCTCD AIID ACCEM ED PCR L ADi$st PROC. HO RM 1* 'Q *r v e's 9.-.. 'p !ij,,).'j.h"*................_..._.._..._............._...._,..,,..:. e .T.C.s.4P E R U.*O '.r _. % m - # '.dl.0413 R.CICGT.t litCM.t.Y I 45e.CCI ED A:tD ACCf P. T ED P E R ' l q. - 7.. W A1 El: est t.,*:t* .,c e m ,. tj,f. e -r' I " iT TI 4 C% t.*< E C A:.............. _ABLE OF Cc;tre, t'.t:isG 10 HY DFs0ST ATIC, TEST r EC'Jt,it tEHTL OF j
st 1&S*d
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'[~..".} 'o rd T.~.'.t.w. g m, '. t' jT Cc -- CTHTHIANA. ELH1'uCKY 4tC3g l I",.* ti .e f3 O e w e:s. - a 's ? /. uo ...g ! l - \\. I" "**k
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SR Attif!FJ, ONT., CAN AD_A IAETALLLlRGICAL ?AATERIAL ANALYSIS KEPORT"'D.
e i ((.I. 8i wen o.n... - .xs ' ']. cuvo. o.ei. o F7h63?A l'$ - l qttime rt_.CC35t a > B-4197 ' V H. " '*~ ""** ' "' ** * * *)"''*? "h.id&fd .>'2 0 65',6-gd'l', ~215~~3EiD A.VE. HORTH C111CAGO B~tIDGE P, IRON CO. _/ d.O.,DOX.370 ' * ~ ~
4. P if? 'f'f. T " f c g #*
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4. ". '. - ~
it' 'j [ j. ' " ' ~ ' * " ' ' ' BESSEf4ER, ALA 35020 I,Mii - CY ITHTAl!A. l*Y. 1. 12//./71 d..- 3.T:.' .- W T.w. W 4 yLg. .i!Ld. il ;/-:- K.< ~ . d ;.5. r, t i _ CO ' "W.'WM,. " ' ' 'p' 72C2Gi[. ttEAT 1 IIEM NO. 57ECHCAnors SCE.filord $;: 4.u.. lD
AS!4E SAS16 Gr. 70 1 PC. 6 XH CA t.Fl{ l',l: H
.,, j 7,H V UI CA L P R O P E @lQ.V $ 6+ . CHEMICAL C O M P O 5i T,5 O f4, Y.td E5t" Utilmes K$l Etong.".' i! C hg '.;.. %y
v 5
53 . Ni U.ca =j
Mo ' I. Cu d
4,. t. .. Ma e,. P . * + et. l j[,J,,2j9f.96 ' /011.,.o26 L.18. i:- ,, qg:1 s: ii i c, app .4816;>s aa,2.s'.,. 23 y. - ;;,p .~.. T. . IEA5 ' '.A* -:]; e ;.M.. t,. CO.Q.f. l,)i., c ? .....i DESCR.7IIC: 4 . p J. . ;p'... . SikOitCAllON. !,- .( item #40.., - ..e. s. p<.r ., p.; P H Y E I C A L P R O P E R ilE ' i< C H E MIC A L C O M P O S IIIO f4
o Cu V
Y eld'K51 UtgMee K58' Elong.% i R,,:
..l y d t s *
,3 5 Si, Ni l '
Cr #
h.C.n'dI I. Ma ' n h 3, .yf j'4 t- , 4y.-
d
+ i-I 't, riiAt CODE,- DE5CEl?i!Or4 SriOi30^bO84 llEMP40.. ' J [.: P H Y S I C A L P R O P E st 11 E t.- C H LMIC A L C O M P.0 5 t i l O N l Ce Mo Cu '. V, Yield K51 Utiierate K58 Elon2.** Ret. aw <. mh{( j Mo P 5 Si Ni o
2.e
.m DESCRIPisON CODE- - _ _ - HEA: Mr'Of lC AisOta grFM NO. 7,...., C H E M I C A L C O A. P O 5 t i t O N y P H Y S I C A L P R O P, E R T I E " un C
.i P
5 . Si Ni Cr.. Mo Cu 'y Yield K51 Ultimoie,K58 Elong.*. I;c-l 5 'f: -. =. i l. ..q'- ~: .7 1653 'iug i iin m li:n. _ A, _ _., Ayo HAvE eEEH 2 ho.srs. IN ACCORD %,Yly LOAP .ifE5Ut15 WEPE ACCKrAstE /4E5ULTS Y#ERE Af/ EPI,$hE 5, b M/ 7.g" ]?. NOTES: .iCASf4E SECTIO 1 III - f UCLEAR. CLASS TC N .;.UE HEREDY CERTIFY Til.\\T THE IMTERIAL REFERENCED HEREIN Coi! FORMS To SA516 Gn,.70 ' t.E 721 Cat M-N., .i .s,. i,920 REV. O f. QA3-318 REV. O )-
HS-u RILI. TEST REPORT ATTACHED
"' i j .a .f Ab:ve fitting as meta.1 starnped using round 0 hs 3 i ceci) ~c mailiin f a,. \\- WJ i .(L) 6" XH SAbl6 GR. 70 Fil:UH /. E, lW n Of hMl c'; Mr big M. Y.
1'199~.. f' b s @ c-n
~ '9 L occAM W/i93mW 4-CERT:Fy IHAT 10 THE BE$t'0F MY AtiOW $ $CRISED APID 5WORH 10 BEFORE ME THis . t,}... F AllD SElli7 THE ABOVE,REPORI 15 f, RUE / Ara .a0 ~ CORR Cl..,'. .; + ,q* '.. g,g gL '8' - 19 DAY Of . y ;. . o u.U' I 2 e J ~ METALLU T ~ ' ' p g} -NO Any f urusC p. ? taxi *,5 tora EXP RE5
-n g A) *.. /T . :...,. m;} h,.'.d.....;;..&. M2tCTI 211515 R C0011 M __ .:r '? h.j p.= $.. p h rl~ & I R '.Q ' + '~ (y-) s .\\ .... MIA I I. a 4 ~ L2 ( h h --- ' March 10, 3 e, i Fining a Equipment Corp. - CUDAHY, WIS., 19 PURCHASER l B2191 jgy 1 f-M 1.3fo'N o. Q s .\\ A71290A Chg. C 1 PuRCHAsEn'.5 ORoEn No. t l . s. 7.* INVQ N O. ~ AD DRESS (- --:2 m; Decscuer ; M.:. 35020' .h CMDAICAt COMPO51 TION [ ll-kHY11 cal PROPCRTI H CAT N CI. q OCSC7tfFT10N,c .'.s... YiCLDk ULTIM ATE gg,onE ...,h AND$, '#s ? ' f.' i .f.'.' M - N D. AND COO [.I' .C MN-P= $.r gg. Nif C ST NGT STRENGTH PCS. W ' '5PCCIFIC ATION/. ,4 v 1 .;,,.J - (L) 'litNSII2_ h . IR2 AIDIYSI i... 6.. ;ce. . l. Iten.9 m o.cQ
.m
..025 30.. FJ:;3, f.nj.:; 7 - 52 3 80 3: 32e; ~c :
.m= =..
v 20tS:.. v..:S/.$d, Cap N N 1.' F'1%..JIIh>.22.. :-= 2 +...
. :-g ;
.+:3...w m=.y y 3_ 1.01- .010 5 a.,.s :, - n.v..y: .x : ~"~y ~ 3 -* g VET notcli C: tarpy impa :t's~ a a.0*IT (10 x 10 x 3"sp~ccine n)'". "i i .Y t c.xx v,,.' i e:g Y ;i-- L. 20.0
20. 0.- 2 L. O
?t. Ib,s jS 425 44# 3 hear. 025 "id2 S..3 t n7_. ,ateral e::panr ion f-4_ d.:.;..n. j ASIE S'A 234 WP3 perf ,Idb[e b. ? =. lN 9"Wo ?: Tren j'. g-r - IO5ufacture l' A'roIY.a be'co Norbir I t'o ACIE S. 5 Ch *' an 'S oec. u,s, 9032 Rev. 0~ Fittings h:ne a aa:cin 2m ha 'dness of 197 BEN. 1; a: Fittingsite'.'e he it tr 3ateciIU[a^ntcaling atlgs0*Fi yg. ~' m' a'nd QAS 321. @// '.teld twc AO Mk/gd CI qd65' J E ~ hours at co lor. Furn ice c aoled to 100aI. pel hour 6 600*F. e== ' Copy of IIi1 L 'Tes aRep 3rt - httached. g _ 3 Ue he:?eby c artif r tha b the material hert uith conform ; to the , @iE i g:: ASIE 3A 234 WP3 CBI S ?ec. : IS-9032 nov. c QAs l321 specificati ons. l Marlt: 8120 306-2 36y / 1DD/ i SUB5CR! BED AND SWORN TO BEFORE ME THIS I HERESY CE't:.tFY THAT TO THE BEST OF MY XNOWI.EDC ~ i
4..**
AND BELIE't BOVE 'POR S TRUE AND CORRES I 10th
March
" 75 stgtv o y B _ 'Ildde(d
f _ _,.(
s .y d % M . cArr 4'-Sd -77 ~ 22 1976. occx gj' ff3ns g .N fr.. ;. u u.3 : :,,.
/ -
g rruouc S.
y. $'/) d / 0 gr4 kOc.
- MY COMM!5310H EXPIR M'.;ugust r.. .-.... f.. e. - t i. ,c .v ~ ~ -5 .h ^ .4 g.=p..m, m;y3.,..%..JJ4 M 32.,L,'.!.I.8. .Y
. g.,tn..yA..i. ;p...,.e. y..,,r.N..,.m$
~..!.',..'.;,. ~ c.--- p;. en.2..'.. . O d M .M. h.'a.U my.gs. $. pg.:: s,:...s;,.. 7...p;.- 3 e,.3.y s* p. 4.t;... . 'LCO (04 22. - s sea.. g,.. ,. ~..,. g..n 1 z.q.. .yeo c..g..gg. g,yr. q n. M. . x w n e.
P 5 d_C f /M CU'STOM ALLO nt.sta. Cariron..b)...y o783oQRPORATION ~ p)} :'a'.. ( 2ai.822 7 ii. Twxt sio 2as-sss2. Tetex:is.c458 Heat No. ........."...C;,?,,','.,.
.g
@ occk (u r M" " ,9 A.S/5 19 MANUFACTURERS TEST REPORT g, ~ k,,, h-PRODUCT DESCRIPTION CUSTOMER DATA .....M.c.,,,J,unhin.,,, Corporation.,,h g& h O . Cp........... NAMC ercu h 34-20923 ^ 6" PS P.O. N O. M M SIZE .S...c..h.... 8. 0.. ...(.4.. 3.. 2.. )............................. WALL A i.i.o y ...W.... ................. ; o a N o...N...0.. 3..5. 0. 4..... 3... p,- nda. n o* """""707"1-q~60073 ASME SA-234 Sec III CL II 197: 3o s,Faier vi; Is51irain,.s varsiliEE"T971",""Ed'de " " ' " " ' " " " ' " " " " " " " " ~ " " ~ ~ ~ ~ " MILL CHEMICAL ANALYSIS viero Tensa'e ciono. na w u steria Point Strength in 2" Control N o. and PSI PSI Specification C Mn P S St Cr Ni Mo Cb '*'e .28 .79 .007' .03d 225 i 54,100 78,800 22.0 CAC 3361 38 220 73>820 38.0 ASTM A-515 - Cn ca Grade 70 L:ou Check OTHER TESTS _PJBf0_B_MED CN THIS PRODUCT [ Hardness 72 RBD T.Chgr w v nr ur PT suCV w. d i G=5) Material was hot formed and finished between 11500F - 1800 F. EED - REMARKS: Then cooled in still air. b, Stato of New Jersey J y County of Hunterdon .t Acceptance and Approvas my customer N c--- ft Sworn and subscribed before me th's nearetenistive/ inspector C:2 I certiry the above product has been rnanutscturce in ac. 7g hy of,,,, 4 <*
19,,,,,,,,
cordance with att olicabte parts of the aeovo oraer and / 1' # P A Pudt." S k f ',,0 j fn"[,0",b ~ My r u.stan Empi s' Dec. 30.197 ,,i
A-_i-c.- - Q", :. u .... C .i. ..o - r. f.
.4
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' O_C ~ RPORATION ~ .l C ged USTOM ALLO -)...yc282o ~ .m ............C,;,?, ni.si3,C nron.L 76 201832 7111. TWX: 510 235 3362. TEl.EX: 13-6456 Heat No. e MANUFACTURERS TEST REPORT g, h.h PRODUCT DESCRIPTION CUSTOMER DATA \\. - .o ,,,,_,,,Mc.,,,Junkin,,,Co,g, oration,,,,,,,,,,T,,[ O . CAR,,,,,,,,,,,,,,,,,,,,,,,,_,,...,,,,,,,_,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, 6 iTCu u,uc ~ .6,',',,,,,,PS,,,,,,,,,,,,,,,,,,,,_,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,_,,,,,,, M-MM Q h sizC P,o, u o, .S.c h,,.8.0..(.d 3D..... 9 wau. a u.o y ...W....................... a o a r. o. .N...0.. 3..5..0. 4..... 3... p. ASME SA-234 See III CL II 197: nda. no* """""7071-q' ~60073 3,Ediff6ii",""fif5Ndi,s"YiiiiiiiiEE"T97I"""Ed' e 3a " " " " " " " " " " " " " " " " ~ ~ " " " ~ " ' " n d MILL CHEMICAL ANALYSIS vloid Teamh I Eto ng. naw Material Polni St,ength in 2" Control No..nd C Mn P S SI Cr NI Mo Cb PSI PSI Specificatio,s t * .28 .79 .007' .036 225 l 54,100 78,800 2'2.0 CAC 3361 3g A8X 38.0 ASTM A-515 Ca2.w i.uis Crade 70 @htCW OTHER TES,TSEEREOfLMED ON THIS PRODUCT m ~ [ Hardness 7." RBD 1. Char w v l RT UT PT HUCY g4 w 3 y/ 0 0 g i REMARKS:. Material was hot formed and finished between 1150oF - 1800 F. p 9 Then cooled in still air. g Ststo of New Jersey b c) C unty of Hunterdort y 1 m Sworrt end st:bscribed before me this A..e.t.nc..no A.ppro,v.a...my c.u.stomer 4 n...... n.. 192f i.., u r, i n..........,.....,,,....... i,. B2off f;,, of,4 c. 'hl4(f &$ ~"*
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NATION Al. FORGE CO. ORDER: 60-5732 %9 USTOMER: Alabama Power Company FthP2-100 SPECIFICATION: ASME Code, Sec. II Table T-1.0 ASME Sec_III USTOMER ORDER NO.:- R q.
46555-SS-Const.-10531200-SS-1102-68 described on Scuthern Services DrW.
N A-1102-68-2, Rev. 4 & detailed Spe h. DESCRIPTION: TEM.SE RI Al. NO. Flued Head Penetration. " Inquiry No SS-1102-68,Rev. 4" '3-001 linit #2. Eenetrction.No.- rawing 4-01801, Rev. O AstE-SA-105.71, Code Case 1519 n/ Q2H1192" EBB 1A Ni. Cr. Mo. V. ITEM. SERIAL HEAT NO. C. .. M n. P. S. Si. 23-001 2 25 .87 .009,,.008 23 .10 .20 03 .03 f = CUDP ULTIMATE YlELD IMP = aREDUCTION e{,ghggp F g LAt. EXD. -). TENSILE . D, F.LONGATION oF ./' STRENGTH AREA (P.S.t.) (P.S.I.) 23-001A 77,000 50,500 32.0 74.5 14% 37.0 .048 23-00ln 76,500 52,000 32.5 73.3 31% 28.0 .064 A 32% 72.5 .060 NilSTENITIZE3-1500 F for 9 hrs, Q 1enched in water 0 254 54.5 .052 1240 F for 9 hrs 30% 63.5,. .063 - B 0 IEMPERED @ IESTS STRESS RELTEVED E 1125 F for-i hrs. hea : rate 38% 7 0. 0 .065 10tP F/hr. Furnance cooled 1000F/h :. down to 6000F HEAT TREAT PROCEDilRE lir-60-A-5732-0A, Rev. C, 10/16/73 IMPACT PRO DliRE LT-60-A-5732 0A, R zv. B, 12/20/73 FREE OF ME RY AND RADI0 ACTIVE CONTAMINATION R0!!ND FLitT INCOT MOLD' BRINELL 174-174 FORGING TESTED AND ACCEPTED IN ACCODANCE WIT i APPLICAB ;E ORDER SPEC. / 4 %,'.) AND DRAWING,RE0"IREMENT' S State cf P;nnsyIvania Warr:n County ss: I . C. h, h. B; fore rne, a Notary Public in and for above County, personally appeared Yrs /O d I abo e cf the National Force Company, who being duty sworn according to Law, deposes and says that th c rrect copy of tests as contained in the records of the Company.y.[hh...... h. ....'.p n DN, jD g 70 acribed a$,d Sworn to [p O y 6 February g,,,7 4 this day of Y..UAU.SL.A/. @. V . !)..~ m, t +- :r s:1.:en :, ty D f My Commission expires Mr.: t.;; :.., u gg Q2A18.{ v < re..,. s,., ;.,, t i.,n
Sisriio.tossa w h'. NATIONAL FORGs COMPANY I. .i TEST CERTIFICATIO.N m,..= NATIONAL FORGE CO. ORDER: 60-5732 11STOMER: Alabama Power Company
USTOMER ORDER NO.:-
FNP2-100 SPECIFICATION: ASME Code, Sec. II P-Table T-1.0 ASME Sec. III ' Reg.
46555-SS-Co'nst. -10531200 -SS -1102-68 described on Southern Services Drw. 4 1 TEM.SERI At. NO.
DESCRIPTION: A-1102-68-2, Rev. 4 & detailed SpecF %D01 IInit Flued Head Penetration. " Inquiry No. SS-1102-68,Rev. 4", Fen;tration Drawing 4-01801, Rev. O K. Q2N11x32 3-1B ASME-SA-105.71, code Case 1519' ITEM. SERIAL HEAT NO. C. Mn. P. S. Si. Ni. Cr. Mo. V. ,m 4001 T14-2684j .29 .92 .C09 .010 .27. .10 .20 .03 . 0'3 Cuar ULTIMATE YlELD I"
REDUCTION, ej Shear
. TENSILE F 5 LAt. Exo. '
23u, ELONGATloN oF STRENGTH AREA (P.S.I.)
(P.S.I.) 32.0 71.8 36% 51.5 .049 $4-001A 81,750 62,000 33.0 70.8 57% 88.5 .071 A $4-001B 82,000 61,000 50% 83.0 .072 kilSTENITIZED-1500 F for 9 hrs, Quenched in water 43% 73.5 .076 FEMPERED @ 1220 F for 9 hrs 40% 81.5 .073 D l 'ESTS STRESS REETEVED 6._12409Ffor ihrs, hea : rate 43% 60.0 .05 100) F/hr. Furnance co:)1ed 1000F/h c. down to 6000F (EAT TREAT PROCED11RE Hr-60-A-5732-1A, Rev. C, 10/16/73 fMPACT PROC' D11RE LT-60-A-5732-0A, Rev. B, 12/20/73 FREE OF ME ORY AND RADIOACTIVE CON CAMINATIONI 10!!MD FL11TE INGOT MOLD 3RINELL 174-179 kORGING TESTED AND ACCEPTED IN ACC0 1 DANCE WIT I APPLICAR LE ORDER SPEC. '\\ WD DRAWING REO"IREMENT3 k[ Statx!P;nnsyivania f ss: ' rr;n County TRUCDO U#' ..d B; fore me, a Notary Public in and icr above County personally appeared.. cf the National Forge Company, who being duty Sworn accord;ng to 1.aw, depous and says that the abo e ,q-i c::rrect copy of tests as containedin the records of the Company. .5h _ h... k. : O) ... f e - ich f, ~+ 4, , ~oscribed and Sworn to 18th de,y of.. February.... - 19 7 4 h.s C# C.'.\\ a j q,7 th.is .'f..U.2u!M ! y' S mm m v D 2 "Q2Al8.05
My Commission expires

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a NATIONAL FORGE COMPANY ',2 ) TEST CERTIFICATION f e Q(() ..3_ -.!! Wt ~ NATIONAL FORGE CO. ORDER: 60-5732 - e.-.9 JSTOMER: Alabama Power Company FhP2-100 SPECIFICATION: ASME Code, Sec. II JSTOMER ORDER No.:. Table T-1.0 ASME Sec. III BCq. #46555-SS-Const.-10531200-SS-1102-68 described on Southern Services Drw. EM.SERI Al. NO. ' DESCRIPTION _: A-1102-68-2, Dev. 4 T, detailed Spec { t-001 Unit _ ' 2 w Flued Head Penetration " Inquiry No. SS-1102-68, Ret. 4"
natration "3 Drawing 4-01801,'Rev. O X Q2N11x32" ERB-lC ASME-SA-105.71, Code Case 1519' l
ITEM, SERIAL HEAT NO. C. Mn. P. ' S.' Si. Ni. . Cr. Mo. V. .5-001 L J _2,7_08 .30 .90 .010 .008 . 25._,. 10 .18 .03 .03 .~, Charpv ULTIMATE YlELO IMPACT. i TENSit.E REDUCTION N Shear . OF MF4.M51. LAt. Exo. - . 2n, Et.oNGATION.:- p) STRENGTH T3' .L U t U (PS.I.) (PS.I.) AREA g !5-001A 83,000 57,000 32.0 71.6 25% 45.0 .041 t5-001B 81,500 61,000 32.5
71. t.1 34%
51.0 .050 A 43M $0.0 .047 IISTENITIZE0 - 1500 F for 9 hrs, Quenched in water 0 34%
52. 0..
s.049 EMPERED @ h RELTEVED @ 12400F for 9 hrs 36% 59.5 .060 - B ESTS STRES 11250F for i hrs. hea rate 41% 60.5 .058 In(P F/hr. Furnanee coaled 1000F/h :. dosn to 6000F EAT TREAT PROCE011RE Hr-60-A-5732-QA, Rev. C, 10/16/73 MPACT PRO DilRE LT A-4732-0A, R av. B, 12/10/73 REE OF ME URY AND RADIOACTIVE CON CAMINATION OllND FL11T INGOT MOLD RINELL 174-179 3RGING TESTED AND ACCE?TED IN ACC0 1 DANCE WIT I APPLICAR LE ORDER S?E j ND DRAWING, REQ"IREMENT3 p g,,- tatacf Pannsylvania !arr n County ss: N. C. Bax ' r, J. Bef:re me. a Notary Public in and for above County, personally appeared cf the National Forge Company. who b,eing duty Sworn according to t.aw. depas an rflg O, v corr:.ct copy of tests as contained in the records of the Company. g,,[..... 8gg... .J O.cribed a.-J Swur. to $j cgf. (q c/ - MAR 20 1974 t-iSth February .19-.~7 4 jf g .nis _....... .. day c,f ~ ..N..U CL@A.D.) -~. - - C n.A www 5"m h ' D**
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Y ' N*~?N TIONAL FORGE COMPANT ~ I jy TEST CERTIFICATION USTOMER: Alabama Power Company NATIONAL FORGE CO. ORDER: 60-5732 g 4 USTOMER ORDER NO.: FNP-2-100 SPECIFICATION:' ASME Code, Sec. II eq. #46555-SS-Const.-10531200-SS-1102-68. Table I.l.0 ASME Sec. III described-REM.SERI A1. NO. DESCRIPTION: on Souther'n Services Dwg. A-1102-68-17 26-001 Unit Flued Head Penetration Rev. 4 & detailed Spec. " Inquiry No. ':nitratio'ri'h.2. Drawing 4-01802, Rev. B SS-1102-68, Rev. 4" ..f.3 Mk.Q2N21x14"DBB-1A ASME-SA-105.71, Code Case 1519 ITEM. SERIAL HEAT NO. C. Mn. P. S. St. Ni. Cr. Mo. V. (hM.,23 .85 .012 .015 .22,.17 .31 .05 .04 26-001 a= LONGITUDINAL ULTIMATE flELo QiAtW REDUCTION lMPACT TENSILE
pg ej "^
ELONGATION @ WOW 3 LAT.EXP. 8"EN,G,W p (Ps u A A 2%L 76,250 52,000 32.5 ^ 72.9 61% 118.5 .081 $USTENITIZGD @ 15000F for 6 hrs, ( UENCHED Ib WATER ~ f25*.' 4 ~ TEMPERED G 1240 F for 0 hrs EESTS STRESS RELIEVED G ll250F for 5 hrs., ht at rate 1000F/hr Furnance c6oled 10(PF/1 r. down tci 6000F EAT TREAT PROCEDURE HT-60-5732 -0A, Rev. C, 10/16/; 3 OMT6fj,N [MPACT PROCEOURE LT-6(1-5732-0A,Rjv.B,12/q0/73 8@h.' r. /t . O,7,. FREE OF ME{CURY AND RAVI0 ACTIVE COETAMINATIOF d ^ o kOUND FLUTr.D INGOT MOLD m BRINELL 179-174 . Apg <, RDANCE WITH APPLICAI LE ORDER SP $ !y R FORGING TEg'TED AND ACCEPTED IN ACCC 4 .A AND DRAWING REQUIREMEN7S OO ( C. N D 5 tate c! Pennsylvania i W!rren County sst N. C B ter, Jr B;for2 me. a Notary Public in and for above County. personally appeared that the.a ove Rgp/ ortis a true and cf th? '.4ational Forge Company, who being duty Sworn according to 1.aw, deposes and sa cc.rrect copy of tests as contairied in the records of the Company. ~2 .; [ ' A -- ,f - - - - - ~ ~ ~ ~ 4:ribed and'S ;orn to "18 bh " ' ' day of AE#i1 19 7 # ~ . 20g.2nw gj My Comrnission expires g y.,.g -, g., pg, C.yv e,.:ne, v.:....., m.v. P:n.m re:3.N='sw P m Snt /J QQ ne cun..u:;.n ts;ms Jw'r 2. In.s _q,,,,,g 03 FNP
M. \\ NATIONAL FORGE COMPANY - W' TEST 2:ERTIFICATION . g-g .h h. o 0. i S NATIONAL FORGE CO. ORDER: 60-5732-: p-r. l
USTOMER

Alabama Power Company 7
USTOMER ORDER NO.: FNP-2-100 SPECIFICATION: ASME Code, Sec. II J
Table I.1.0 ASME Sec. III described tg. #46555-SS-Conse.-10531200-SS-1102-68. on Southern Services Dwg. A-1102-68-2, 'l DESCRIPTION: YEM.SE RI Al. NO, l27-001 Uni ,2 Flued Head Penetration Rev. 4 & detailed Spec. " Inquiry Nog.5 ' tnatration Drawing 4-01802, Rev. B .SS-1102-68, Rev. 4" ?M-Mk.Q2N21xiti"DBB-1B ASME-SA-105.71, Code Case 1519 ITEM. SERIAL HEAT NO. C. Mn. P. S. Si. Ni. Cr. Mo. V. i~3 i86, 26 .96 .011 .007 .19 .17 :.27 .06 .05 h - i 27-001 , gg: ~ = f/= LONGITUDINAL, CHAKt'X ULTIMATE YIEt.o wPAct reduction % SHEAR .) LAT.EW* ~ ]. TENSILE
pg ELON G ATION,,,
OF STRENGTH (P.S.I.) ' AREA
(P.S.I.)
27-001 L 83,000 61,500 30.0 73.5 68% 123.5 .086 76% 144.0 ~085 RUSTENITIZyD @ 15000F for 6 hrs, (UENdHEDIt WATER 82% 145*0
103 TEMPERED @ 1240 F for 0 hrs TESTS STPISS RELIEVED Q ll250F for 5 hrs., heat rate 1000F/hr. Furnance c6oled 1000F/br. down t6 6000F HEAT TREAT PROCEDURE HT-60-5734-0A,Rev. C, 10/16/~ 3 IMPACT PROCE00RE LT-60-5732-0A, Rgv.-B, 12/F0/73 TAMINATION FREE OF ME{CURY AND RA 10 ACTIVE CO:
ROUND FLUTr.D INGOT MOL BRINELL 179-179 PTED IN ACCOP3ANCE WITH APPLICAI LE ORDER SPEC. FORGINGTEjTEDANDACCI: AND DRAWINN REQUIREMEN':S A-% 1 / ff,0 % Stat 2cf Pennsylvania Warren County sst ~ N. C. Baxter, Jr. CM.U.b[8(., B2fcre me. a Notary Public in and for above County. personally appeared a true r[O, cf the National Forge Company. who being duty Sworn according to Law, depese: and says that the ab r. O c:rrect copy of tests as containedin the records of the Company. A & A s.g..g % q.. l
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.,.v. a.. :.,, O .E eribed and Sworn.tb.. " 'h'..'..,". x 18.t.h diy of,..i..L.co.e.r.il 19...7$. n ?T cot ...h. b '.:3.O.,W /f My Commission expires. i p [h y-rf-?y .x
tol~$P NATIONAL FORGE COMPnNV TEST CERTIFICATION . ei mm}D m]D[ ] D w@N
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1 60-5732- $ e: NAT10NA1. FORGE CO. ORDER:
USTOMER:
Alabama Power Company s. \\ fuSTOMER ORDER NO.: FNP-2-100 SPECIFICATION: ASME Code, Sec. II pg. #46555-SS-Const.-10531200-SS-1102-68. Table I.1.0 ASME Sec. III described TEM. SERIAL. NO. DESCRIPTION: on Southern Services Dwg. A-1102-68-2, Flued Head Penetration Rev. 4 & detailed Spec. " Inquiry No. l 28-001 Unit 2 Mn ;tration @ Drawing 4-01802, Rev. B SS-1102-68, Rev. 4" -M; ? ' ASME-SA-105.71, Code Case 1519 Mk.Q2N21xltl"DBB-lC 1 ITEM. SERIAL ' HEAT NO. C. Mn. P. S' Si. Nr. Cr. M o. V. 1 l 28-001 1 35016 25 . 9t; .010 .009.31 .17 ,20' .05 S ~ ~ ~ - a. - n. L= LONGITUDINAL ULTIMATE YB E t.o CH AKt'X REDUCTION IMPACT 'e TENSILE po/ .@W0bf1 LAT'.EXP.
YPs:""E
-[P s.t.1 = 7* A A ~ 2 )1 L 77,500 55,500 33.0 75.1 64% 130.0 .092 ~ 73% 135 5 ~097 AUSTENITIZVD @ 15000F for 6 hrs, (UENCHED Ib WATER 66% 124$0
089 TEMPERED @ 12400F for 0 hrs s~STS STRESS RELIEVED 0 11250F for 5 hrs., heat rate E
1000F/hr. Furnance cooled 10CPF/hr. down tch 6000F HEAT TREAT PROCEDURE. HT-60-573E-0A, Rev. C, 10/16/; 3 IMPACT PROCEDURE LT-60-5732:0A, Rgv. B,12/2 0/73 +f TAMINATION FREE OF ME{CURY AND RAyI0 ACTIVE C0! d ~~ ROUND FLUTr.D INGOT MOLD 4P,7 BRINELL 170-174 H APPLICAI,LE ORDER SPEC. o 8d' k FORGING TE' TED AND ACCEPTED IN ACCORDANCE WIT O s/. AND DRAWIN REQUIREMEN'?S s N h$0< stat 2cf Pennsylvania Warren County ss: N.
aner, Bif;te rne, a Notary Public in and for above County, personally appeared df the National Forge Cornpany, who being duly Sworn according to t.aw, deposes and says tha (h, above peportis a true and
. -[ ' correct copy of tests as contained in the records of the Cornpany.-.A *- $ ,.., t sl.,.. cribed and Sworn'to ' ((o 4 ~ C t.t I', April 39..7 4 18'.th day of 4g ".. S..C.).AlS. M ~ - g N My Cornmission expires.... r /....J.// win., *. t.,,en Cav,. **yr. o. 24'M / P.Va Seal M.03ThE N ~/8 AAy Coent:sm tigen J sary 2,1778 e,at t Y t W VeAt
~'" %" U I NAT O AL' FORGE COMPANY - gyl TEST CERTIFICATION e. O**D~*D 8 2 .a oa w
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%.~-*.?* NATIONAL FORGE CO. ORDER: 60-5732 ~ CUSTOMER: Alabama Power Company SPECIFICATION: ASME Cod", Sec. II CUSTOMER ORDER NO.: FNP-2-100 R:q. #46555-SS-Const.-10531200-SS-1102 68 Table I.1.0 ASME Sec. III described on Southern Services Dwg. A-1102-68-2 Rev. 4 & detailed Spec. " Inquiry N N,s-ITEM. SERF At. NO. DESCRIPTION: U .29-001, Unit #2 Flued Head Penetration SS-1102-68, Rev. 4" !P ngtration # @ Drawing 4-01803 ASME-SA.-350,GRLF2 hk.Q2G24x2"CBB-3A . ITEM SERIAL HEAT NO. C. Mn. ' P. S. Si. Ni. Cr. M o. V. {4-2$~32]'.25 ~ ~ .90 .0,10 .012 . 21.,,. 10 .17 .03 .03 29-001 L= LONGITUDINAL (for 2nformation') /- ULTIMATE YlELD CHARPY TENSILE REDUCTION IMPACT-oF 4 SHEAR (FT.LBSJ LAT.EXP. gT- . STRENGTH . PT EloNcATIO& m. (P.S.I.I (P.S.I.) AREA @ +10 F 29-001 L 82,750 61,750 30.5 7,4.1 84% , 143.0 .094 78% / \\ 131.0 085 AUSENTITI7,E3 @ 1500)F f3r 2 hrs, QU INCHED IN NATER 68% \\l11.0 . 81 TEMPERED @ 12400F for 2 hrs TESTS SPECliENS STRESS lELIEVED @ l L25 F for i brs, hea : rate 100 F/hr./Furn mce cooled 10 ) F/hr. down to 600 F \\ / HEAT TREAT PROCEDURE HT-60-5732-0B, Rev. B, o E 10/16/73 N IMPACT PROCEDURE LT 6 0- A-5732-0B, Rgv. B of 12/20/73 / T FREE OF MERCURY AND RAD 10 ACTIVE CONTAMINATION / / STR ROUND FLUTE 3 INGOT MOLD O FORGING TESTED AND ACCE MED IN ACCO IDANCE WIT:I APPLICAB1EORDERN3C.h'.k[,@ycp Sp-n p AND DRANING REQUIREMENT 3 e I O 7Sf4 ,O'[ State ef Pennsylvania o Warren County ss: N. C. Baxter, 1 B2 fore rne. a Notary Public in and for above County, personally appeared.... cf the National Forge Cornpany, who being duty Sworn according to Law, deposes and says that the ab kortigag c rrect copy of tests asgontained in the records of the Cornpany. "' - ~ ~ ~ ~ - - - - - - - d, J' i .osciihed and c .his,,,,13,,[*y,..... day of......M,,p.y.......... ,.,,3 9,,,,7,!,s,,, g ",' Okyf ...RLh9w. r e.>swd threon b.'4 My Cornndssion expires,,,,,,,,,, lev;ne, S.'/serra Cu. e.. Permy!ien3, ibia*y Pub'e W T5Eal My Corn.nisnin ty*.res Jer. vary 2,1978 n.,.unns vnp A '/hd <
v-NATIONAL FORGE COMPANY \\ \\ r . TES C ERTIFICATION A9 9, ,3 e a e y NATIONAL. FORGE CO. ORDER: 60-5732. g _. CUSTOMER: Alabama Power. Company +... SPECIFICATION: ASME Code, Sec. II CUSTOMER ORDER NO.: FNP-2-100 Table I.l.0 ASME Sec. III described _,' R:q, #46555-SS-Const.-10531200-SS-110'2-68. on Southern Services Dwg. A-1102-68-2, DESCRIPTION _: No. Rev. 4 & detailed Spec. " Inq ITEM.SERI At. f40. 30-001 . Unit #2 Flued He'ad Penetration - SS-1102-68, Rev. 4" e P;natration $ Drawing 4-01803 ~~ ASME SA 350 Grade GR LF2 [ Mk. Q2G24i2"CBB-3B. f lTEM. SERIAL HEAT NO. C. ' M n. P. S. Si. Ni. Cr. Mo. V. 25 .95 .010,.008 20'.12 .21 .04 .03 30-001 - p 19 2586 L= LONGITUDINAL . ~ ' CHAKt'X ULTIMATE YlELo wpAcr REDUCTION TENStLE
pg
% SHEAR @-5T0@ LAT.EXP. ELONGATION sT STRE N,G,m gp (p,,, 1L '82,500 61,750 3 0. 0,. '75.9 100% 17 1.0 .102 77% 152.0 .095 AUSTENITIZFD @ 15000F for 2 hrs,,(UENCHED Ib WATER 77% 144.0 .084 r TEt;PERED @ 1240 F for 2 hrs TESTS STRE$S RELIEVED 11250F for 5 hrs., heat rate 1000F/hr Furnance c oled 1003 FAr. down to 6000F HEAT. TREAT PROCEDURE HT-G0-A-5732-0B, Rev. B of 10/16/73 IMPACT'PROI EDURE LT-60-A-5732-0B, Rev. B of 12/20/73 FREE OF ME CURY AND RA I0 ACTIVE CO2TAMINATIGO i ROUND FLUT D INGOT MOL FORGING TESTED AND ACCMPTED IN ACCORDANCE WITH APPLICAT LE ORDER' SPEC. AND DRAWINC REQUIREMEN':S ' rara 9reTlW/TiPTM M' :. Stat:rcf Pennsylvania d Warren County ss: O.,, ( h Before me. a Notary Public in and for above i.oun'y, personally appeared... ..h.k..C.,,},,,,f,,{,,,,,,y,r. ot and Stays-thaShe,ve~~ cf the National Forge Company, who being du!) korn according to Law, depose c rrect copy of tests as contained in the records of the pany. _.a.... 4:"'.. 9 L. '.z (C.',,y,27, W 8 ubscribed and $dc,rn'to. b*'4 M.. 'N., ........1. 7...t..h.... W.. day of......Ma Y.. . 19.....il p is ~' .Y.YkL N0M.. bd 'd GI. C3NTRO

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NATIONAL FORGI: C; OMPANY. ~ .~. 'W l. TEST CERTIFICA' D oo .~
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g, Q. NATIONAL FORGE CO. ORDER: 60-5732 %.~-T- . CUSTOMER: Alabama Power Company SPECIFICATION: ASME Code, Sec. II CUSTOMER ORDER NO.: FNP-2.100 Table I.l.0 ASME Sec. III described ~ R::q. #45555-SS-Const.-10531200-SS-1102-68 on Southern Services Dwg. A-1102-68-2, ITEM-SERF At. NO, DESCRIPTION _: .Rev. 4 & detailed Spec. " Inquiry No.h, 31-001 Unit 2 Flued Head Pene'tration. ~ SS 02- ~ Pznatration No. Drawing 4-01903 g g 3 Mk. Q2G24x2"CBB'- ITEM, SERIAL HEAT NO. C. Mn. P. S. Si. Ni. Cr. Mo. V. 31-001. 24-2586;} .39. 75 .009 .008 20 28 1.00 .31 .04 . i *' (For Information) L= LONGITUDINAL CHARPY ULTIMATE YlELD REDUCTION IMPACT TENSILE 2o/a ELONGATION OF % SHEAR (FT. LBS.) LAT. E)(P. f STRENGTH AREA
-(PS.I.)
(P.S.I.) 31-001 L 82,500 61,750 .30.0 . 75.9, @ 10 F ~ 100%' 171.0 '102 AUSTENITIZyD @ '1500 F for 2 hrs, QUENCHED IN WATER 77, 0 095 TEMPERED Gl1240 F for p hrs 77} 152l0 0 yqq 0 .084 IMPACT PRO:EDURE LT-60.-A-5732-03, Lev. B, of 12/20/73 HEAT TREAT PROCEDURE' HT-60-A-5732-0B, Rev. B of 10/10/73 FREE OF ME CURY AND RADIOACTIVE C0b TAMINATIOb ~ ROUNDFLUTDINGOTMOLh [T,A~ggm FORGING TE TED AND ACCCPTED IN ACCC RDANCE WI'1 H APPLICAI LE ORDER. EPE 4 Cf ' AND DRAWIN REQUIREMENTS g AtJG 5 1.974 D U Stata cf Pennsylvania j 55: Warren County ...N,,,,,C,.,, Ba x,t,e,r,, Jr. f Before me, a Notary Public in and for above County, personal ppear cf the National Forge Company, who being duty Sworn accord o L.aw, depos4 and says that the above Re rue and ~ CONST
b correct copy of tests a contained in the records of the Company.
/1.. W / ..x. C ~:3.. ..:2i.. .\\ O bscribed and Sworn to' $edu a.h n 8 ........... 19..7 4 l AUGUST g 3 1s t;.t......... da. 'f...1 - o th.is........ 4 Cl 1 -E lb.. L..h.1}%o !...... 0( n r
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l My Commission expirN.... .........T.&rd..!!c.'.':M ' ?.r........... CO::T i,..:viv3,,.n r u f. r.<n +..:. m,, rew s..' 3.m J . ik/ Cw.m:t*tn t'a,*,4s J my 2. 1918 u
NATIOK A T. FORGE ~ COMPANY ,y; @.ST CERTIFICATION O o**Jo
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~ v_.g { %'M.T.l CUSTOMER: Alabama Powei Company . NATIONAL. FORGE CO. ORDER:60-5732 CUSTOMER ORDER NO.: FNP-2-100 SPECIFICATION:.ASME Code, Sec. II '-l R:q. #46555-SS-Const.10531200'-SS-1102-68 Table I.10 ASME Sec. III described ITEM.SERI AL NO, DESCRIPTION: on Southern Services Dwg. A-1102-68-1,.f Rev. 4 & detailed Spec. " Inquiry No? Flued Head Penetration 38-001 Unit @2 ASME-SA-350LF2 P;natration Drawing 4-01810, Rev. B Mk.Q2P15X3/8"CBB-4A ITEM.SERIAt. HEAT NO. C. Mn. P. S. Si. Ni. Cr. M o. V. 38-001 4,4-2532,3 .25 .90 .010 .012 .21.10 .17 .03
03
~ ULTIMATE YlELD Y. L.htMic'I P T, EloNGATIOIt- -, RE DUCTION % SHEAR IMPACT (7 TENSILE 'A oF (FT. LBS.) LAT.EXP. g STRENGTH U (Pil.1 (P.S.I.) AREA 38-001 77,500 54,750 32.5 74.9 @ +10 F AUSENTITIZED @ 15000y for 3 hrs, QUENCHED TN WATER 27% 51.0 . 0 tl6 TEMPERED G 124CP F for 3 hrs 86% 156.0 .093 HEAT TREAT PROCEDURE HT-60-5732-B, Rev. B of 10/16/?3 70% 147.5 .090 IMPACT PROCEDURE ~ LT-60 5732-0B, Re. B of 12/20/73 FREE OF ME CURY AND RA IOACTIVE CO ITAMINATI0tl P ROUND FLUT D INCOT MOL -FORGING TE$TED' AND I'C PTED IN ACCORDANCE WI'E APPLICALBE ORDER 3PE "\\, AND DRAWINd REQUIREtL.T'S h h . =; Statacf Pennsylvania W:rr:n County 55: N. . Baxte, Jr.- 82f:re rne, a Notary Public in and for above County, personally appeared..... cf the National Forge Company, who being duly Sworn according to Law, deposes and ays that the a ove Report is a true and correct copy of tests as contained in the records of the Company. QdQ..Qt ( scribed and Sworn to ... da y t,............D..e'e...c..m. b. e r..... 19... 7 4 O '/3 ' 7#/ ..s.....1.. 2...t..h..... h5N & )bf t.6.1 n.mn m nr . My Comniission ex pireg.g.. 3.y.j,...,,.7. ..,.,.,.,.g..y,,..g,n,.;.p,.m.a.. (_ (@ O,u. a.td.1,Aes Juay L 1778
FORM NO;10,%A n I NATIONAL FC R E COMPANY l TEST CE TI ICATION DT 6b w .. +f: W. NATIONAL FORGE CO.OROER: 60-5732I. YC CUSTOMER: Alabama Power Company SPECIFICATION: ASME Code Sec. II CUSTOMER OROER NO.: FNP-2-100 R q. #46555-SS-Const.-10531200-SS-1102-68 Table I.l.0 ASME Sec. III ' described on ITEM.SE RI AL NO. DESCRIPTION _: Southern Services Dwg. A-1102-68-2, l 39-002
Unit, Flued Head Penetration Rev. 4 & detailed Spec.." Inquiry No. g.
C5 - P;natration NoO 5 Drawing 4-01810, Rev. B SS-1102-68, Rev. 4" ~ Mk. Q2P15x3/8"CBB 4B ASME-SA-350-LF2 ITEM.SERIA1. HEAT NO. C. Mn. P. S. Si. Ni. Ci Mo. V. 1 39-002 2h:2.23 1.03 .014 014 .23 .12 .17 .05 .04 L= LONGITUDINAL CHA ULTIMATE YlELo TENSILE REDUCTION IMP op % SHEAR (FT. LBS.) LAT. EXP. PT. ELONGATION _. ~ STRENGTH (PS.I.) ,, * (PS.I.) -... ' AREA @ 10 F L 71,500 55,250 33.0 79.4 100% 239.0 .122 43% 112.0 .100 100% ~240.0 .122 AUSENTITIZ ED @ 1500 F for 3 hrs, 0.UENCHED IM WATER 0 TEMPERED G 1230 F, for 3 hrs TESTS SPEC MENS STRESS RELIEVED @ 11250F for 5 hrs, Fijrnance cooled 100 F/1 rtobelowl 0 IlEAT TREAT PROCEDURE HT-60-A-673y-0B,Rev. B of 10/16/73 600 F i IMPACT PRO :EDURE LT-60.A-5732-0B, Rev. B of :.2/20/73 FREE OF ME 1CURY AND RA3I0 ACTIVE COMTAMINATION ROUND FLUT CD INGOT MOLD FORGING TE 3TED AND ACCsPTED IN ACCORDANCE WI'11 APPLICABLE ORDER SPEC. AND DRAWING REQUIREMENTS i p'18UCIl O4 Stat >cf Pennsylvania O o 55: Warren County .i.te{,, .pgg Before me. a Notary Public in and for above County, personally appeared ve Re ortis a true gI cf the National Forge Company, who being duty Sworn according to Law, deposes and says that the a gt) 171 e.o c rrect copy of tests as conta*ned in the records of the Company. ? h
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O T v.,. ( " +bscribed and Sworn {to C 'IE NT f..'......b.. u...n e....... ........ 19...7 4... 10.th'.............. d d o\\ 4....... .8eLN j..i.u.ouw........ g ,110a_ gj, M y C o m mission o pi,e,..... e = v....... ae....,,m. ,.., m,., jg
'.Foplu NO 10m-l~ NATION L FORGE COMPANY TEST CERTIFILO ON D t t .'f t m . w. ea NATION AL FORGE CO. ORDER: 60-5732 CUSTOMER: Alabama Power Company SPECIFICATION: ASME Code, Sec. II ~ CUSTOMER ORDER NO.: FNP-2-100. Table I.l.0 ASME Sec. III described on ~ Rrg. # 46555-SS-Const.-10531200-SS-1102-68 ITEM. SERIAL NO. DESCRIPTION: Southern Services Dwg. A-1102-68-2, Rev. 4 & detailed Spec. " Inquiry,Nati f 40-001 Unit #2 Flued Head Penetration SS-1102-68, Rev. 4" B-4C Drawing 4-01810, Rev.B ASME SA-350 LF2 Mk. 02P15x3/(" Panatration j ITEM. SERIAL HEAT NO. C. Mn. P. S. Si. Ni. Cr. Mo. V. 40-001 ' M 2,53,2,h .25 .90 .010 .012 .21 .10 .17 .03 .03 5 L= LONGITUDINAL ~ ULTIMATE YIELD ChARW REDUCTION IMPACT TENSILE PT % SHEAR (FT. tasa LAT.EXP j
37 EtoNcATIQ3:
oF . STRENGTH (PS.I.) (PS.I.I AREA @ 10 F 40-001 L 79',500 57,500 31.5 72.5 76% 136.0 .100 64% 122.0 - .089 AUSENTITI2 ED @ 1500 F for 3 hrs, QUENCHED IN WATER 0 TEMBERED E 1240 E for.3 hrs HEAT TREAT PROCEDURE IMPACT PRC CEDURE FREE OF ME RCURY AND RLDIOACTIVE CCfTAMINATION FORGING TL STED AND ACC EPTED IN ACCORDANCE WITH APPLICABLE ORDER SPECIFICA AND DRAW 3 G REQUIREMEbTS CONSTg i 5b?fhn z. W -n Stat 2cf Pennsylvania /8/g 55 Wir'.en County g i ....h.....,....D a,?;,t .,b. C Defore me, a Notary Public in and for above County, personally appe ed Y cf the National Forge Company, who being duty Sworn accordin o law depose nd say that the abov pr goo c:rrect copy of tests as contained in the records of the Company. b ...n.- ' " " *;, 5:,. s )f. W y e () bscribed and Sworn to f# g p) this... 2 d. .... day of..... S.9. ..U... P.. 19.'.7.9... g,,,j-1 1.' !d
f....Y......f....
...............?....... f My Commission expires................. ..T.h'. O"f."..h.M! S A listas. Vier <a Cos.m anayfu,i a. New y PJ.tle $,d r ur u l-,
w. m Q74.18.05
7gm n. 7 .g NATIONAL FORGi COMPANY TEST CERTithTLON o
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O .. >g* NATION A1. FORGE CO. ORDER: 60-5732 CUSTOMER: Alabama Power Company, SPECIFICATION: ASME Code, Sec. II CUSTOMER ORDER NO.: INP-2-100' R;q. #tl6555-SS-Const'.-10531200-SS-l102-68 Table I.l.0 ASME Sec. III described on OESCRIPTIONi Southern Services Dwg. A-1102-68-2, ITEM.SERI At. NO. Rev. tt & detailed Spec. " Inquiry No m. 'M" 32-001 Unit 2 Flued Head Penetration SS-1102-68, Rev. 4" Panetration' "1 Drawing 4-01804, Rev. B AsME-SA-182-F 3'Ott Ek 02Ellx10"CCB-32A ITEM,SERIA1. HEAT NO. C. Mn. P. S. Si. Ni. Cr. M o.. V. h ^ 32-001 ' $5632Ar-... ' 6 1.67 .021 .008 _,3.9.,. 9.27 19.49 .2t& .0 L " LONGITUDINAL ULTIMATE YlELD HARDNESS -REDUCTION IMPACT TENSILE =D ELONGATIO_N,, AREA hT OF (FT. LBS.) STRENGTH V
(PS.I.)
^ ~ (P.S.I.) 32-001 L 85,250 t10,500 65.0 81.5 AUSENTITIZED E! 1900 ?, for 3% hrsQUENCHED EN WATER 0 I HEAT TREAT PROCEDURE 4T-60-5732-0 2, Rev. B af ti/10/73 IMPACT PROCEDURE LT-60 -5732-0C, Re f. A, of 6/22/73 FREE OF MEhCURY AND RA 3I0 ACTIVE CO'4TAMINATIO'1 !\\ ./ , s FORGING TESTED AND ACCCPTED IN ACClRDANCE WI CH APPLICA3LE ORDER SPEC. AND DRAWING REQUIREMEN ES / 4 m \\- l State pf P:nnsylvania ) 55: WarrenCountyJ Bef re me, a Notary Public in and for above County, personally appeared.. .N.,,,C. Baxter,,,,,Jr.__ of the National Forge Company, who being duly Sworn according to Law, deposes and says that the abo IN 0/8:h//D.p02,89..o. c* rrect copy of tests as contained in the records of the Company, ?% -.e...<..C ( scribed and,.S$orn is ' '. f. N g 25 ,,,,,,7,,g, g,o. b.... k..,t h,,,,,,,,,,,,,,,,,,,,,,,,,,,Apr.,1,3,.iG-..9tsO..souh>.-... a y& 64 n, Commission e pires. _ m.,._~, m.4..t ~ e.- c.,. _,m.,sz-m,
.1 e m. NATIONAL FORGE COMPANY ' W TEST CERTIFICAT D** 3 n .J, ow w ~ s %.. t!STOMER: Alabama Power Company
NATION AL FORGE CO. ORDER:
60-5732'. DI
USTOMER ORDER NO.;
FNP-2'100 SPECIFICATION: ASME Code, Sec. II ~ R:q. #46555-SS-Codst. 10531200-SS-1102-68 Table I.1.0 ASME Sec. III described on. YEM sERIA1. NO. DESCRIPTION: Southern Services Dwg. A-1102-68-2, Rev. 4 & detailed Spec. " Inquiry Nos;r 33-001 Uni 2 Flued Head Penetration SS-1102-68, Rev. 4" ^.; M T "- Pc.nstration Drawing 4-01805, ASME-SA-182 F 304 ~ Mk Q2Ellx12"ECBuI4 Rev. A - ITEM-SERIAL HEAT NO. C. Mn. P. S. Si. Ni. Cr. Mo. V. -~~ 33-001 ~ 1-3591; .07 1.64 .029.009 48 9.13 19.05 21. j L9 LONGITUDINAL ULTIMATE YLELD HARDNESS REDUCTION IMPACT i TENSILE 24 ELONGATION OF (FT.LBSJ /^T STRENGTH 4 '() ? (P.S.I'.) .(P.S.I.)
1- - AREA 33-001 L 88,750 40,000 74.0
.82.4 AUSENTITIZED @ 19250F, for 2% hrs , QUENCHED EN WATER HEAT TREAT PROCEDURE IT-60-5732-0 :, Rev. B of 4/10/73 IMPACT PROCEDURE LT-60,-5732-0C, Re f. A, of 6/22/73 , 82 FREE OF ME, CURY AND RA)I0 ACTIVE COlTAMINATIO1 / /N./ FORGING T STED AND ACCEPTED IN ACORDANCE lI IH APPLICA3LE ORDER 3\\ AND DRANI G REQ'1REMENIS l / \\? -9.. Stita,ef Pennsylvania 55' Warren County Bef::re me, a Notary Public in and for above County, personally appeared. ...N.,,C,,,,,B,a,x,t,9,g,, Jr. and - cf the National Forge Company, who being duly Sworn according to Law, deposes and :ays that the a a.c d. 0b...., Mhg correct copy of tests as contained in the records of the Company, I'Q 7
cribed and Sworn to s.
April.. 19 7# Q,Q ..ns day oI .e,M.b. My Comme s si on expires...... __..y,,3,.f.g.,,_,.h-76 - - ~ ~ Pr2n Werem Cog.cy. Pcmuylver43. II:'J'Y h hDC N
Y NAT.EONAL FOREE CO.MPANY _ TEST CEI;Tt (CATION e g I%@ 60-5732' N.k. NATIONAL FORGE CO. ORDER: CUSTOMER: ' Alabama Power Company \\ SPECIFICATION: ASME Code, Sec. II CUSTOMER ORDER NO.:.FNP-2-10D Table I.l.0 ASME Sec. III described on Reg. #46555-SS-Const.-10531200-SS-110268 ITEM.SE RI At. NO, DESCRIPTION: Southern Services Dwg. A-1102-68-2, Rev. 4 & detailed Spec. " Inquiry No. 34-001 Unit "2 Flued. Head Penetration SS-1102-68, Rev. 4" Penetration Drawing 4-01806 ASME-SA-182 F 304 ~ Mk. 02Elix1 CB-32B ITEM. SERIAL HEAT NO. C. Mn. P. S. Si. Ni. Cr. Mo. V. 34-(j01 I41-3568[ .06 1.64'.019 .007 .43 8.87 18.67 .22 M ~ ~w ..g ..s :r L9 LONGITUDINAL ULTLMATE YtELD HARONESS TENSILE REQUCTION IMPACT
ST
.D ELONGATION oF (FT. LBS.) STRENGTH .
TPS.I.)
(PS.I.) .- ; -- AREA 3 1L 87,250 42,500 65.0 .78.8 AUSENTITIZED G 1900 F for 4 hrs , QUENCHED [N WATER ~ HEATTREAT PROCEDURE 1 (T-60-5732-0C, Rev. B.c f 10/16/73 IMPACT PROCEDURE LT-,0-5732-0C,F ev. A of E/22/73 FREE OF MIRCURY AND R,DI0 ACTIVE CC NTAMINATICN I ROUND FLUTED INGOT MOI \\D FORGING TESTED AND ACCEPTED IN ACCORDANCE WI DI APPLICA 3LE ORDER SPEC. AND DRAWINd REQUIREMENIS / y / N St;txf Pennsylvania 5" Warrs;n County B;f are me, a Notary Public in and for above County, personally appeared N. C. Baxter,,,j 3J g cf the National Forge Company, who being duty Sworn according to Law, deposes and says that the vI is a true + -O correct copy of tests as contained in the records of the Cornpany. 4..
c. NEE
........ Y ~. f 0 M'neribed an.d b rn to. ' s.. 2b.h.h......... day of[ ' April j g,,,,7,,9,,, Q,,M \\ N N ~ ~ .~ - i --t y m u.tv - My Commission expir'es..
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FO5%4 NO. t056A NATIONAL F3RGE COMPANY 1 TEST TIFICATION D oo m s .-:s-w -~ h3,7.' ~. CUSTOMER: ALABN% POWER COMPANY NATIONAL FORGE CO. ORDER: 60-5732 SP"..:IFICATION:ASME CODE, SEC.II TABL551.1.0 CUSTOMER ORDER NO.: FNP-2-100 AS.'E SEC. III DESCRIBED ON SOUTHERN REQ iW6555-SS-CONST.-10531200-SS-1102-68 ITEM.SERI AL NO. DESCRIPTION: SERVICES DWG. A-1102-68-2, REV. 4 S
4. --
5V.
DETAILED S'PEC. " INQUIRY NO. SC-1102 '68f R 35-003 UNIT 2 FLUED HEAD PENETRATION 4" ASME SA-182-F30f+ PENETRATION 1 DRAWING 4- 01807, REV. A MX Q2E11X12"ECET-13 ITEM. SERIAL * - HEAT NO. C. fan. - P. S. Si. Ni. Cr. M o. V. ~ 5 1-9674 }2.071.61 .023 .006.47 9.10 19.12 .18 35-003 ~: _. j(;g.of' L: LONGITUDINAL ULTIMATE YtELD HARDNESS STREt*GTH ..2b ELONGATIOS~ ~. REDUCTION I ,TEt4SILE fl O f f~y OF (P.S. I.) (P.S.I.) AREA h 35-003 L 89,500 43,500 69.0 80.1 y g5 sEP O < ) AUSENTITIhED @ 1900*F FOR 5 HRS QUENCHED [N IMTER HEAT TREAT PROCEDURE H1 5732-0C, REV. B IMPACT PROCEDURE LT 5732-0C, REV. A OF 6/22/ 73 FREE OF MjRCURY AND RADIOACTIVECONTyINATION ,.! sN, FORGING TESTED AND ACCE PTED IN ACCOR0ANCE WITH APPLICABLE ORDER SPEC, AND DRAWING REQUIREMENT @ ,V ) I 'N. v itatacf P2nnsylvania ) 55 i' S WARREN CO. 7 B1 fora rne, a Notary Public in and for above County, personally appeared... ....N. __C.NO, ER,,,, tig_, ff / 6 cf the National Forge Company, who being duty Sworn according to Law, deposes and says that the above Reportis a true and carrect copy of tests as contained in the records of the Company. dc W ...n..... ribed and Sworn to %. _.....15.Tli.' L a.l.iy of.......MP.IE@ER..__...1D.. 21 _ s.........,... .g..g... ~ ~- ~ -. . ~ ~ ~ ~...... ..,.. j.....,...,
......,.....

...w...rds r. 3-at fay Comrnissioh 4, ites...
tw.sv 2. 1978 n. a i
m.- NATIONA.l.f0RGE COMPAN W f,&., TEEGERTIFICATION Q UX/2?l974 o r 0 0" ALABAMA POWER CO. J. mm o Wc WJ 4 CONSTRUCTION DEPT. FARLEY NUCLEAR PLANTG.i'l. NATIONAL FORGE CO. ORDER: 60-5732 +--- T - NSTOMER: Alabama. Power. Company REQ. #46555-SS-Const.10531200-SS-1102-68
USTOMER ORDER NO.:
FNP-2-100' SPECIFICATION: ASME Code, Sec. II 1. Table I.l.0 ASbE Sec. III described liiEM.SERI AL NO. DESCRIPTION: on Southern Services DWg. A-1102-68-2, 16-001 Unit 2 Flued Head Penetration Rev. 4 & detailed Spec. " Inquiry No., tnatration Drawing 4-01807 SS-1102-68, Rev. 4" W-> ASME-SA-182 F304 ik. Q2E21x3"ECB 1 ITEM. SERIAL HEAT NO. C. Mn. P. S. Si. Ni. Cr. Mo. V. 16-001 { .05 1.73 .019 .005'.43 9.25 19.12.23 ,= LONGITUDINAL ULTIMATE YlELD
HARDNESS TENSlLE REDUCTION IMPACT STRENGTH
.. D ELONGATioM.~.... OF (FT. LDS.) .T , e' (P.S.I.) (P.S.I.) ~ ' AREA 16-001 L Bf,500 43,500 62.5 81,2 \\USENTITIZED @ 1900 F for 2 hrs, QU INDIED IN 1ATER { EAT TREAT PROCEDURE HT-60-A-5732-0 2, Rev. B af 10/16/7 3 [MPACT PROCEDURE;LT-60-5732-0C, Rev. A of 6/22/74 FORGING TESTED AND ACCEPTED IN ACCO H)ANCE WIT I APPLICAB ',E ORDER SPEC. IND DRAWING REQUIREMENTS x / State cf Pennsylvania ),4gyMg ~. r s \\- N. C. B,,,ter, Jd,.,1,,,fkg f trran County ss: 4 BIf tre me, a Notary Public in and for above County, personally appeared cf the National Forge Company, who being duly Sworn according to Law, deposes and says t'the agve R krt h t A O. l.9g ccrrect copy of tests as contained.irith'e r'r$ l,Elrof the Company. YW..S. g %.... d i /..,ey-CONTCg .o)cribed and Sworn to . '.l.*
s,,
G i, - mis......lS..th..... day di.' '... ...Ma.y........... 19.2.8.l... .h M).fAJ.0M S ) 2-1/'M A-My 'C:mmission cxp1ecs.......... .hL::..:.t.:.*. s.. frv aa, Vr re.a (" a e,. i'., n.., M. >, n.,cy sw: 3 IAO Cc.v.s :. t.t f a;,*s. : J.:, wf 2, 3<,73 j
pino..ms.A q [ NATIONAL FC GE COMPANY$}@QT TEST CthFICATION MAY P.71974 e p m 'a-ALAsAMA powE:a co ) 2: 0= CONSTRUCTION DE?T. ']' FARLEY NUCL5A, Ptifdr$[ I
~
NATIONAL FORGE CO. ORDER: 60-5732 USTOMER: Alabama Power, Company- 'I SPECIFICATION: ASME code, Sec. II
USTOMER OROER NO.:
FNP-2-100 Table I.1.0 AstE Sec. III described..on tg. #r4G555-SS-Const.-10531200-SS-1102-68 Southern Services Dwg. A-1102-68-2, lTEM. SERIAL NO. DESCRIPTION: Rev. 4 & detailed Spec. " Inquiry No d s" ~ 37-001 Uni 2 Flued llead Penetration SS-1102-68, Rev. 4" Pc.natration Drawing 4-01809 ASME-SA-182 F30l+ F1K. Q2E21x3"HCB-4 , ITEM. SERIAL HEAT NO. C. Mn, P. S. Si. Ni. Cr. Mo. V. 37-001 1'--3629 .05 1.73 .019 .005..r43 9.25 19.12.23 HARDNESS ULTIMATE YlELO TENSILE REDUCTIO;4 IMPACT pq -.D ELONGATIO2r.'. OF (FT. LBS.) 5 STRE NGTH -k/ I P.S.t.) (P.S.l.) AREA 37-001 L 85,500 39,000 65.0 80.0 AUSENTITIZE D @ 19000F for 1 hr, QUENCIED IN MATER FORGING TESTED AND ACCEPTED IN ACCORDANCE WIT:1 APPLICABE,E ORDER SPEC. AND DRAWINC REOUIREMENIS IIEAT TREAT ?ROCEDURE !!T.-60-A-5732-0C, Rev. B of 10/16/7:4 IMPACT PROC DURE LT-60-A-5732-0C, Rbv. A of 6/22/73 gGIIbIOUg7 State ef Pennsylvania cYf,. Warren County 55: N. C. Baxter 82 fore me, a Notary Publicin and for above County, personally appeared....... -n cf the National Forge Company who being duty Sworn according to l_aw, deposes and says that thea ove f, grt is a tr dp D/., (orrect copy of tests as conteined in the records of the Company. ,,T '/:y, ...s...<....d. 6 W Q.r.w,. .scridcd and'swern io Q4 %. co8 s 2 May ,,,,,,,,,,,,,,3 g,,,7,11,, this...... 2. U.h.'....... da o f... t .ORuku'.................... 5)14l19 \\ My Comm.. s s. on c x pu e s................:...._uu..-.s.. u.....- ,a.,..... c...,.,......... .c.... 8 n MoL
("Y NATIONAI fdR E COMPANY \\ TEST CERT Fit ATION rL(*3 lD . M. W 1." 303' }/AL 2 e o .v .t 60-5732'.g-NATIONAL. FORGE CO. ORDER: +
USTOMER

Alabama Power Company USTOMER ORDER NO.: FNP-2-100 SPECIFICATION: ASFE Code, Sec.'II Table I.1.0 ASME Sec. III described on rR:q. 46555-SS-Const.10531200-SS-1102-68 TEM. SERIAL NO. DESCRIPTION: Southern Services Dwg. A-1102-68-2, Rev. 4 & detailed Spec. " Inquiry No. 41-001 Unit Flued Head Penetrabion SS-1102-68, Rev. 4" ' M' JCnatration Drawing 4-01811, Rev. B ASMS SA-182 F304 Mk. Q2P15x3 "CCB-54 ITEMsSERIA1. HEAT NO. C. Mn. P. S. Si. Ni. Cr. M o. V. 41-001 $359O . 07 -1.64 .029 .009 .48 9.'13 19.05 2.1 ~ ~ LvLONGITUDINAI' ~ ULTIMAT E YlELD HARDNESS. TENSILE REDUCTION IMPACT TEST
2f STRENGTH
/3 ELONGATION oF (FT. LBS.) AP.S.I.) (P S.t.) AFtEA 4 L 87,500 42,500 65.0
79. 3 s.
AUSENTITIZED @ 1925 F for 2 hrs , QUENCHED EN WATER 'N llEAT TREAT PROCEDURE IT-60-5732-0 2, Rev. B of 5/7/73 k ^ IMPACT PROCEDURE LT-60 -5732-0C, Re i. A, of 6/22/73 / b FREE OF ME.lCURY AND RA3I0 ACTIVE C0 4TAMINATIO1 s N FORGING TEjSTED AND ACCEPTED IN ACCORDANCE WI EH APPLICA 3LE ORDER 3PEC. IRIP/Pl? MY/@ LIM'l M AND DRANIhG REQUIREMEN ES State of Pennsylvania c7 h-n 55: Warren County ., Bcfore me a Notary Public in and for above County, siersonally appeared.. . __B_ x t, r,,,,,JpEp,,,,,, a e 6 s that the e Report is a true ancf cf the National Forge Company, who being duty Sworn according to Las c:rrect copy of tests as contained in the records of the Company. h T..CDk@ N L.c ?s.. m.- .~, ', t bscribed and Sworn to '.
7. #
( 1 8' . : ce,Pof.'.... August. 19... 7...tl.. 20th f 0 Ql.}.04 M.......... ,g My Comminion expires..... i~w. w. - <......x. :....,,:.,,,,.3,,, 3,,, g ,. Sf' s Hi cu..-.
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NA270NAL FO. E COMPANY
W TEST CERTI IC, TION L
( ) 60-5732" ih-NATIONAL. FORGE CO. ORDER: pSTOMER: Alabama Power Company l USTOMER ORDEF1 NO.: FNP-2-100 SPECIFICATION: ASME Code, Sec. II Table I.l.0 ASME Sec. III described on - teg. 46555-SS-Const.-10531203-SS-1102-68 TM.SERI AL NO. DESCRIPTION : Southern Services Dwg. A-1102-68-2, Rev. 4 & detailed Spec. " Inquiry No. - 42-002 Uni Flued Head Penetration SS-1102-68, Rev. 4" s.. -L.. m 4-01811, Rev. B S , / # ~ 5/mdel (L. na pl5 g fg "A-182-F304CC8-s's per b 3 0F3n fM benntration - Drawing Mk Q2P15x3 55 # #2 kee. um w.wa. it o trf. u 4 w, %, s ga.esuave,an 9 6 Si. Ni. Cr. FAo. . V. 7 6W-77 C. r P. ITEM-SERIAL HEAT NO. 6 53090;.J.b8 1.74 .033 .009 .44 10.35 19.08. 19 12-002 ~
LONGITUDINAL ULTIMATE YlELo HARDNESS TENSit.E REDUCTION IMPACT egg.
2or STRENGTH /3 ELONGATION OF (FT.1.83.1 [3 \\ ! (PS.I.) . (PS.I.) =: -- - AREA x.- 42- 02 L 87,750 43,000 78'.9 68.5 S.e@S LUSENTITIZED @ 19250F, for 2 hrs , QUENCHED EN WATER IEAT TREAT PROCEDURE IT-60-5732-0 :, Rev. B dtd 10/16/i3 p 7,1 p [MPACT PR DURE LT-60 -5732-0C, Re i. A, of 6/22/73 i FREE OF CURY AND RA3IOACTIVE C0:1TAMINATIO:4 1 ON-s., 20RGING TESTED AND ~ ACCEPTED IN ACCORDANCE WI GI APPLICA 3LE ORDER 3PEC. AND DRAWII/G REQUIREMENIS tatecf P;nnsylvania $5' farren County B: fora me, a Notary Publicin and for above County, personally appeared. ...,.l.l C,:,,_ Ba,, ter,_Jr.,_,,,;,,,,,__ of thi National Forge Company, who being duty Sworn according to Law, depose nd say hat the a ve Reportis a true and / carrIct copy of tests as contained in the records of the Company. .~.. ef...C..C...c ' '[$. f.e,.,,, / O. bed and Sworn to 25 ri s ...... 9k @ day o't.~ d..nu,ay,g,,,,,,',,,,,,,,,j g,,7 5, ...S !.e. A._. ]>]/ l df v f.............T81 \\ 1'. ~ f Y* ~ ' " 7 - E~' ~ [ g/ Ay Commission expires r,. f.......,y p.3,3 g, ,.,, c u..a.: u,.:.. n,.x,, 2. i,g . p t.+, w,..... e.
unm y. w r NATIONAL FOFGH COMPANY TEST CERTI. r ATION h d 4y 7 1 .y-- NATIONAt. FORGE CO. ORDER: 60-5732'
tJSTOMER: Alabama Powsr Company
~ SPECIFICATION: ASME Code, Sec. II
USTOMER ORDER NO.:
FNP-2-100 Rxq. #41723-SS-Const.-10734800-SS-1102-68 Table I.l.0 ASME Sec. III described on ITEM.SE RI Al. NO. DESCRIPTION: Southern Services Dwg. A-1102-68-2, 43-002 Unit Flued Head Penetration Rev. 4 & detailed Spec. " Inquiry No. E'.: P. natration @2 Drawing 4-01812, Rev. B SS-1102-68, Rev. 4" MK. Q2P15X3/8CCB.56 ASME-SA-182 F304 -= ITEM SERIAt. HEAT NO. C. Mn. P. S. Si. Ni. Cr. Mo. V. 43-002 ( .06 1.67 .021 .008.39 9.27 19.49.24 L= LONGITUDINAL ULTIMATE YIELD HARDNESS TENSILE -
2%
REDUCTION IMPACT --*T /s STRENGTH ELONGATION. oF (FT. LBS.) (/ ' lPS.t.) (P.S.I.) ~ ~ ~.. AREA 43-002 90,000 43,000 68.0 80.0 AUSENTITIZED @ 19000F for 2 hrs, QUENCHED Ib WATER HEAT TREAI PROCEDURE 1(T-60-A-5732-0C, Rev. I, dtd 10/16/73 IMPACI PRC CEDURE LT-6C -A-5732-0C, Rev. A, dt'd.6/22/73 FREE OF ME RCURY AND RADI0 ACTIVE CCNTAMINATICN. FORGING TE STED AND ACCEPTED IN ACCORDANCE WITH APPLICI BLE ORDER S C. AND I RAWING REQU,IREMED S 3 A mb Aa n YN7 7- / T. 5 tit 2cf Pennsylvan.ia 55 Warren County B;fera me. a Notary Public in and for above County, personally appeared... R.,S. New. 1 ,_,,,,,,,,},_ cf tha National Forge Company, who being duly Sworn according to Law, deposes d says that t above Reportis a true and c rrect copy of tests as contained in the records of the Company. cribed and Sworn to .... 0.?............ day of.. -....... * ~. ~ ~ ..N A Y.L9d--.~~~~ My Commbsion expires. - % < A H-- ---4 w ~~~ ~ s t.*ee, 'tr c o n... .:... e t e, hya, 3,,t h j f o $~~ ., *.i 8., g_ & t g3
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MILL HEAT NUMBER i enetration i Sleeve Door Ring Door Face 84
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802J14990 y 802J14990 80lJ13870 Equipment 801K02660 Hatch i A l i O I e d ] O b 4c. .*e--A+ .-*-w -e-w-..r+ww- + - .-w- -- *.,., + - - vv. .- _.r - w ar., .-~.s., .,w,e.w._,y.. ,..~,-,7.-,,,,~m,,,,.y,..p,, ..,..ww.,v,.,.,,3.,,,,,
BETHLEHEl[ ST PORATION - 3F737sIF[O) r'. ./ S METALLUte ARTMENT a ( ) REPORT OF T .4ND ANAL.YS15-L an: smrats o. DAf(sNfM) CAR of VEWCd t40. \\ ntW.N S H A RP.t)R 803-06636 03-16-73 PC LOUISVILLE LN . Pl.E 000734 PAGE 2 CHICAGO BRIDGE C IRON CD CHICAGO BRIDGE G IRON CD E BOYLES AL 00X 277 [{ BIRMINGHAM AL 35202 j -Q #%O 5'z a cuaxmv // y um ''c'f38*g,' e'"ol7o" " ,te,c, staat wtAt statuera statucia is. s ,,, j y,w wune muu g,,,,, l w;,,,,,,,, ,l t,,,,, w,,,,, esco enorn wo. REV O' QAS-321 REV O ASME SA516 GRI70 PVQ C ASME SECT 3 LONG CHARPY V NOTC'EH IMP ACT. OF 2'O FT LB AlT MINUS Sd DEG F PER IEEL PLATES-MS-66'.1 NB 23'00 ' '~)20 63-2 SHE ET 51 GH 025'26680 / I I K 10231-01'B01J13870 1 1-3/ B 124-1/ 2 462 22t:06 4 00 75 00 8' O .Le m M v.u [h ( c==== LLD: n: $s o, n. a Tbn 2Eso @r4 (?f-at'c7+5ettheregn4*cmenta V tho spocification numbora shown horcon hav boon 60t. j Pl.ATES NORMAbl ZED /AND STAMPED MT / O<9e/O 4-dddf/ / / / / / CHEMICAL ANALY$!$ f2/[.g [ f,gg ca ,gn r/ s/ si f c. m c, u. v u c/ W ~ rs 31J13870J .1/2 L.1'2.009'. 029 .48 SUBSCRIBED AND SWORN-TO'BEFORE ME h<. THIS /[ DAY Cff..-h - 1928 . ;.7 : ggg q M Y ' l U D L.10 p POPTER COUNTY INDIANA MY C01th!ISS10N' EXPIRES 11AY 10.1075 h P m. & O gg,/g c. , tw i n rs;fr the abov,,,,,N rh N ermro m crtfTr0 lb rDo M c f (Do Pethlehem Steel Cornflos..
h, k'.. f~ = - - &g., RECORD OF HEAT TREATMENT g i, gg.,% c =1 . CUSTOMER: Chic.ago.Bridga.and Iron Company. .l ' ' 'I '. ' '.' PAdE ' 2.1TTACH$ENT ~ .* SHIPMENT.NO. 803-06636 ~ DATE SHIPPED 3/16/73 gg - ~ PI. ATE'NORxALIZING CYCLE unE g=@ . FmaacE g stR14t wuxstR + rtxr oP.; ...m. _utir vuxstR _C00t1NG g x 10231 m 8.0ia1387 ,g: . 16 f16y <..,,7;, .;ggccot f.. '. : . ;:br. 3,;.. a. s.. n. -o .p- ..... o. ...(y. a. i. . q-.< ,=,...s. -.m n ,/ s,- ,e p. N
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IMPACT PROPERTIES .i.. s .\\ PAGE 2 ITTACHMENTl - ^ CUSTOMER: Chicag'o Bridge and. Iron Company. 't a . SHIPMENT NO.803-06036 . i: . DATE SHIPPED 3/16/ ~ c. -50 0 LONGITUDINAL'CHARPY V-NOTCH TESTED 0 F_ ./ : 5. /; y-SERIAL NUMBER ., HEAT NUMRER, .. CHARPY SIZE - FT.4 B % DUCTIL, FRACTURE AREA LATERAL EXPA dION-(IN)~ -t' 10231(1) 801d13870 FULL' .27-37-42 40-33-55.- .7 .o.pr 035;;0,J ' c:s ,O =. - ,..7 .y O .y,,. .. C
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I ..S. b ueTAttch .UAtmENT m BETHLEHEM ORPORATION ,,, k m 4i. 1 ( REPORT OF W AND ANALYSIS W r A sn, tur no. o.n sarna ca os vtaca no. rwn BURNS HARBOR 807--07600 01-?A-79 FrK NTI t FR 701 n non17 DArp CHICAGO BRIDGE & IRON CD CHICAGO BRIDGE IRON CO o y r. Z BOX 277 'BOYLES AL E BIRMINGHAM AL.35202 r o ./ sin a curwisty - U V wan w ss ,1,1, ,,gg,t,
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su== 1,..iv., u-u-uum 1 wm - o;.. i.... w.i.. O ASME SA516 GP 70 PVQ C ASME 5,ECT 3 LON G I ,co. REV;0 FT. LB AT MINUS 50 5 TEEL PLATES-MS-661 REV 0 ;C OAS-321 g DEG F. PER NB 2300 OF 2 CHARP.Y V NOTCH IMPACT 2065-2 SHEET 51! GH 025J2668C 1~$8 61000(/ 7880~0' 27 di ' I. r-K 11576-01802J14990; 1 3 63 373 19.972 B 81000 <2 28?:dKfi T 61200 l ./ t K-11919-01802J14990', 1 3 j 63 373 19972, B 59800', 81700' 2 28':0F [ ' T 59400 /' 81790' 2J&Op I h .b. b:@, T l l {$dk l + i I l l i l I j I t --r-p r t ify p_ Liw icsiac.;nt; obl.ho specification numbers shoen horcon have toen ect. 4/>V6 d-ddd M PLATES OUENCHED AND TEMPERED. O STAMPED MT A //.o f/ carnicAt Asetysis ( "unu e/
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r/ s ss c. ni e, u. v sn @02J14990 . 2'3 1.0'8 .0 66. 0W .f4 SUBSCRIDED AND SWOIUJ TO DEFORE ME PJ b 'HII3.'3 DNI CF.U' ' '.s ,3g 73 3dov KY 1*UCl.1C I'ORTER - COUNTY INDIANA i / hlY CO!! MISSION UXRIRES11AY 10.1075 (,f m.% e~l-[% c. 1 ._..,..._.s..,..... ~ ~, ~. ~. ........ ~
~ , f,. O RECORD OF HEA A REATMENT U' CUSTOMER: Chicago Bridge & Iron Company PAGE1 ATTACHMENT SHIPMENT NO. 803-07600 DATE SHIPPED 3/28/73 HARDENING CYCLE TEMPERING CYCLE STRESS RELIEVING CYCLE ' , RATE OF RATE OF 'F/HR TEMPr?F)f, TIME COOLING . HEATING HOLDING'. (MIN) / *F FURNACE TIME QUENCHING FURNACE TIME K 11576(1) 802J14990 1655/ 660 202 ~ (TYPE) TEMP 'F_. _(MIN)_ COOLING SERIAL NO. HEAT NUMBER TEMP 'F/(MIN)_ t$rTER 1290 /, 192 AIR (.115/125 1150 ( 90 5/160 , AIR JL 115/1W'115h 9 C %160 1290/ 147 X 11919(1) 802J14990 1645 170 WpER -
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r\\u'R O ~~ O O kMPACTPROPERTIES, ) CUSTOMER: Chicago Bridge and Iron Company PAGE 1 ATTACHMENT . SHIPMENT No. 803-07F00 DATE. SHIPPED 3/28/73 LONGITUDINAL CHARPY V-notch TESTED G x(E OF SERIAL NUMBER HEAT NUMBER CHARPY SIZE FT,-LBS. ' % DUCTILE FRACTURE AREA LATERAL EXPANSION (IN) I;11576(1) 802dit 990 FULL (d-7FF./, 'J 29.051.0f M 05E, t .0! 9.d5J.d.05.2 - 6 K 11919C1) 802J1t 990 FULL E14>9 6 29- -29 t _J i=26 c, g n ~
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.n access teen for. eenerat. le d. stoned and eenstreet.d under the rutos of f eetten ff, af .h. me cod. f o r M u CLl e gg.__,,, i ,, J. Va..ete. This form do.e not cow.r th. tel,lar whe reby the.1gst,,,t a att.a%.A.. .s. .b containment v...et, fste wate el pi (en.14. ap.a d.,. .# i.. a a.h-s.=. E m. e.1 I 8'u o. d.4 M T .d ,a If P.. w a -. -i,..-..== We ee.eify th s the.eueme.es.de in this rep.es.re e.e'ect e.eia i..,-=-..>...,,u. ain. ad thu shi. n.cl... e...el conf.c= . rut.. [ coa. ores. ti.e of sh. ASME Code secni.a tit. /. ( ) p'* ou, t - 7_ i,He si, d cereaco sesocs smos e0.,m u .J. ^ <= w.e--> y 8,
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.ir.in,he., 1. Mf g by Chacaos Bridge & Iron Co. Alabama Year CG&l Contrac 71 - 2 M 5 -3 I *t . Mfd foe Daniel Construction company For Alabama Powe r company - ~ 2. s.. :. . g..,c.. .+.. raiocatio,,a e e~s, m a,.
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a,.d * .i 72-2 la accordance with ASME code section TTT Pub-geetime M*. 1971*Ntaten w1*k .g.r re k n f erren.nt of veneel open1Q Addenda thru Summer. 71. No consideration has been given to 'e, (B) Detah. Construction. 8e Inspectaoes in accordance with e raw i a er a 7o .n4 P1 w%4e% * *
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54 t SAS16=70 6. Matis: Material Charpy trnpact 20 eg.tp p Mp 'tg. . 'F ~ Material Charpy impact f -t> @ t Material Charpy Irnpact fr-fp 9 'F 7. Heat Treatment PWHT all category *D* joints and all butt welded'jotests over 1 1/2* thick,. Nor>CatM 8. Joint Category; A C D . J0'"15 rutt Palt rtilet weld on Attachmento on1. autt chi. autt Penetration Penetration within 4* of norren - Type .,. 4 - 100% RT 100% R7 M7 NT MT T q Esandanation Equipment door in steel liner for concrete nuelear conta[nuent essee1 9 Part $ervce:
10. RMs. This certification covers shop work eniv. Putt velde in 2 f e%fe* insert 7ehen plates are to be 1004 radiographed. Attachaeot welds of anchorage parts to the equipment door and insert are covered by thle certifteation. Ancherede naren av.
+ covered only to the extent of using weldable grade piatorial. ri.id we gd of As sembly 70-C. ..E'. to the 1/2" thick liner plate and field weld of 70- A to 71-A are not covered by this I! certification. 4 II. Dirnens onal checks are limited to those required by the ASME Code.6ection 14. Class Mcfor Cod. veneas, ~
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g ono.lsmo otseniPrion 3:d Carbon S*.cci, ASME SA-105 -71 as Modified by 209A4373P003 - ij ~ / / Hea : Treat Charts att3ched. / / 9,1.iquid Quenched Before Temperlitg. Roiigh"Maclitried'. ' ~ ""j'.G'rY [9'iNpM'T"ETI'6'13[12 " _SCR.80 i 'I.0T#15626,,C)F, J' petic Particle Insp6Etied In Accord, With Para.'. 2545 of HE Item..L1 _ yton_'TTT ASME loil.cr# '.F_ressure Vessel Code 1971 Edition by - ' P.0d_282-Y-7.438 ASMESA-105-71,I'_Q&T,2,_E.'209A4373P003, d"*.*dTI.p.Jp0(Rcv _ {,,(8/31/72) &, Addendum I,(4/27/l73), Spec IE Boiler & Pr, css,ure_Ve,ssel Code Section.II &.III'.(1971)' on Magnetic Particle Inspection Satisfactory, Conam Inspection s ,./ ddeni a Through S,ummer 1973 Class 2 Components. ___.Inc. ' Certificati0nh {ttached..__.lheir_. Job __No..C /. l ' / //d O 11 // P}es"12"Sch. 80 Pipe, 13" long I.
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a s - REACTOR SYSTEMS BRAfiCH QUESTION 210.2 REQUEST: t I Wi.en shutting down or starting up the plant, certain automatic safety injection signals are blocked to preclude unaanted actuation of these systems. Describe the alarms available to alert the operator to a failure in the primary or secondary system during this phase of operation and the time frame available to mitigate the consequences of such an accident. Justify the time frame available. The following scenarios should be in. cluded in your discussion: a. In the event of a LOCA following the closure and power lockout of the accumulator isolation valve during plant shutdown, what equipment and procedures would mi tigate the consequences. Provide an analysis showing that the consequences of such an event would be less limiting than the design basis LOCA. b. Discuss the scenario and consequences of a moderate energy pipe break in the RHR system immediately af ter initiating RHR operation while shutting down. Identify ins trumenta tion that will provide alarms for operator actions. Describe opera tor actions required to terminate the event and, using an event time table, show the time available (af ter receiving the alarm) to perform the action (s). Identify the consequences of the occurrences. m v w RESP 0f1SE: a. During the shutdown the following operator actions pertain to the isolation of ECCS equipment and would effect a LOCA during the time accumulator isolation valves are closed with power locked out. (Start-up is not addressed since shutdown is more limiting due to the higher core decay heat generation). (i) . At 1900 psig, the operator is instructed to manually block the automatic safety injection (SI) signal. This action disarms the SI signals from the pressurizer level and pressure transmitters along with the steam flow transmitters. ^ 1 other SI signals, including contain-ment high press re and high steamline differential pressure, are armed and will actuate safety injection if their set-points are exceeded. Manual SI actuation is also available. (ii) At 1000 psig and 425 F, the operator closes and locks out the SI accumulator isolation valves. He also locks out and tags one high head charging pump. At this time, two ) Residual Heat Removal pumps (LH safety injection) would be available from either automatic or manual SI actuation. L" (iii) At less than 400 psig and 350 F, the operator aligns the Residual Heat Removal (RHR) system suction to the Reactor l Coolant System. The valves in the line from the Refueling Water Storage Tank (RWST) are closed. ,= w a =-.--- d
~. Reactor Systems Branch Response to Question 210.2 (Cont'd) Page 2 The significance of these actions on the mitigation of a LOCA when power is locked out to the isolation valves is that: (1) Between 1000 psig and 400 psig, a portion of the ECCS may be actuated automatically on containment high pressure or high steamline differential pressure sig-nals or manually by the operator. The equipment that can be energized are two RHR and one high head charging pumps. Subsequently, the operator would reinstitute power at the motor control centers to the other high head charging pump and the accumulator isolation valves. (ii) Below 400 psig, the system is in the RHR cooling mode. The operator would realign the RHR system per plant emergency procedure, as the RHR and the high head charging pumps could still be initiated by an auto-matic high containment pressure signal, or by manual actua tion. Subsequently the operator would reinstitute power at the motor control centers to the other high head charging pump and the accumulator isolation valves. Safety Significance During Shutdown: Comparing plant cooldown and heatup, the limiting case for a LOCA w'ould be during a plant cooldown rather than a plant heatup because the core decay heat generation would be higher. The ECCS analysis presented in the Farley FSAR conforms to the Acceptance Criteria of 10 CFR 50.46 so that initiation of the LOCA is at 102% of full licensed power rating and corresponding RCS conditions. Some of the reasons why the analysis presented in the Farley FSAR would be more limiting than LOCA during shutdown are: (1) a LOCA initiated during shutdown would have reduced decay heat generation since the reactor, in general, would have been at zero power for an extended period of time; (2) the core-stored energy during shutdown would be reduced due to the RCS uniform temperature condition at a reduced temp-erature; and, (3) the energy content of the RCS would be lower. Furthermore, the probability of the occurrence of a LOCA during this period along with the critical flaw size needed to rupture the RCS piping at reduced pressure clearly indicates that a LOCA is considered to be incredible. These arguments are provided in the following sections. a
Reactor Systems Branch R:sponse to Qu stion 210.2 (Cont'd) Page 3 (1) Between 1000 psig and 400 psig: For the purpose of calcu-lating the probability of a LOCA, a conservative time of 7 hours is assumed to cool the plant from 5000F to 3500F. The annual probabilities of small and large LOCA were estimated at 10-3 and 10-4 per year in WASH-1400.* Assuming this same failure rate holds at reduced pressure (this assumption is not realistic since normal operation serves as a proof test for lower pressure operating modes as discussed later), the probability of a LOCA during heatup/cocidown periods (assuming two heatup/cooldown cycles per year) would be: Small LOCA 3.2 x 10-6/yr. Large LOCA 3.2 x 10~7/yr. These can be compared to the total meltdown probabilities for small LOCA and large LOCA initiating events analyzed in WASH-1400: Small LOCA 2 x 10-5/yr. Large LOCA 3 x 10-6/yr. L-Therefore, even if there was no pipe rupture protection v i for these heatup/cooldown periods, it is concluded that such events add only a small increase to the meltdown risk due to the short time periods involved.
~~WASH-1400, " Reactor Safety Study", U.S. NRC, October,1975.~~
(ii) Rupture of RCS piping at reduced pressure: Below 1000 psig, RCS piping rupture is considered incredible under these low pressure conditions since normal operation serves as a proof test against rupture. Calculations of critical flaw size for the reactor coolant piping show that at 1000 psi internal pressure: 1. Rupture cannot occur for a part through-wall flaw regardless of orientation. 2. For a circumferential through-wall flaw, a catastrophic rupture is not possible. 3. For a through-wall longitudinal flaw, the criti-cal flaw size is in excess of 70 inches. A ,nu
R: actor Systems Branch Response to Question 210.2 (Cont'd) Page 4 Therefore, postulated RCS piping flaws of critical size for internal pressure below 1000 psig cannot exist since they would have previously failed at the normal operating pressure (2235 psig). (iii) Below 400 psig: After several hours into the cooldown pro-cedure (a minimum time is approximately 4 hours) when the RCS pressure and temperature have decreased to 400 psig and g 350 F, the RHR system is placed in operation. This system has a 600 psig design pressure and rupture of this system is also considered highly unlikely. However, the proof l test argument given above for RCS piping does not apply to the piping in this system. The provisions to isolate these lines and the ECCS capa-bility for core cooling should a leak or rupture develop during this mode of operation are as follows. Any leakage of the RHR system piping would be expected to occur when the system is initially pressurized at 400 psig. The RCS is at this time under manual control by the reactor operator. The reactor operator is monitoring the pressur-izer level and the RCS loop pressure so that any significant leakage from the RHR system would be immediately detected. i When leakage is detected, then the operator would isolate j l the RHR system and identify the location and cause. Since the decay heat generation four hours after shutdown is about 1.2% of full power, the PCS fluid temperature is at about 0 350 F and the core stored energy is essentially removed, l the operator would have ample time to isolate the RHR loop. l Therefore, in spite of the low probability of occurrence and the fact that f-certain failure modes for pipe rupture do not exist during cooldown at a. RCS pressure o'f 1000 psig, the plant operation procedures are as follows: j 1. At 1000 psig, the ogerator will maintain pressure and proceed to cool down the RCS to 425 F. 2. At 1000 psig and 425 F, the operator will close and lock out the accumulator isolation valves. The above plant operating procedures will ensure that the accumulator isola-i tion valves will not be locked out prior to about'2-1/2 hours after reactor o shutdown for a cooldown rate of 50 F/hr. I A conservative analysis has detennined that the peak clad temperature result-ing from a large break LOCA would be significantly less than the 22000F Acceptance Criteria limit using the ECCS equipment available 2-1/2 hours af ter reactor shutdown. e i
Reactor Systems Branch Response to Question 210.2 (Cont'd) Page 5 The following assumptions were used in the analysis: 1. The RCS fluid is isothermal at a temperature of 425 F and a pressure of 1000 psig. 2. The core and metal sensible heat above 425 F has been removed. 3. The hot spot occurs at the core midplane. 4. The peak fuel heat generation during full power operation of 12.35 Kw/Ft (102% of 12.11 Kw/Ft) will be used to calculate adiabatic heatup. 5. At 2-1/2 hours using decay heat in conformance with Appendix K of
~~10 CFR 50, the peak heat generation rate is 0.167 Kw/Ft.~~
6. As previously noted in the original response, two low head SI pumps and one high head charging pump are available from either manual SI actuation or automatic actuation by the containment HI-l signal. However, for this analysis the loss of one low head safety injection pump was assumed. 7. No liquid water is present in the reactor vessel at the end of the blow-down. p m 8. A large cold. leg break is considered. For a postulated LOCA at the cooldown condition of 1000 psig, previous calcula-tions show that the clad does not heat up above its initial temperature during blowdown. Proceeding from the end of blowdown and assumjng adiabatic heatup of the fuel and clad at the hot spot, an increase of 682 F was calculated during the lower plenum refill transient of 142.2 seconds. During reflood, the core and downcomer water levels rise together until steam generation in the core becomes sufficient to inhibit the reflooding rate. At that time, heat transfer from the clad at the hot spot to the steam boiloff and entrained water will commence. This heat removal process will continue as the water level in the core rises while the downcomer is being filled with safety injection water. The reflood transient was evaluated by considering two bounding cases. 1. Downcomer and core levels rise at the same rate. No cooling due to steam boiloff is considered at the hot spot. Quenching of the hot spot occurs when the core water level reaches the core midplane. 2. Core reflooding is delayed until the SI pumps have completely ' filled the downcomer. No cooling due to steam boiloff is considered at the hot spot until the downcomer is filled. The full downcomer situation may then be compared with the results of the ECCS analysis for Joseph M. Farley Nuclear Plant-Unit 2 to obtain a bounding clad temp-A- erature rise thereafter.
O ' Reactor Systems Branch Response. to Question 210.2 (Cont'd) Page 6 For Case 1 described on the preceding page, the water level reaches the core midplane 75.4 seconds af ter bottom of core recovery. The tempergture rise during reflood at the hot spot from adiabatic heatug is 362 F, which results in a peak clad temperature of approximately 1470 F. For Case 2, the delay due to downcomer filling is 100.0 seconds. The corgesponding temperature rise at the hot spot from adiabatic heatup is 480 F, which gives a hot spot clad temperature of 15880F. The clad temperatures at the time when the downcomer has filled for the - 0.4 submitted to satisfy 10 CFR 50.46 requirements are 1965.6 F DECLG, C and 215894 F at the 6.0 and 7.5 foot elevations, respectively. Co're reflooding in the shutdown case under consideration will be more rapid from this point on due to less steam generation at the lower core power level in effect; decay heat input at any given elevation is less in the shutdown case. The combination of more rapid reflooding and lower power in the fuel ensures that the clad temperature rise during reflood will be less for the shutdown case than for the design basis case. b. 7he occurrence of a break in a Residual Heat Ree "al (RHR) line during a l nonnal shutdown or heatup when in the RHR mode at, ieration is considered highly unlikely.. For example, the Reactor Coolant System (RCS) pressure in that mode would be 400 psig compared to a RHR design pressure of l 600 psig. However, such a postulated event has been addressed and dis-cussed below. The analysis has been performed for a postulated RHR moderate energy line break during shutdown. The analysis is conservatively based on the break occurring githin four hours after reactor shutdown, the RCS at 400 psig and '350 F, and the pressurizer level at 21.4%. An assessment was also made to determine the equipment necessary to mitigate an RHR l line break to ensure core covery. L The analysis has shown that the operator has 40 minutes (Refer to Table l 210.2-1) af ter the initial alarm to take any appropriate action to ensure r core covery. The analysis further established that (1) one charging /SI i 1 pump will provide adequate flow to sustain the system in a safe condi-l tion and (2) an initial alarm signal low pressurizer level deviation alarm - at 16.4% (5% below zero power programned level of 21.4%) will occur within 17 seconds of the event initiation, followed by another alarm (low level heater cutoff) at 15% and then another alarm (low level SI trip set point) at 5%. The analysis conservatively assessed the largest RHR line which could adversely impact both RHR trains simultaneous-ly. The break area was developed consistent with goderate energy line break criteria and was established to be 0.0152 ft. Results of the analysis confirm that the required make-up can be provided by the i in-service charging /SI pump.
Reactor Systems Branch - Response to Question 210.2 (Cont'd) Page 7 Additional conservatism applied to the above analysis would be excessive. However, even if (1) a 10-minute delay time for operation action and (2) a single failure were assumed, they would not result in an unsafe condition. Specifically,30 minutes should still remain available for initiation and effective operation of necessary equipment. Moreover, it is only if the single failure assumption is invoked that operator action to start the back-up charging /SI pump would be necessary. The operator can initiate starting of the pump from within the main control room and flow can be established within one minute. Primary coolant loss through the break will lower level in the reactor vessel to the hot leg nozzle elevation, assuming no charging /SI pump flow, at 30 minutes from break initiation. Start of charging /SI pump flow in 11 minutes (i.e. > 10 minutes) will delay that time and level will stabilize at the hot leg nozzle level until the break is isolated. Following isolation of the break, original pressurizer level will be reestablished within 72 minutes. The operator, from within the main control room, can initiate closure of the RHR isolation valves and closure will occur within one minute. In conclusion, the conservative assessment of the RHR line break confirms ' the ample margins available in retaining the core in a safe condition. %I w 7 i 'b
Reactor Systems Branch Response to Question 210.2 (Cont'd) Page 8 TABLE 210.2-1 Event (Assuming No Operator Action) Time i RHR Line Break 0.0 (at approximately 4 hours into shutdown) Alarm Annunciation 16.5 seconds ] End of Blowdown to Bottom 1801.7 seconds or of Hot Leg Nozzles 30 minutes ' Beginning of Core Uncovery 2471.12 seconds or -~ 41.2 minutes l I l Operator action (i.e. charging /SI pump initiated and/or RHR line ] isolation) must occur prior to 41.2 minutes. ,O s
v . = REACTOR SYSTEMS BRANCH QUESTION 201.3' The Regulatory Requirements Review Committee, in a memorandum from E. Case, Committee Chairma_n, to L.LCossick', Cxecutive Dirtetor for Operations (dated February 16, 1978), has approved a new staff. position (BTP RSB 5-1) for ' the Residual Heat Removal System (RilR). The_ technical requirements for your j ' plant are described below. Please respond to these requirements in suf-ficient detail to enable the staff to reviev "our compliance in an exped-7 itious f ashion. ^ REQUEST : 1. Provide safety-grade steam generator dump valves, operators, air and power supplies which meet the single failure criteria (given credit for limited operator action to correct failures). 1
~~RESPONSE~~
r The Farley Nuclear Plant (FNP) steam generator atmospheric relief valves (one per steam generator) are designed to seismic Category 1 requirements. These valves are normally operated by control grade cir-cuitry and air supply. If operation by these normal means is prevented, operation will be initiated from the emergency compressed air system. f The emergency compressed air system consists of two train-oriented, seismic Category 1, air compressors with associated reservoirs and air piping. The air compressors are powered from separate and independent emergency bueses which are energized by the emergency diesel generators 7 should a loss of offsite power occur. Operation of the atmospheric relief } valves using the emergency compressed air system is independent-of the I normal air-supply and all control power. The atmospheric relief valves can also be operated manually using valve hand wheels. Three separate channels of class lE instrumentation and scismically qualified control j: board indicators are provided to monitor steam generator level and pressure { from the control room. i The most limiting single failure would be the loss of one train of.the emergency'compres ad air system. Each compressor la designed to 100% cap-ability for operation of the atmospheric relief valves so this failure wouldEbe of no consequence. A second limiting single failure would be a complete failure of one of the three atmospheric relief valves. Should this occur, the plant would l continue to be cooled using the remaining two valves and if necessary the failed valve ce Cliie repaired -(manual isolation capability is provided). .e W,
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~~2. ' Provide the capability to cool down to RRR cut-in in less than 36~~
. hours assuming the most limiting single f ailure and loss of of fsite power or show that manual actions inside or outside containment or return to hot standby ~ until the manual. actions or maintenance can be -pertormed to correct the failure provides an acceptable alternative.
~~RESPONSE~~
-Cooldown to RHR cut-in 'is accomplished using auxiliary feedwater and the atmospheric relief valves. Natural circulation is established,and thus heat is removed from the reactor, by establishing a AT between the reactor (heat source) and steam generators'(heat sink). ~ The AFW is an engineered safety feature system and is deslaned to seismic Category I. In NUREG-Oll7, "Saf ety Evaluation Report Related to Operation of Joseph M. Farley Nuclear Plant, Units 1'and 2", it was determined that "... the auxiliary feedwater system design is in conformance with our technical position APCSB-10-1 regarding diversity of power sources, system flexibility and redundance including the combination single active failure and high energy line break and is, therefore, acceptable". f A detailed description is provided in Attachment I, Cold Shutdown Scenario and a single failure evaluation is provided in Attachment II. p .- u A. . O l 's 3 a.:._ e,
- 3. Provide the capability to depressurize the reactor coolant system with only safety-grade. systems assuming a single failure and loss of of fsite power or show that manual actions inside or outside containment _ or remaining at hot standby until manual _ actions or. repairs are complete provides an acceptable alternative. RESPONSEi Two methods are provided for depressurization of the RCS with a loss of offsite power. ~ (a) Auxiliary spray utilizes either 4 weight percent boric acid. from the - boric acid tanks or refueling water (2000 ppm) from the RWST discharged from the ~ charging pumps through an air-operated valve into the pressurizer steam space. Descriptions of.the water sources and charging pumps are provided in response to question (4). The auxiliary spray valve in the flow path is seismic Category I.- The operator is statically qualified. If the normal air supply to the auxiliary spray valve is not available, a portable gas bottle would be connected to facilitate opening the valve. If the operator fails to-function, corrective maintenance would be performed to restore operability. (b) Two redundant PORV's are provided to allow the pressurizer steam space to be vented to.the pressurizer relief tank which is located inside containment. The PORV's and air supply meet seismic Category I requirements. The operators are statically qualified. Due to the extremely low probability of both PORV's being inoperable after a seismic event, reliance on these-valves is acceptable during the period from initial operation until the PORV's are upgraded af ter 4 "~ hompletion.of EPRI testing program on Safety Relief Valves and PORV's (NUREG-0578).' Each PORV has a -motor-operated isolation valve which meets seismic Category I requirements and is powered f rom emergency busses. Three channels of narrow range and two channels of wide range, Class lE pressure-instrumentation, three channels of Class lE pressurizer level instru-mentation, and seismically qualified indicators are provided for monitoring depressurization from the control' room. d O h = l-
4. Provide the capability for borating with only safety-grade systems assuming a single failure and loss of offsite power or show that manual actions inside or outside containment or remaining at hot standby until manual action or repairs are completed provides an acceptable alternative.
~~RESPONSE~~
Borating the RCS to cold shutdown conditions is accomplished using two boric acid tanks (BATS)..The BATS are seismic Category I tanks and contain 4 weight percent boric acid. I.evel and concentration are main-tained (technical specification requirement) in the tanks to ensure sufficient inventory is available to borate to cold shutdown conditions. The BATS provide suction to two, train-oriented, safety grade transfer pumps which discharge into the charging pump suction header via two paths. The primary path is the emergency boration flow path. The charging pumps, have dual function as high head safety injection pumps, are engineered safety feature equipment, seismic Category I, traEn-oriented. Discharge-into the RCS is accomplished via the cold leg injection lines of the safety injection system. Only one charging pump is required for boration evolu tions. Additional makeup is provided by the refueling water storage tank (RWST). The RWST is a seismic-Category I tank and contains 2000 ppm boric acid and is the water source for the safety injection system. The RWST 'b'rovides direct suction to the charging pumps via redundant safety-grade lines. Emergency boration can be accelerated by directing flow from the charging pumps through the boron injection tank (BIT). The BIT is a seismic Category I tank, with parallel inlet and outlet valves, designed as engineering safety feature equipment. The BIT contains 900 gallon of 21,000 ppm boric acid maintained by technical specification surveillance. ~ k j .l U
t' 5. Provide the system and-component design features necessary for the prototype testing of both the mixing of the added borated water and the cooldown under natural circulation conditions with and without a: single-failure of a steam generator atmospheric dump valve. These tests and analyses will be used to obtain information on cooldown ti::les and. the corresponding AFW requirements.
~~RESPONSE~~
The plant design provides the capability for conducting natural circu- . lation cooldown~ tests, if required, llowever, Diablo Canyon, a Westinghouse designed pressurized water reactor will conduct such tests in the proximity to the start-up of Joseph M. Farley Nuclear' Plant Unit 2, and because of the similarity in design between all Westinghouse pressurized water reactors, it is anticipated that these tests will be representative of the natural - convection cooling and boron mixing capability of Paricy. de Diablo Canyon test will demonstrate any limitations on cooldown via natural circu-lation. A comparative analysis of the natural circulation capability between Farley and Diablo Canyon is being prepared and will be provided in the near future. j l [ ~I \\ n l s dA'
n -, =. c + 1 - 6. Connit. to providing specific procedures for cooling-down using natural . circulation and submit a summary of these procedures. , RESPONSE: Specific procedures will be written for cooldown of Farley Nuclear Plant - by natural circulation. A summary of.the procedures is provided in Attach-ment I. m O g,, e' e -A. I l i 4 2' - 4._,
7. Provide or. require a seismic Category I AFW supply for at least 4 hours at Hot Shutdown plus cooldown to the RHR. system cut-iu based on the longest. time (for.only onsite or of fsite. power and assuming the. worst single fail-ure),-or 'show that an adequate alternate seismic Category I source will be ' -available.
~~RESPONSE~~
The auxiliary feedwater pumps take suction from the condensate storage tank. This tank has a capacity.of 500,000 gallons, 150,000 gallons of which is dedicated ~to the AFW. This is suf ficient quantity for remaining at hot standby for.2 hours after a reactor trip, followed by a 4-hour cooldown.to 3500 F, at which temperature the RilR system would be initiated. The tank is. ~ designed to seismic Category I requirements. The service water system provides an alternate source of water to the AFW. The service water supply to the AFW including, the storage pond is designed to seismic Category I requirements. Switchover to the alternate _ water source is performed by the operator in the control room. Redundant, Class lE tank level instrumentation and seismically qualified-indication are provided on the main control board. A tank level low alarm is provided which allows the operator at least twenty minutes to switch over to the alternate water source. a
ATTACHMENT I COLD SHUTDOWN SCENARIO (Assuming loss of all non-seismic Category I equipment) The safe shutdown design basis of Farley is hot standby. The plant can be maintained in a safe hot standbycondition while manual actions are taken to permit achievement of cold shutdown conditions following a safe shutdown earth-quake with loss of of fsite power. Under such conditions the plant is capable of achieving RHR initiation conditions (approximately 3500 F. 400 psig) in approximately 36 to 48 hours, including the time required for any manual actions. To achieve and maintain cold shutdown, four key functions must be performed. These are: (I) circulation of the reactor coolant, (II) removal of residual heat, (III) boration and makeup, and (LV) depressurization. Instrumentation required for monitoring the cooldown is listed in Section V and a discussion of maintaining RCS temperature and pressure without letdown is given in Section VI. I. Circulation of Reactor Coolant Circulation of the reactor coolant has two stages in a cooldown from hot standby to cold shutdown. The first stage is f rom hot standby to 3500 F. During this stage, circulation of the reactor coolant is provided by natural circulation with the reactor core as the heat source and steam generators as the heat sink. Steam release from the steam generators is initially via the steam generator safety valves and occurs automatically as a result of turbine and reactor trip. Steam release for cooldown is via the steam generator power operated relief valves which can be operated using the safety-grade compressed air system or manually with their hand-wheels. The steam generator power operated relief valves are accessible for local operation. The status of each steam generator can be monitored in the control room using Class lE instrumentation and seismically quali-fled indicators. Three separate channels of indications for both steam generator _ pressure and water level are available. Feedwater to the steam generators is provided from the Auxiliary Feedwater System which has a minimum reserve of 150,000 gallons in the seismic Cate-gory I condensate storage tank as the primary source and two separate Seisnde Category 1 piping sub-systems. The first sub-system is composed of two motor-driven pumps each powered from a different emergency power train, j and the second sub-system incorporates a turbine driven pump which can re-ceive motive steam from either of two steam generators. An additional j backup water source is frem the fully qualified Seismic Category I Service Water System via manual tie-in valves. The operation of the auxiliary feedwater system can be monitored in the control room using Class lE in-i strumentation and seismic Category I indicators. There is a single indica-tion of the flows into each steam generator, and pump operating status ] lights for the motor driven pumps. Redundent, safety grade indication is provided in the control room for the level in the condensate storage tank. A low level alarm is also provided to allow 20 minutes for the operator to switch over to the backup water source. Indication for the suction and 4x l NNI *-
L discharge pressure _for'all three AFW Pumps is provided in the control room and locally. L The second stage of Reactor Coolant circulation is from 3500 to cold -shutdown. During this stage, circulation of the reactor coolant is pro-vided by the Residual Heat Removal Pumps and a fully qualified, redundent system. II. Removal of Residual Heat Removal of residual heat also has two stages in a cooldown f rom hot. standby to cold shutdown. The first stage is from hot standby to 3500 F. During this stage, the steam generators act as the means of heat removal from the reactor coolant system. Initially, steam is released from the steam generators via the steam generator safety valves to maintain hot I standby conditions. When the_ operators are ready to begin the cooldown, the steam generator power operated relief valves are operated by means of the emergency compressed air system or by lccal operation with their hand-wheels. As the cooldown proceeds, the operator will occasionally adjust these valves to increase the amount they are open. This allows a reason-able cooldown rate to be maintained. Feedwater makeup to e steam genera-tors is provided from the Auxiliary Feedwater System. The Auxiliary Feed-water System has the ability to remove decay heat by providing feedwater to l all three steam generators 'for extended periods of operation. Communica-tions for these actions will be by use of either sound-powered headsets. 7 l the plant private exchange or by public address. l The second stage is from 3500 F. to cold shutdown. During this stage the l. Residual Heat Removal (RHR) System is brought into operation. The Residual Heat Removal Heat Exchangers in the RHR System act as the means of heat re-l moval from the Reactor Coolant System. In the RRR Heat Exchanger, the resi-l dual heat is transferred to the Component Cooling System which ultimately transfers the heat to the Service Water System. The Component Cooling and the Service Water. Systems are both designed to Seismic Category 1. The RHR l system includes two Residual Heat Removal Pumps and two Residual Heat 'Re-L moval Heat'Exchangers. Each RHR Pump is powered from dif ferent emergency . power trains and each RHR Heat Exchanger is coeled by a dif ferent Component Cooling loop. If any component in one RHR loop becomes inoperable, cooldown of the plant is _not compromised; however, the time for cooldown would be extended. Indication of the pump discharge flow and pressure and pump operating status is provided in the Control Room. Component Cooling water temperature from the discharge of tae RHR heat exchangers, and RHR heat exchanger discharge temperature is provided locally. L [ III. Boration and Makeup i Boration is ' accomplished using portions of the Chemical and Volume Control System (CVCS). Boric acid (4 wt.%) from the Boric Acid Tanks, is supplied f-o b L p
o l\\ i: to the suction of the Centrifugal Charging Pumps by.the Boric Acid Transfer Pumps. The Centrifugal Charging Pumps inject the borated water into the . Reactor Coolant System via the safety injection system. The'two Boric Acid _ Tanks, two Boric Acid Transfer Pumps, and the associated piping are of Seismic Category I design. Sufficient boric acid is maintained to ensure the capability to achieve a cold shutdown condition, shutdown margin of 1% ALK/K. The Boric Acid Transfer Pumps are each powered from different emergency power trains. The Boric Acid Tank level can be monitored from the Control Roca to verify 'he operability of the boration portion of the CVCS. Makeup.in excess 'of that provided by the ' boric acid tanks is provided from the Refueling Water Storage Tank (RWST) using Centrifugal Charging Pumps and the same injection flow paths as' described for boration. TVo motor operated valves, each powered f rom dif ferent emergency powerf trains and connected in parallel, will transfer the suction of the changing pumps to the RSWT. Separate redundent channels of RWST level indication are pro-vided in the Control Room. IV. Jepressurization Depressurization.is accomplished using portions of the Chemical and Volume Control System (CVCS). Either 4 wt.% boric acid or refueling water can be used as desired for depressurization with the flow path being from the Cen-trifugal Charging Pumps to the auxiliary spray valve to the Pressurizer. The Centrifugal Charging Pumps of the CVCS are of seismic Category I, and are powered from different emergency power trains. An alternate means of depressurization is use of the redundant pressurizer PORV's which vent the steam from the pressurizer to.the pressurizer relief tank. i The pumps can be operated from,and their operating status monitored in, the ~ Control Room. The-depressurization of the reactor coolant system can be monitored using safety grade instrumentation in the Control Room. Avail-able to the operator are three channels of Pressurizer pressure, three channels of Pressurizer level and two channels of reactor coolant pressure. i l l i p ba_ m
V. Maintaining RCS Temperature and Pressure Without Letdown In performing the cooldown, the operator will integrate the functions of heat removal, boration and makeup, and depressurization so that these functions can be accomplished without letdown from the Reactor Coolant System. Boration, cooldown, and depressurization will be accomplished in a series of short steps arranged to keep Reactor Coolant System ten-perature and pressure and Pressurizer level in the desired relationships. The boration requirement will be evaluated by the operator prior to initiat-ing cooldown and depressurization. Based on initial plant conditions, the operator may elect to borate using the contents of the boron injection tank and/or the contents of the boric acid tanks. Once the plant is cooled to 3500 F. and depressurized to 425 psia, Residual Heat Removal System operation is initiated and the Reactor Coolant System is taken to cold shutdown conditions. To demonstrate that boration and depressurization can be done without let-down, a simpler scenario can be used. First the operators integrate the cooldown and boration functions taking advantage of the cream space avail-able in the pressurizer and the RCS inventory contraction resulting from the cooldown. Finally, the operators use auxiliary spray from the CVCS to depressurize the plant to RHRS initiating conditions. The calculation to demonstrate this capability assumes worst case boration requirements based on core end of life / peak xenon conditions and the following RCS initial conditions following plant trip: RCS Temperature 547 F. RCS Pressure 2250 psia 1 Pressurizer Water Volume 350 ft3 Pressurizer Steam Volume 1050 ft3 Pressurizer Quality-X 0.34 The cooldown from 5470 F. to 3500 F. decreases 'the volume of water in the RCS by approximately 1413 cubic feet. This assumes that the pressurizer is not cooled and the water level is maintained at the initial condition. Makeup for contraction is supplied by 4 wt. % boric acid stored in the boric acid tanks at 700 F. A boric acid tank volume of approximately 1130 cubic feet will expand to approximately 1250 cubic feet as it is heated to the RCS temperature of 3500 F. A boric acid tank volume of approximately 1000 cubic feet is required to maintain the reactor within the technical specification shutdown requirements at 3500 F. The volume required for boration require-0 cents at 350 F is less than the contraction volume available at 3500 F. To calculate if depressurization can be accomplished without letdown and without taking the plant water solid, it was assumed that the Pressurizer was at saturated conditions with 350 cubic feet of water, 1050 cubic feet of steam, and the Pressurizer metal, all at 6530 F (2250 psia). It was further assumed that no additional water would be removed from the pressur-izer by the cooldown contraction. With these assumptions, and including the effect of heat input from the pressurizer metal, it was determined that spraying in approximately 29,505 lbm of 700 F. water would provide saturated conditions at 425 psia (4500 F.) with a water volume of 908 cubic feet and a steam volume of 492 cubic feet. .L...
= a J (Once depressurized to 425 psia, R11RS operation is initiated and cooldown is continued-to cold shutdown conditions. The cooldown from 3500 F. to ~2000 F further' decreases the volume of water in the'RCS1by approximately '350' cubic feet..This assumes that the pressurizer is not cooled and the pressurizer. water level is maintained at the level resulting from depressuri-zation. Makeup for contraction.is'again supr~ ?d by 4 wt % boric acid. A ' boric acid ~ tank' volume of approximately 340 coolc feet will expand to approxinately 350 cubic feet as it-is heated to the RCS temperature of' 200 F. An additional boric acid tank' volume of only approximately 250 cubic. feet is required to maintain the reactor within the technical specification shutdown requirements of'2000 F. The additional volume-required. for boration requirements at 2000 F.is less than'the additional contraction volume available at 200" F, thus ensuring that technical speci-fication requirements for. cold shutdown conditions are satisfied. The results of the calculations described above demonstrate that boration and depressurization can be accomplished without letdown and without taking full credit'for the available volume created by the cooldown contraction. VI. Instrumentation Class IE instrumentation and seismic Category I indication is available to monitor the. key functions associated with achieving cold shutdown from the-control room. This instrumentation is as follows: 1. -RCS wide range temperature 2. RCS wide range pressure p-s: '~ 3. Pressurizer water level 4. Steam generator water level (per steam generator) 5. Steam-line pressure (per steam line) 6. RUST. level 7. Boric acid tank level (per boric acid tank) 8. Condensate storage tank level This. instrumentation is sufficient to monitor the key functions associated with cold.~ shutdown and to maintain the RCS within the desired pressure, temperatureLand' inventory relationships. Operation of_the auxiliary systems that service the RCS can be monitored by the Control Room operator, ~ if desired, via remote communication with an operator in the plant. A.1 7 .m
~ AITACINENT II SINGLE FAILURE EVALUATION 1. Circulation of the Reactor Coolant 1. From Hot Standby to 350 F (Refer to FSAR Figures 5.5-1, 5.5-2, and 5.5-3) - Three reactor coolant loops and steam generators are provided, any two of which can provide suf ficient natural circula-tion flow to provide adequate core cooling. Even with the most -limiting single failure (of a steam generator power operated relief valve), two of the reactor coolant loops and steam generators remain available. 2. From 350 F to cold shutdown (refer to FSAR Figure 5.5-6) - Two RHR pumps are provided, either one of which can provide adequate circu-lation of the reactor coolant. II. Removal of' Residual Heat 1. From Hot Standby to 350 F (Refer to FSAR Figures 6.5-1, 10.3-1, Sheet 1 and 2.) a. Steam generator power operated relief valves - Three are pro- ' 7 vided (one per steam generator), any two of which are sufficient for residual heat removal. In the event of a single failure, two power operated relief valves remain available. Also, manual iso-lation capability is provided to facilitate repair of an inoperable
~~valve, b.~~
Emergency Feedwater Pumps - Two motor driven and one steam driven emergency feedwater pumps are provided. In the event of a single failure, two pumps-remain available, either of which can provide sufficient feedwater flow, c. Flow control valves' - Air operated, fall open valves are provided. In the event of a single failure of one flow control valve (which effects flow to one steam generator from either the AFW motor' driven pumps or the turbine' driven AFW pump) emergency feed flow can still be provided to all three steam generators from the other i . pump (s). d. ' Backup' source - A backup source of auxiliary feedwater can be i provided to the suction of the auxiliary feedwater pumps from either train of the Seismic Category I Service Water System. -2. From 350 F to 200 F (refer to FSAR Figures 5.5-6, 6.3-2, 9.2-5 and 9.' 2-3. ). a. RHR'Suetion Isolation Valves - 8701A and 8701B (to RHR Pump 1) and 8702A and 8702B'(to RHR Pump 2) - The two valves in each RHR subsystem are each powered from differnt emergency power trains. 1 Failure of-either power train could prevent initiation of RHR .) I 2:
i cooling in the normal manner from the control room. In the event of such a failure, the affected valve can be opened with its handwheel. Valve power could also be provided by temporary connections made outside containment. Any other single failure can be tolerated as it would only affect one of the RHR subsystems, and adequate cooling can be provided by the redundant' subsystem. b. RHR Pumps 1 and 2 - Each pump is powered from a different emergency power train. In the event of a single failure, either pump can provide. sufficient RHR flow. c. RHR Heat Exchangers 1 and 2 - If either heat exchanger is unavail-able for any reason, the remaining heat exchanger can provide suf ficient heat removal capability. d. RHR' Flow Control Valves HCV603A and B - If either of these normally open fail open valves closes, sufficient RRR cooling can be pro-vided by the unaffected RHR train. e. RHR/ SIS Cold Leg Isolation Valves 8888 A, B - If either of these normally open, motor operated valves, which are powered from different emergency power trains, closes inadvertently, sufficient RHR cooling can be provided by the unaffected RHR train. The affected valve can be deenergized and opened with its handwheel.
~~p.,~~
~~ f. Component Cooling Water System - Two redundant subsystems are provided for safety related loads. Either subsystem can provide suf ficient heat removal via one of the RIEt heat exchangers with heat rejection to' the service water system. g. Service Water System - Two redundant subsystems are provided for safety related loads. Either subsystem can provide sufficient heat removal via one of the CCW heat exchangers. III. Boration and Makeup (Refer to FSAR Figures 9.3-3, 9.3-4, 9.3-5 and 6.3-1) 1. Boric Acid Tanks 2A and 2B - Two boric acid tanks are provided. Each tank contains sufficient 4% boric acid to borate the reactor coolant system'for cold shutdown. 2. Boric Acid. Transfer Pumps 2A and 2B - Each pump is powered from a different emergency. power train. In the event of a single failure, either pump can provide sufficient boric acid flow. 3. Isolation Valve 8104'- If valve 8104 which is supplied from emergency power and is normally closed, cannot be opened due to power train j or operator failure, it can be opened locally'with its handwheel. If ) valve 8104 cannot be opened with its handwheel, an alternate flow path ) ..^~ is available via air operated, fail open valve PCV-ll3A and normally closed. manual; valve 8439. L A z_
4. Refueling Water Storage Tank Isolation Valves - LCV-1158 and LCV-ll5D - Each valve is powered f rom a dif ferent emergency power train, only one of these normally closed motor operated valves needs to be opened to provide a makeup flow path f rom the RWST to the Centrifugal Charging Pumps. 5. Centrifugal Charging Pumps 2A, 2B, and 2C - Pumps 2A and 2C are powered from a different emergency power train. Pump 2B is a spare pump which can be manually aligned to either train. In the event of a single failure, any one pump can provide sufficicent boration or makeup flow. 6. Charging Pump Suction Isolation Valves 8131A and B and 8130A and B - If one of these normally open, motor operated valves, each of which is powered from a different emergency power train, closes inadvertently, operator action can be used to reopen the valve with its handwheel. 7. Normal Charging Flow Control Valve FCV-122 - This normally open valve fails open on loss of air or power. If FCV-122 closes inadvertently, the charging pumps can operate on their mintflow circuits until operator action can open bypass valves 8403 and Qv606. 8. Normal Charging Isolation Valves - 8107 and 8108 - If either of these normally open, motor operated valves, each of which is powered from a dif ferent emergency power train, cannot be opened by normal means, operator action can be used to reopen the valve with its handwheel. 9. Charging Pump Discharge Isolation Valves 8132A and B and 8133A and B If one of these normally open, motor operated valves, each of which is powered from a different emergency power train, closes inadvertently, operator action can be used to reopen the valve with its handwheel. 10. Reactor Coolant Pump Seal Injection Isolation Valve 8105 - If this normally open, motor operated valve closes inadvertently, operator action can be used to reopen the valve with its handwheel. 11. Reactor Coolant Pump Seal Injection Flow Control Valve - HCV-186 - This normally open valve fails open on loss of air or power. If HCV-186 closes inadvertently, the charging pumps can operate on their miniflow circuits until operator action can open bypass valves 8389 and QV607. 12. Boron Injection Tank Isolation Valves 8803A and B - Each valve is powered from a different emergency power train; only one of these normally closed, motor operated valves needs to be opened to provide an alternate path and source for boration. 13. Boron Injection Tank Isolation Valves 8801A and B - Each valve is powered from a different emergency power train; only one of these normally closed, motor operated valves needs to be opened to provide an alternate path and source for boration.
b, p HECHANICAL ENGINEERING BRANCH QUESTIONS l110.1 Request- .There are several-safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor-coolant system (RCS) pressure. There are also some systems which- =are: rated at full' reactor pressure. In order to' protect.these systems 1from RCS pressure,.two or more isolation valves are placed 'in series to form the interface between the high pressure RCS and the low pressure systems..The leak tight integrity of these-valves .must be ensured by periodic leak testing to prevent exceeding the design pressure of the low prensure systems thus causing an inter-system LOCA. Periodic-leak testing of pressure isolation valves is required to be performed af ter all disturbances to the valve are complete. 1 Pressure isolation valves are required to be Category A or AC per l 1WV-2000 and to meet the appropriate valve Icak rate test require-J ments of IWV-3420 of Section XI of the ASP.E Code and as discussed below. Limiting Conditions for Operation (LCO) are required to added to the technical specifications which will require corrective action i.e.,- shutdown or system isolation when the final approved leakage limits are not met. Also surveillance requirements which will state the-acceptable leak rate testing frequency, shall be provided in the technical specifications. The' staff's present position is that leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute per valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy .] of the redundant pressure isolation function and give' an indication I of -valve degradation over a finite period of time. Significant increases over this limiting valve would be an indication of valve degradation from one test to another. 1 1 L'eak rates higher than 1-GPM will be considered if the leak rate l' changes -are.below 1 GPM above the previous test leak rate or system l design precludes measuring 1 GPM with sufficient accuracy. These items will-b'e reviewed'on a case by case basis. The Class 1 to Class 2 boundary will be considered the isolation point . hich must be protected by redundant isolation valves. w i, In cases where pressure isolation is provided by two valves, both will be independently leak tested. ~ When three or more valves provide isolation, only.two of~the valves need to be leak tested. Provide a list of all pressure isolation valves included in your testing . program which will!be categorized "A" or "AC" and discuss how your test-ing-program will conform to the above staff position. W
~~l I~~
3 b/ d
y
~~RESPONSE~~
I.' Valves which protect low-pressure systems from Reactor Coolant System (RCS) pressure willlbe considered pressure-isolation valves. The leaktightness of these valves will be ensured by periodic leak testing to prevent exceeding _ the design pressure of the low-pressure systems thus preventing an intersystem LOCA. These valves will be classified as Category A or -AC per IWV-2000 of Section XI of the ASME Code, 1974 Edition with Addenda through Summer 1975, and will meet the leak rate test requirements of IWV-3420 with the exception that the maximum permissible leakage rate shall be 1 GPM per valve. In cases where direct pressure-isolation is provided by two or three valves, all will be independently leak tested. Where indirect pressure isolation is provided by four or more valves, two of these valves will be leak tested. Utilizing probability figures developed by the Reactor Safety Study (WASH-1400), the NRC staf f in NUREG-0677 has established upper limits for the probability of the intersystem LOCA caused by the failure of RCS pressure-isolation valves. When the probabilities of all pressure-isolation interfaces are summed for the total-plant-intersystem-LOCA probability, the result must approach 1 X 10-6 per reactor year. The systems and valves which provide direct pressure isolation are listed below. These valves are the only barrier between the RCS and low-pressure piping outside containment. The probability of failure of these valves, based on NUREG-0677 figures, is summarized in the attached Table A. DIRECT PRESSURE-ISOLATION SYSTEMS / VALVES Systems Valve Numbers Residual Heat Q2EllV016A ~ Removal (Suction) Q2E11V001A Q2EllV016B Q2EllV001B Low Head 02EllV051A Safety Injection, Q2EllV051B Cold Leg 'JE11V051C (And RHR Discharge) Q2E11V021A Q2EllV021B Q2E11V021C Q2EllV042A 92E11V042B Low Head 02E21V077A Safety Injection, Q2E21V077B Hot Leg 02E21V076A Q2E21V076B Q2E11V044 . j _s=- j
-m .~.. 4 77-f.: ' Tha RRR:Suetien.valycs'End the:LHSI C'ald Lag (valvso will' ba laak tcoted i a F > l crch r:fualing_ outtga..Tha LHSI Hot Lig valves will ba leek taats.d each-- 1 refueling outage and eacht time flow occurs through the valves.:.Using theset'sging; frequencies,J the total intersystem LOCA probability is' e 5.2 X 10 EperJ reactor... year. This totalLis-significantly lower than'the goal of 1 X.10-6-and is satisfactory. 1 1The RHR-Suction valves and_the LHSI Cold Leg valves must be excluded from.the requirement to leak test e'ach ti=e the valves are disturbed.. The RHR Suction valves cannot be leak tested when the RCS 'is' above 0 212 F because any leakage.would immediately flash into steam making" accurate measurement of the _-leak rate' irpossible. Whenever -the RCS is below 3100F, Technical Specification 3.4.10.3 requires that these RHR .-Suetion' valves be'open. - Technical Specification 3.4.1.4 - requires that .in Mode'5 (Cold Shutdown) both trains of RHR must be operable._.To test ~ these_ valves will require securing each train successively, thereby-i placing the plant in a degraded mode of operation and under an LOO. To minimize the number of LCO's,.a testing frequency of once each ' refueling outage will be used. _The valves will be tested in Mode 5 as the plant.is coming up;from Mode 6 and while the RCS is below 2000F. ' The LHSI Cold ' Leg valves are also limited as to when they can be leak 4 tested. - Whenever the RCS is-above 212 F, any leakage past these valves would flash into steam making accurate measurement of the leakage-rate 0 impossible. When the RCS is below 310 F, these valves serve as the - RHR Discharge to RCS; consequently, they have flow through them. The valves will be isolated and tested in Mode 5 as the~ plant is coming -up from Mode 6 and while the RCS is below 2000F. II..There are other systems which also connect the RCS with low-pressure ~ h~ - . piping outside containment. However, these systems are rated at reactor pressure and contain' four or more high-pressure normally-closed valves. These, systems serve only an indirect pressure-isolation function. However, in order to' comply with the NRC staff position that the Class 1/ Class 2 boundary be considered the. isolation point, the following Class 1 valves will be leak tested each refueling outage: INDIRECT PRESSURE-ISOLATION VALVES Q2E21V062A Q2E21V062B Q2E21V062C Q2E21V066A Q2E21V066B Q2E21V066C-Q2E21V077C Q2E21V078A' .Q2E21V078B Q2E21V078C -Q2E21V079A Q2E21V079B _ Q2E21V079C 4 , 4 !z. u._ - uv.u.uum..u ',E--
e Lcak te: ting threo valves each time.they move would be unnecessary and unwise. The probability of four or more high-pressure normally-closed valves failing sLoultaneously is insignificant 1y small, especially since .the Class 1 valves can be leak tested each refueling outage. Since these valves can be tested only when the RCS is below 212 F, the plant would have to go to cold shutdown any time any one of these valves is disturbed. Subjecting the plant to unnecessary transients is an unwise method of operation. In addition, all of these thirteen (13) valves .are limited in the amount of reverse leakage that they can pass. In line with these valyes are locked, throttled globe valves whose maximum flow rate (193 GPM) is controlled by the plant Technical Specifications Surveillance Requirement 4.5.2(i). The thirteen valves listed on page 2 will be leak tested each refueling outage in Mode 5 as the plant is. coming up from Mode 6 and while the RCS is below 200 F. III. One other system which contains low-pressure piping and which connects with the RCS is the Safety Injection Accumulator System. Although this system is located entirely within containment, the Accumulator Discharge check valves will be leak tested each refueling outage and each time the valves are disturbed. This will help to minimize the possibility of a LOCA inside containment. These valves are listed below: Q2E21V032A Q2E21V032B Q2E21V032C Q2E21V037A ~ Q2E21V037B Q2E21V037C IV. In the event that the Class 1 valves should fail and the high-pressure Class 2 piping is subjected to RCS pressure, a LOCA would not result. The high-pressure Class 2 piping is rated at the same pressure as reactor coolant piping. The ASME Code Section III requires that Class 2 piping welds receive a radiograph and dye penetrant examination to ensure structural integrity. In addition, the piping receives a hydro-static pressure test, a visual examination for leakage, and a visual inspection of support capability of all hangers and supports. The preservice inspection of Class 2 piping was originally planned to be in accordance with the 1971 Edition of Section XI with Addenda through Winter 1972. The PSI program was voluntarily upgraded to the 1974 Edition with Addenda through Summer 1975. The NRC has approved the Farley Unit 2 preservice inspection program and has stated that the PSI is in compliance.with the ASME Code and with 10 CFR 50.55a(g)(2). The Unit 2 inservice inspection program is essentially identical to the. Unit 1 ISI program which has also been approved by the NRC. The PSI and -ISI programs exist for the purpose of ensuring that the systems function ~ safely over the 40-year life of the plant. V. The position on pressure-isolation valves as stated above is the basis for the attached Technical Specification.
L P TABLE A. J. M. FARL'EY NUCLEAR PLANT UNIT 2 RCS PRESSURE-ISOLATION VALVES f Intersystem LOCA Probability Flow Path: Flow Path: Total System: Number of Type of Original - Revised - Revised System Flow Paths Interface No Testing Leak Testing Probability 1.8 x 10'
4.2 x 10

8.4 x 10 ~
~~Residual 2 Two Heat Removal locked-closed (Suction) motor-operated gate valves -9 3.0 x 10'0~~
7.4 x 10

4.4 x 10' Low-Head 6
~~Three check Safety Injection, valves Cold Leg ~ Low-Head 2 Two check valves 2.8 x 10'~~
1.5 x 10~

3.0 x 10 Safety Injection, and one normally-Hot Leg closed motor-operated gate valve

Leak testing will be performed every *cfueling on all valves in the " Type of Interface" column for the RHR Suction and LHSI Cold Leg Systems.
C* Leak testing will be performed every refueling and whenever flow occurs through the flow path on all valves listed in the " Type of Interface" column for the LHSI Hot Leg System. -8 5.2 x 10 Total Intersystem LOCA Probability a-
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited to: a. No PRESSURE BOUNDARY LEAKAGE, b. 1 GPM UNIDENTIFIED LEAKAGE, c. 1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator, d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e. 31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 20 psig f. 1 GPM leakage 'from any Reactor Coolant System Pressure Isolation Valve APPLICABILITY: MODES 1, 2, 3 and 4 speci fied in Table 3.4-1. ACTION: a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. b. With any. Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.7.2.iReactor Coolant System leakages shall be demonstrated to be within each of the above limits by; a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours. b. Monitoring the containment air cooler condensate level system or containment atmosphere gaseous radioactivity monitor at least once per 12 hours. [c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of )1 the affected system from the low pressure portion within 4 hours by use of at least two closed manual or deactivated automatic values. or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. j FARLEY-UNIT 2-3/4.4-17
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump c. seals at least once per 31 days when the Reactor Coolant System pressure is 2235 1 20 psig with-the modulating valve fully open. d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours. The provisions of Specification 4.0.4 are not applicable for entry into MODE 4. Monitoring the reactor head flange leakoff system at least once per e. 24 hours. 4.4.7.2.2 Each Reactor. Coolant System Pressure Isolation Valve specti ;f in Table 3.4-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit: a. Every 18 months during startup. b. Prior to returning the valve to service following mcintenance, repair or replacement wcrk on the valve affecting the seating. capability of the-valve. s Following valve actuation due to automatic or manual action or c. flow through the valve for valves identified by an asterick. D FARLEY-UNIT 2 3/4-4-18
O-i REACTOR COOLANT SYSTEM TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES Q2EllV016A Q2E11V001 A Q2EllV016B Q2E11V001B - Q2E11V051 A Q2EllV0518 Q2EllV051C Q2EllV021 A Q2EllV021B Q2EllV021C Q2E11V042A Q2E11V042B Q2E21V077A* Q2E21V077B* Q2E21V076A* Q2E21V076B* Q2E11V044* Q2E21V062A Q2E21V062B Q2E21V062C Q2E21V066A Q2E21V066B Q2E21V066C Q2E21V077C Q2E21V078A Q2E21V078B Q2E21V078C Q2E21V079A Q2E21V079B Q2E21V079C Q2E21V03?A* Q2E21V032B* Q2E21V032C* - Q2E21V037A* Q2E21V037B* Q2E21V037C* .FARLEY-UNIT 2 3/4 4-19 =
1 t REACTOR'C00LANT SYSTEM BASES 3/4.4;7 REACTOR COOLANT SYSTEM LEAKAGE 3/4.'4. 7.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor. Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. 3/4.4.7.2 OPERATIONAL LEAKAGE Industry experience has shosn that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional-leakage. The 10 GPM IDENTIFIED -LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose-presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems. The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 31 GPM with the modulating valve in the supply _line fully open.at a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses. The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a s. mall-fraction of Part 100 limits in the-event of either a steam generator' tube rupture or steam line break. The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents. The 50.0 gpd leakage limit per steam generator ensures that steam generator tube integrity i-s maintained in the event of a main steam line rupture or under LOCA conditions. PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of?an impending gross failure of the pressure boundary. Therefore, ~ the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD. SHUTDOWN. FARLEY-UNIT 2 B.3/4.4-4
INSERT TO PAGE B 3/4 4-4 The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. a h a H o, e
TMI Related Question on Section II.B.2, Plant Shielding REQUEST: Your response to Item 2.1.6b/II.B.2 Plant Shielding is incomplete. Provide the following information.
l.. Item 78 - last paragraph - Figures showing all radiation zones used in your plant shielding analyses to determine direct radiation and sources of indirect radiation. All figures should be clearly legible.
2. Item 7c - the projected doses to individuals for necessary occupancy times in access-controlled areas in the auxiliary building. 3. Item 7F - specify completion dates. 4. For other modifications, which allow access to areas where access would be useful but not vital, you should specify the anticipated modifications and the scheduled completion date for modification. It is our position that all items in Section 7 must be completed prior to full power licensing except plant shielding modification k which is January 1,1981, or full power operation, whichever is ' ~ _ later.
~~RESPONSE~~
1. Item 7b are the drawings identifying all the radiation zones used in the Farley Unit 2 shielding analyses to deter-mine direct radiation and sources of indirect radiation. 2. Item 7c The projected doses to individuals based on the estimated post-accident dose rates shown in Table 2, for necessary occupancy time in vital areas are shown in Table 1. Where occupancy times for a vital area are greater than 12 hours the dose to an individual is projected, based on a maximum of 12 hours of exposure, unless otherwise noted. Tables 1 and 2 are enclosed. A w
.~T 3. Item 7F-Shielding-modifications developed in the design review will be implemented by January 1,1981, or prior to full power operation, whichever is later. 4. Other Shielding Modifications No modifications are presently anticipated for areas where access would be'useful but not vital. Entry into areas where access would be useful but not vital would be accom-plished using hallways and corridors of Unit 1 for each elevation. j-- e
4 l i L Page 1 of 4 Table 1 Post-Accident Projected Personnel-Ooses - Farley Unit Nc. 2 Projected Designated Necessary Dose Per 3 ' Vital Areas Radiation Zonel Occupagcy ' Individual Post-Accident Time (REM) Conrnents Aux. Building. Elev. 155': Primary Access Point (Security Center) . A-I 24 hr./ day 0.03 Assuming average dose rate, 0.0025 rem./hr. . Control Room (2401) A-! 24 hr./ day 0.03 , 0.0025 rem./hr. . Operational Support Center (Rm. 2401) A-I 24 hr./ day 0.06 , 0.005 rem./hr. Technical Support Center A-I 24 hr./ day 0.06 , 0.005 rem./hr. .(Rm. 2453) Unit No. 2 Entrance Passageway (2402) A-I 1 hr./ day 0.015 , 0.015 rem./ar. I 11way (2409) A-III 1 hr./ day 0.5 0.5 rem./hr. Containment Pur e Air Equipment Room 2429) A-IV 5 min./oper.5 0.225 Assuming average dose rate, 30 rem /hr., minimum shielding (Aux. Airlock) (A-VIII) 0.5 cm iron and 3 meters distance. Aux. Building & Containment Purge Ventilation Equipment Assuming average dose rate, 30 remdhr., minimum sheilding 5 Room (2418) A-IV/A-VII 10 min./oper 0.45 0.5 cm iron and 3 meters distance. Also, assuming portable shielding in place to reduce streaming through door from personnel access airlock or 10 hours decay time. Motor Control Centers Hallway adjacent to Room 2418 A-III 10 min./oper.5 0.085 Assuming average dose rate 0.5 rem./hr. MCC Room 2478 A-III 10 min./oper.5 0.08$ , 0.5 rem./hr.
L i L i [ Pap 2 of 4 Table 1 Post-Accident Projected Personnel Doses - Farley Unit No. 2 -a Projected ^ Designated Necessary Dose Per 3 Post-Accident Occupancy Individual Vital Areas Radiation Zonel Time 2 (REM) Conrients Aux. Building. Eley. 139: Pcst-Accident Liquid Sampling I hr./ day Panel (Hallway 2322) A-III (3 samples / day) 0.75 Assuming average dose rate. 0.75 rem./hr. Load Center Rooms (2343 & 2335) A-III 30 min./oper.5 0.25 , 0.5 rem /hr. Cable Spreading Room (2318) A-II 30 min./oper.5 0.024 0.05 rem./hr. Electrical Penetration Rooms (2334) A-VI 10 min./oper.5 0.315 Assuming average dose rate,1000 rem /hr., 2-inch lead (2333 & 2347) A-VI 12 min./oper.5 0.375 shielding and average distance of 3 meters. Counting Room (2326) A-III 15 min./oper.5 0.525 Assuming average dose rate, 2 rem./hr. Radiochemistry Lab (Rm 2324) A-III C min./oper.5 0.5 , 5 rem./hr. Spectrophotometer Lab (Rm. 2325) A-III 5 min./oper.5 0.5 , 5 rem./hr. Filter Rooms Access Corridor A-VI 16 min./oper.5 0.72 Assuming 3" lead shield on chemical volume control tank transfer line in corridor and no entry into filter rooms. Aux. Building, Elev. 130: Steam Generator Blowdown Control Panel (Passageway 2602) A-I 30 min./ day 0.005 Assuming average dose rate, 0.01 rem./hr. Aux. Building, Elev. 121: Radioactive Gaseous Effluent Assuming 1-1/2" lead shields in use and three samples per day. Sample Panel (llallway 2209) A-IV 30 min./ day 0.39 Volume Control I.mL Rmth Rods (Hallway 2209) A-V 10 min./oper.5 0.85 Assuming average dgse rag 01 decayfhr. and no entry for 5 rem. 10 hours af ter acc dent ~
1 Page 3 of 4 Table 1 Post-Accident Projected Personnel Doses - Farley Unit No. 2 Projected Designated Necessary Dose Per Post-Accident Occupancy Individual 3 Vital Areas Radiation Zonel Time 2 (REM) Comments Aux. Building, Elev. 121: Hot Shutdown Panel Room (2202) A-I 24 hr./ day 0.12 Assuming average dose rate 0.01 rem./hr. Battery Charger Room (2225) A-I 30 min./oper.5 0.005 0.01 rem./hr. Switchgear Rooms (2229 & 2223) A-!/II 30 min./oper.5 0.015 Assuming average dose rate in A-I, 0.01 rem./hr. A-II, 0.1 rem /hr. Penetration Room Isolation Panel (Corridor 2208) A-III 1 hr./ day 0.5 , 0.5 rem./hr. Aux. Building, Elev. 100: Component Cooling Water Heat Exchanger Room (2185) A-III 15 min./oper.4.5 0.375 Assuming average dose rate. 0.5 rem./hr., based on no entry for 10 hours after accident (90% decay) A-IV 15 min./oper.4,5 0.875 Assuming average dose rate, 3.5 rem./hr. based on no entry for 10 hours after accident (90% decay) Aux. Feedwater Pump Assuming average dose rate,1 rem./hr., based on no Room (2193) A-IV 10 min./oper.6 0.17 entry for 10 hours after accident (90% decay) Motor Control Center Panel Assuming average dose rate, I rem./hr., based on no (Room 2190) A-IV 10 min./oper.5 0.17 entry for 10 hours af ter accident (90% decay) Plant Heating Equipment Assuming average dose rate, 5 rem./hr., based on no Room (2189) A-IV 10 min.oper.4,5 0;85 entry for it hours after accident (90% decay) Equipment Room (2194) A-IV 10 min./oper.4,5 0.85 Assuming average dose rate, 5 rem./hr., based on no entry for 10 hours af ter accident (90I decay) Assuming 1" portable lead shield in use and 5 meters Radwaste Systen Control Panel (RM. 2103) A.IV 1 hr.j x: 0.165 distance to control panel. / w
. - - ~ --... t i Page 4 of 4 Table 1 Post-Accident Projected Personnel Doses - Farley Unit No. 2 i Projected -Designated Necessary . Dose Per Post-Accident Occupancy Individual 3 l Vital Areas Radiation Zonel TimeZ (REM) Cornents l Aux. Building. Elev. 83: Corridors (2103 & 2104) A-Ill I hr./da'y 0.5 Assuming average dose rate. 0.5 rem./hr. Assuming maximum dose rate. 0.1 rem./hr., and 6 hour Hydrogen Recombiners 4 (' (Rms. 2105 & 2106) A-Il 24 hr./ day 0.6 maximum exposure per individual per day. Monitoring Control Panel Race A-V 5 min./ day 0.85 Assuming average dose rate. 10 rem./hr. (2110) Aux. Building. Elev. 83: Waste Monitoring Tank Pump Room (2109) A-IV 5 min./ day 0.85 Assuming average dose rate,10 rem /hr. ' Waste Evaporator Feed Pump Room (2122) A-V 5 min./ day 0.85 ,10 rem./hr. i Floor Drain Tank Pump Room (2121) A-V 5 min./ day 0.85 ,10 rem./hr. (1) Limits given in Table 2. (2) Estimates based upon operational experience from Unit No.1. (3) Maximum occupancy time,12 hours / day per individual. (4) Two or more individuals required to accomplish job. l (5) Operation perfomed on an infrequent schedule (not daily) l \\ l
Table 2 Designated Post-Accident Radiation Zone Dose Rate Limits Farley Uni t flo. 2 Zone Designation Zone Dose Rate Limits (Rem./Hr.) A-I 0 < D < 0.015 A-II 0.015 < D < 0.100 A-III 0.100 < D < 5 A-IV 5 < D < 50 A-V 50 < D < 500 A-VI 500 < D < 5000 A-VII 5000 < D < 50,000 A-VIII 50,000 < D < 500,000 t v A
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~ l TMI Related Question on Section II.F.1, Additional Accident Monitoring Instrumentation.- REQUEST: Provide a description of the two high range containment monitors required and specify the location of.these monitors inside con-tainment. The description of the monitors should include: 1. name of manufacturer and model number of monitors; 2. verification that the monitors meet the specifi-cations of Table II.F.1-3; 3. verification that the monitors are or will be l 1 operable on January 1,1981, or prior to full power operation, whichever is later, and; 4. a plant layout drawing showing the location of the monitors. l 5. The monitors should be located in a manner as to provide a reasonable assessment of area radiation conditions inside containment. The monitors should f s be widely separated so as to provide independent . measurements and should " view" a large fraction of the containment volume. Monitors should not be placed in areas which are protected by massive shielding and should be reasonably accessible for replacement, maintenance, or calibration. Place-ment high in a reactor building dome is not rec-ommended.
~~RESPONSE~~
1. The manufacturer's name is Victoreen, Inc. and the model number of the monitor is 875 and includes a model 877-1 detector and model 876A-1 readout. 2. The monitors meet the following specifications: REQUIREMENT The ability to detect and measure the A radiation level within the containment during and following an accident. ~ 6.. 9
7 RANGE 1 R/hr to 10 R/hr for photon radiation (overlap with normal radiation monitor (s) range by a factor of ten.)
~~RESPONSE~~
-15% at 80 Kev and 8% from 100 Kev to 3 Mev. REDUNDANT Two physically separated monitors located 90 apart. RELIABILITY Fulfills requirements of Regulatory Guide 1.97.and will meet Regulatory Guide 1.89 and IEEE 323 (1974). SPECIAL In place calibration by electronic signal CALIB W ION substitution is provided for all range decades above 10 R/hr. An in place calibration check will be performed using 137 a 230 milli curie Cs source for the 1 to 10 R/hr range (assay date 8-27-76) SPECIAL Victoreen calibrated and type tested rep-fg .resentative samples of detectors on at 3 least one point in each decade of range 6 from 1R/hr to 10 R/hr. Victoreen will provide certification of-calibration of each detector for at least one point per decade of range between 1 R/hr and 3 10 R/hr. 3. The detectors, cable, and rack chasis has been ordered and will be onsite by October.1,1980. The readout device has been ordered and is scheduled to be received onsite during the later part of December 1980. Alabama Power Company expects the monitors to be operable by January 1,1981.
4. Figure 1 is attached and indicates the location of the containment high range monitors (RE 27 A & B). 5. The monitors will be located 90 apart just above the operating deck inside containment and will provide a reasonable assessment of area radiation conditons inside containment. The monitors will be located in areas that are not protected by massive shielding and reasonable access will be provided for replace-ment, maintenance, or calibration. y v s L +'ni
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TMI Related Question on Section II.F.2, Inadequate Core Cooling Instrument Data Request The Instrumentation and Control System Branch has reviewed the applicant's response to Item II.F.2 to determine the degree to which the proposed instrument system, a subcooling meter, meets the applicable criteria of the current draft version of Regulatory Guide 1.97, Revision 2. Provide additional infor-mation on the Farley Unit 2 subcooling meter design to indicate how the design meets the requirements of Regulatory Guide 1.97, Revision 2, Table 1, Instrument Category 3. Any deviations should be reported and either corrected or justified. In addition, demonstrate the appropriate criteria have been developed to assure that the output of the computer will be accurate within the overall uncertainty noted 11 Table 1 of the Farley 2 submittal. General criteria for safety computer sof tware are being developed for the NRC at Oak Ridge National Laboratory. Our consultants at ORNL will be available to discuss these criteria with the applicant if necessary. Provide also a description of periodic tests of the subcooling meter to confirm the accuracy of the calculator for temperatures and pressures anticipated during accidents. )-
~~Response~~
It is noted that R.G.1.97 Revision 2 Draft 1 dated 12/79 has been superseded by Draft 2 which has failed to achieve NRC/ACRS resolution and will be superseded by at least one more draft before resolution is achieved. This draft of R.G.1.97 Rev. 2 requires type E instruments, such as the subcooling meter,to be qualified to the conditions of operation and provides no requirements to meet single failure and seismic criteria. The applicable criteria for the subcooling meter are power source supply, quality assurance level, display type, and display method. The attached copy of a revised Table 1 of Unit 2 Action Plan Section II.F.2 indicates that the subcooling meter is in full compliance with the applicable require-ments of R.G. 1.97, Draft 2. Provisions for assuring that the perfcrmance of the unit meets the overall uncertainty, noted in Table 1, include:
1) engineering verification of the software by extensive testing under the directicn of personnel other than those who developed the software, 2) the use of ROM (read only memory) for software, 3) complete test of the unit prior to shipment, and 4) procedures for reverification in the field according to the vendor's technical manual.
A. 4
g l Periodic retesting of the subcooling monitor will be performed at refueling intervals u'nder the Preventive Maintenance Program, utilizing an Instrument Maintenance Procedure (IMP). The IMP is essentially a repeat of the initial calibration and functional test procedure used for acceptance -testing in the initial set-up of the subcooling monitor. l This test is a full functional checkout of proper input conversion, calculational output accuracy, MCB meter indication accuracy, MCB annunciation and local alarm outputs, temperature and pressure auctioneering, display function outputs, and diagnostic output verification. Testing is accomplished following input calibration verification by using analog test signals generated from calibrated plant test equipment to simulate various combinations of temperature and pressure conditions which encompass the normal operating (various subcooled conditions), approach to saturation, saturation and super-heated ranges. For each temperature and pressure test point combination, the subcooling monitor calculational outputs and MCB meter indications are verified to be'in agreement with expected values as determined from the steam tables. Alert and alarm conditions are simulated to verify proper MCB meter indication, annunciation actuation and local subcooling nonitor alarm indications. In addition, all diagnostic and self testing features and display function outputs are verified to be functional. Self-test features are used to test alarm circuitry and MCB meter ~ ' r operation on a routine basis. Inputs which are common to the sub-cooling monitor and Process Computer are routinely verified to be in agreement. Subcooling monitor display functions are also routinely used to check monitor operability. ) i j l H i
Farley tiuclear Plant Unit 2 38 C:cket f;o. 50-364 TABLE 1 IriFORTMTI0tt REQUIRED FOR THE SUBC00LIllG M0tilTOR (1 of 3) DISPLAY 1. Information displayed P - Psat subcooled T - Tsa t superheat 2. Display type Analog and Digital 3. Continuous or on demand Analog - continuous Digital - on demand 4. Single or redundant display Redundant 5. Location of display fleter - ma in control board Microprocessor - main control room instrument racks 6. Alarms (include setpoints) Caution: 25"I' subcooled for RTD 0 15 F subcooled for T/C 0 0 F subcooled for RTD Alarm: 9' and T/C 7. Overall uncertainty Digital - 40F for T/C; 30F for RTD Analog - 50F for T/C; 50F for RTD 8. Range of display Calibra ted region - 1000 psi sub-0 cooled to 2000 F superheat overall; never offscale. 9. Qual ifica tions !!ane at present 10. Quality Assurance Requirements Complies with 10 CFR 50, App. B 11. Power Supply Source Emergency Power Supply C3 Clit ATOR 1. Type Dedicated digital 2. If process computer is used, N/A specify availability 3. Single or redundant calculators Redundant 4. Selected logic Highest Temperature for RTD or T/C and lowest pressure 5. Qualifications None at present 6. Calculational technique Functional fit - ambient to critical point
Forley N'uclear Plant-39 - Unit 2 Cocket No. 50-364-TABLE 1 !!IFORf4ATI0tl REQUIRE 0 FOR IllE SUBC00LIflG t4001 TOR -(2 of 1) It;PUT 1. Temperature (RTDs or T/Cs) RTO, T/C, and Tref 2. Temperature (number and location RTD - 2 hot and 2 cold legs ofsensors) per channel 0
~~3.. Range of temperature sensors RTD 700 F T/C 16500F (calibration 0~~
(unit range 0-2300 F) 4. Uncertainty of ~ temperature 10.n. RID l sensors 5. Qualifications IEEE 3231971 6. Pressure (specify instrument used) RCS Uide Range Pressurizer k {. Pressure (number and location 2 wide range - Loops 1 and 3 t ' of sensors) I narrow range - Pressurizer (per channel) l 8. Range of pressure sensors Wide range 3000 psi Narrow range - 1700-2500 psi l 9. Uncertainty of pressure Wide range - 11% l t! arrow range - il.5% Pressurizer - 11.0%' H
~~10. Quali fica tions -~~
IEEE 323 1971 9 ACKUP CAPABILITY 1. Availability of temperature Temp - Swap between T/C and P.TD. and pressure Press - Can defea t any uf the three-inpu ts. System uses auctioneered low pressure. 2. Availability of steam tables Saturated steam tables and tables to verify required sub. cooled conditions are included in Emergency Procedures. u- - - ~ -
Farley fluc! ear Plant Unit 2 40 Dacket tio. 50-364 TABLE 1 ItiFORMATI0ff REQUIRED FOR fill: SUBCOOLll;f; MutliTOR (3 of ~l) E2.CKUP CAPABILITY (C0tlTIllUED): 3. Training and operators Operators have been trained on the use of the subcooling monitor to determirn! required subcooling co rid i t io rr.. 4. Procedures Emergency procedures have been revised to describe the utilization of the subcore cooling monitor readout arnt appended portions of the s team tables to determine subcooling conditions. A system opera ting procedure has been written to guide operators in the operation of the subcooling moni tor. Appro-pria te personnel have been trained in these procedures. i(. w - - hww
TMI Related Question On Section II.F.2, Inadequate Core Cooling Instrument Data Request Provide sample typeout sheets for the incore thermocouple data, including the core map display and the trend typewriter typeout. Explain all symbols and numbers. For each display give the location of the printout (eg control roon) and time to obtain typeout.
~~Response~~
Sample typeout sheets for the incore thermocouple data, including the core map display and the trend typewriter typeout, are attached. is the short incore thennocouple printout which is printed on the trend typewriter located in the control room. The time to obtain this printout is 3.83 minutes. Attachnent 2 is the long incore thermo-couple printout which currently prints out.in the computer room located two floors below the main control room. The time to obtain a printout is 25 seconds. Upon completion of the Technical Support Center (TSC) scheduled to be completed by January 1,1981, the long printout will be displayed in the TSC. The thermocouple data is stored on a rotating file which contains values accumulated every hour for the last 24 hours plus the current infonmation. >When printing data for a short or long map a " map" value of zero indicates the most current data, "1" indicates data is one interval old, "2" indicates data is two intervals old,.. 24 indicates that data is 24 intervals old. A request for a map includes an entry to indicate which set of data is de-sired. -l
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ML19345A202

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