ML23006A132

0 to Updated Safety Analysis Report, Appendix H, Revision 29, Reactor Pressure Vessel Design Summary Report

ML23006A132
Person / Time
Site:	Monticello
Issue date:	12/19/2022
From:	Xcel Energy, Northern States Power Company, Minnesota
To:	Office of Nuclear Reactor Regulation
Shared Package
ML23006A159	List: L-MT-22-021, 0 to Updated Safety Analysis Report, Appendix K, Revision 37, Renewed Operating License-USAR Supplement ML23006A127 ML23006A128 ML23006A129 ML23006A130 ML23006A132 ML23006A133 ML23006A134 ML23006A135 ML23006A136 ML23006A137 ML23006A138 ML23006A139 ML23006A140 ML23006A141 ML23006A142 ML23006A143 ML23006A144 ML23006A145 ML23006A146 ML23006A147 ML23006A148 ML23006A149 ML23006A150 ML23006A151 ML23006A152 ML23006A153 ML23006A155 ML23006A156
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L-MT-22-021
Download: ML23006A132 (1)

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• MONTICELLO The jet pump instrumentation safe end and penetration seal was replaced with the jet pump instrumentation nozzle penetration seal, using low carbon 316 to replace the original ASTM A508 Class II material. The new stress analyses are contained in General Electric Stress Report No. 23Al939, also included in Design Change No. 83Z049C.

The core differential pressure and standby liquid control, and the jet pump instrumentation modifications invalidated the "Summary of Stress Analysis for Core Differential Pressure and Liquid Control Nozzle, Head Cooling Spray and Instrumentation Nozzles, Vent Nozzle, Instrumentation Nozzles, Jet Pump Instrumentation Nozzles, Drain Nozzle, High Pressure Seal Leak Detector Nozzle anci Low Pressure Seal Leak Detector Nozzle" shown on page 4-28 of Exhibit 4.

Also, in 1984, a corrosion resistant cladding overlay was applied to the inside diameter of the RV head vent nozzle and RV head cooling spray and instrumentation nozzles. The weld overlay of 308L isolated the IGSCC susceptible existing weld butter located in the weld residual stress area from the reactor coolant. As documented in General Electric Stress Report No.

23A4280, part of Design Change No. 842068, stress calculations performed originally at this location are still valid.

The recirculation inlet and outlet nozzles were both modified during the 1984 outage. General Electric Stress Report No. 23Al627, part of Design Change No. 83Z049A, documents the analysis of the redesign and replacement of the recirculation inlet nozzle safe end and thermal sleeve, including the attachment weld and the weld overlay to the recirculation inlet nozzle.

This design change invalidated the "Summary of Stress Analysis for Recirculation Inlet Nozzle" shown on page 4-22 of Exhibit 4.

Bechtel Stress Report No. SR-10040-SS2 (Rev. 3), also part of Design Change No. 83Z049A, documents the analysis of the replacement of the recircula-tion outlet nozzle safe end fitting, a forged,and machined component made of SA 358 Type 316 stainless steel. The "Summary of Stress Analysis in Recirculation Outlet" shown on page 4-24 of Exhibit 4 has been invalidated by this change.

In 1986, new core spray safe ends featuring a tuning fork design with a thermal sleeve were installed along with a section of piping upstream at each nozzle. This modification was performed to minimize the chance of IGSCC from occurring in the Core Spray System. The new stress analyses is documented by Bechtel Document 301-P-5.

Also in 1986, the CRD return nozzle, previously capped in 1977, was again modified. The purpose of the modification was to remove that portion of the existing weld butter layer susceptible to IGSCC, and re-clad the weld prep area with corrosion resistant cladding and install a new nozzle cap.

General Electric Stress Report No. 23A5553, included as part of Design Change No. 862016, documents the analysis .

H.1-3 REV 7 12/88

1-GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPEC . . . o. 21Alll2 ~Ell . . . o. 6 PURCHASE SPECIFICATION 5 SM NO. 4 CON"!" ON SMEET' s.o D!SICN R!OUIRl!'.MENTS 5.1 Operating Conditions 5.1.1 Internal Pressure Design Pressure: 1250 p~ig at bottom of the reactor vessel Normal Operating Pressure: 1000 psig at -top of reactor vessel 5.1.2 Temperature Design Temperature: 575°F Norm.al Operating Temperature: 546°F 5.1.3 Reactor Core and Internal Weight The weight of the reactor core and internal structure, centers of gravity and distribution of loadings are shown on Drawing 886D482.

5.1.4 Water Weight The weight of water contained in the vessel for variqus conditions of operation are presented on Drawing 886D482.

5.1.5 Pipe Reactions The Buyer shall provide the Seller with the pipe reactions which the connecting piping will apply to all nozzles with a nominal size larger than the reactor vessel wall thickness and _those nozzles which in addition are subjected to significant thermal cycling. The reactions will be limited by the Buyer such that the combined stress as due to pipe reactions and design pressure in the vessel shell at the nozzle attachment will not exceed the design stress allowed by the ASME Code,Section III. These pipe reactions shall be used in the detailed stress analysis required by the Code and perfonned by the Seller. This analysis shall include the thin section of the nozzle in the vicinity of the weld preparation for connecting piping, any bi-metal weld and shall take into account the nozzle cladding.

5. 1. 6 Control Rod Drive Weight and Reaction The momentary reactions whjch are suddenly applied to each control rod drive housing in che vessel hotto~ head are presented on Drawing 886D482.

ISSUEC:

MAR - 5 1969

1-11 GENERAL ft ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION SPEC. NO. 21Alll2 REV. NO. 6 SH NO. 9 CONT ON S><EET lC analyeia of stresses under tea,a*a 4!t;ate and transient conditions to detemine uitability of the design *lth respect to the allowable stress given in ASKE Code,Section III, and to determine the opera-tional limitations. with respect to fatigue of the reactor vessel materials over the life of the reactor vessel (Design Objective) using the loading conditions supplied by the Buyer.

6.3 Parts of the Reactor Vessel Assembly to be Analyzed The parts of the reactor vessel to be analyzed shall include:

head closure, bottom head, shell adjacent to reactor core, reactor vessel supports and stabilizers, supports for reactor vessel internals, control rod drive penetration, feedwater nozzle, poison nozzle, emergency core cooling nozzles, drive system return nozzle, and all nozzles 10" or larger in size.

6.4 Closure Head Seal Calculation To assure meeting sealing requirements of the main closure seal as specified in paragraph 5.2.S above, the relative rotations of the flanges shall be calculated. These rotations shall be used to demon trate analytically satisfactory seal performance using t~e following assumptions:

6.4.l The 111ating surfaces of the flanges shall be assumed rigid, 6.4.2 The rotation shall be assumed to cause contact over the minimum area which will sustain the loading between the faces.when stressed to the yield strength at the metal temperature.

6.4.3 The flange faces shall be assumed to diverge from the contact area, specified in paragraph 6.4.2, through the angle of calculated relative rotation less any radial taper machined on the face(s) to accommodate the flange rotations.

6.4.4 It may be assumed that the seal will be maintained if, at both 0-ring seal locations, the separation between flanges is less than the minimum elastic spring-back of the 0-ring.

6.5 Calculations The calculations shall be clear and in sufficient detail to rermit independent checking: Specific references shall be given for all formulas and method~ us~d ?r the formulas and methods shall be derived independently. Calculation shall be submitted to the Buyer for approval:

• ISSMAR - 5 1969

1-15 GENERAL fl) ELECTRIC

• PURCHASE SP ECI FICA TION ATOMIC POWER EQUIPMENT DEPARTMENT SPEC. NO.

SH NO.

2lA1112 13 C:ONT REV. NO.

ON 5'41:ET f,

12 7.2.3.7 A stud sling for the main closure studs shall be provided. The atud aling shall include a swivel and counter-weight spring to support the weight of the stud during turning of stud into vessel flange. Studs are to be provided with a wrenching surface accessible when suspended on sling.

7.2.3.8 All main load-carrying.threads and spherical washers shall be assembled only after cleaning, gaging, and lubricating. In no case during fabrication or testing shall these parts be assembled without lubricant. Only thread lubricant approved by the Buyer shall be used .*

7. 2. 4 Flanges 7.2.4.l The top head flange surface shall be machined or the area around each stud hole spot faced. Spot facings shall be complete and extend beyond washer O.D. to accommodate maximum eccentricity of stud in head flange bolt hole. The top head-flange surface, with or without spot facings, must accommodate and provide proper bearing area for the stud tensioner feet.

J.3 Nozzle Ends

• 7.3.l The ends of all nozzles other than flanged nozzles shall be prepared for welding in accordance with Drawing 107C5305. Nozzle safe ends are considered to be part of the vessel, not part of the connecting piping but in no case shall the safe end wall thickness be less than the wall thickness of the connecting pipe.

7.3.2 Where thermal sleeve nozzles are specified to a nominal size, the size of the pipe through the nozzle as well as the nozzle external end shall be the nominal size specified for the nozzle. Thermal sleeves shall be supplied by the Seller.

7.3.3 The Buyer will furnish information on the wall thickness, t , of all piping connections and will set the inner bore diameter incYuding tolerances and allowances of the connecting piping will follow ASA Standards. The Buyer will use the formulas and allowable stresses of B31.l for establishing the required piping wall thicknesses.

Nozzle safe end wall thickness shall be governed by Drawing 107C5305 and will in general be greater than required by Section III *

• ISSUEO:

MAR - 5 1969

1-21 GENERALt} ELECTRIC .

ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPEC1F1CAT1ON SPEC:. NO. 21Alll2 ,.EV. NO. 6 SM .. 0. 19 C:ONT ON SMEET 20 9.0 FABRICATION 9.1 Procedures 9.1.l The Seller shall submit for the Buyer's approval, all of the following procedures and procedure specifications:

9.1.1,l Heat treatment procedures for all thermal processes exceeding 800°F after the mill rolling or forging or foundry casting oper-ation.

9.1.1.2 Forming and bending procedures for all forming during fabrication subsequent to mill forging or rolling or foundry forming and cladding.

9.1. 1. 3 Welding and weld repair procedures including temporary welds as required in accordance with the ASME Code,Section IX, Paragraphs Q-10 and 11, and QN-10 and 11,Section III, Paragraph N-540.

9,1,1.4 Method of qualifying welding procedures and performance, if other than ASHE Code,Section IX and ~II

• 9,1.1.5 Repair procedures for major and minor defects as define*d in Paragraph 9.4.

9,1,1.6 Drawings showing location and preparation of test specimens, including specimens required in Attachment B.

9.1.1,7 Fabrication schedule including the detailed sequence to be followed in fabrication of the vessel.

9.1.1.8 All cleaning procedures, preserving procedures and a list of cleaning agents and preservatives together with their chemical content which shall be used during fabrication and in preparation for shipment. In lieu of a complete chemical analysis, the Buver shall accept a report which states the chlorides, fluorides and sulfur content. Other harmful elements should also be reported.

9.1.2 All work by the Seller or his sub-suppliers shall be performed in accordance with Buyer approved drawing, and fabrication and test procedures.

9.2 Material Cutting 9.2.l Stainless steel and carbon steel shall be cut to size or shaped by machining, shearing or thermal cutting .

1-27 GENERAL- ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION SPEC. NO. 21Alll2 REV. NO. 6 SM NO. 25 CONT ON IMEET26 10.3.2.2 Lav-Alloy Steel Nozzle Forgings Specimens, as fabricated, shall be taken from locations per ASME Code,Section III, N-313.2 (d) for forged nozzles. At least 2 tensile, 3 Charpy~V and 2 drop weight specimens shall be tested for each heat and heat treatment charge, except that nozzles with wall thickness of less than 4 inches and outside diameter less than 12 inches shall not require drop weight testing. The material shall meet the requirements of Paragraph 10.3.l.

10.3.2.3 In addition to the tests required by the ASME Boiler and Pressure Vessel Code, longitudinal specimens (parallel to the primary rolling direction), as-fabricated, shall be taken from the l/4T x T location. At least 2 drop weight specimens shall be tested from the top end (top as determined by ingot pouring) or each mill rolled plate and each heat treatment charge. The material shall meet the requirements of Paragraph 10.3.1. Additional drop weight specimens shall be required for NDT temperature determination per Paragraph 10.3.1.2 for plates located opposite the center of the core.

10.3.2.4 Castings Tangential specimens, as-fabricated, shall be taken from loca-tions per ASME Code,_ Section III, N-313.2 (d). Castings 1000 lb. weight and under shall have a total of l tensile specimen, 1 metallographic specimen, and 3 Charpy-V and 2 drop weight specimens, tested for each heat and heat treatment charge. Castings over 1000 lb. weight shall have a total of 2 tensile specimens, 2 metallographic specimens, 6 Charpy-V and 4 drop weight specimens tested from which 1 tensile specimen, 1 metallographic specimen, 3 Charpy-V and 2 drop weight speci-mens shall be taken 180° apart and/or diagonally opposite. The metallographic specimens shall be for reference only. Additional drop weight specimens shall be required for NDT temperature deter-mination in accordance with paragraph 10,3.1.2 if the casting is located 1D the core area. The material shall meet the require-ments of paragraph 10.3.l *

• I"MAR' -5 196~

1-36 PURCHASE S" ECI FICA TION GENERAL - ELECTRIC ATOMIC PCWER EOUlPMENT DEPARTMENT s1>e:c. NO 21Alll2 3.5 CONT Re:v NO ON SMEC~

6 F

• 12.2.2.s Instructions and parts list shall be clearly legible and prepared on good quality paper; carbon copies and tissue copies or other flimsy material are not acceptable. Multiple page instructions shall be securely bound.

12. 2. 2. 6 If a standard manual is furnished covering more* than the specific equipment purchased, the applicable model (or other identifica-tion) parts and other information for the specific equipment pur-chased shall be clearly identified.

12.2.3 Photographs The Seller shall provide the Buyer with sets of progress photo-graphs of the ,,easel at each significant stage of fabrication.

One set shall consist of one negative ~nd three glossy 8" x 10" prints.

12.2.4 Engineering Schedule 12.2.5 Fabrication Schedule 12.3 Records The Seller shall maintain records of all material qualifications, all weld and weldor qualifications and all process qualifications required by this specification and the material specifications. In addition, t~e Seller shall maintain records of all tests and inspections (e.g. - ultrasonic, radiography and hydrostatic). A list of the records shall be submitted to the Buyer on completion of the job. The Buyer shall be able to obtain certified copies of such records for a five-year period-, Where the Seller considers the actual test records to be proprietary, he shall submit certified reports containing all pertinent test data excerpted from the actual test reports. These certified test reports shall also be available for a five year period.

ISSUEO:

MAR - ~ 1969

1-37 ATTACHMENT A 1 INSTRUCTION MANUAL, DRAWING & DATA REQUIREMENTS

• Review or approval of drawings, procedures, data, or sµecilications by the buyer with regord t:, general design and co,.trolling di,.,eniions does not canstitut" acceptance of any designs, materials. or equipment which will not fulfrll the functional or pf'rformance reauir"'"""ts est ab Ii shed by this speci !i cation and the purchose contract. DocL.ment s and drawings submitted shall be black Iine and of n au al i ty which will produce readable-prints when microfilm.,d (35 mm) and blown bock on a conv,.ntronal 18

• 24 printer viewer sueh as Fdmac 200 or ltee. Send oil documents ond drawings to L.L. Kleinhessel;nk, GE, APED, (with eopy of tronsm,ttol to .I.PED Buyer, es ,nd,:ateravina* and Dlmenalona CB&l '

Load* - Pre11ure, Temperature Seinlic WelJht CE ..

Jet Poree* n Analytical Model Cl&I * ~

Selemic Analylle CB&l 1/4\\

'l'hennal Analyeie iL GE .. ~

Streu Analylla

,;-atigue A.nalyai 1

PH,PB,PL,Q r

Cl CE

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,- E UJ N

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N Review and Approval

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Ui:.SIGN ANALYSIS Sci1EDULE FOR REACTOR PRESSURE VESSEL FOR MONTICELLO POWF.R STATTON

Ro PENE'l'llAnOH SCHEDULE ASKE Sec tlon Responaible !fil  !!!!

ra1k1 III Streu Category Party A s 0 N D J , H A H J J A 8 (Total Ta1k TlJDe)

>raving* and Dimen1 ions CBl.I ~

3izing, ASHE Calculation CB&l Load1 - Se iamic

• Scram Weight,, Pre11ure Flow llatea and n Temperatures CE f lbermal Analysia -

Steady State Tranaient CE I*

Stress Analysia -

q-imary PH,PL,PB CB&l

• Primary and Secondary PK,PL,PB.Q GI J-.

n 0 :id Fatigue Analy.Ji1 PH* PL ,P1 .Q,P GE I+ .~~.

Review ftnd Approval CE MAN .....I ii:.

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• IIACTOI PUSSUU VESSEL roa OISIGB AllilYSli UIDULI roa lllffICILLO IQID STAnm DZUS*

SCBIOOLI ASIIE Section Re1ponaible .ill! .ill!

Ill 8trH1 Party Taak1 Cateaory A s 0 N D J r M A H J J A s (Total Talk Time)

I Drawlna* and Dimenaiona l

CB&l Sizina, ASME Calculationa CB&l Loads -- Pipe Reaction Seiamic

- Pre1eure1

- Flow Rate and Temperaturea GI l n thermal Analy1i1 -

- Stea.Jy State

- Tran1ient GE t

• Streaa Analysis

-Primary PM,PL,PB CB&I  !

- PTimary and PM,PL,PB,Q GE -

Secondary Fatigue Analy1i1 GE

'MLa*Q,r --I n o::-:i r:, Ill

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Review and Approval GE i u,

Note breakdown of no11le1 to be an.aly1eJ p*r thi1 1chedule on page 6,

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PRESSURE VESSEL REPORT MANUFACTURER'S DATA REPORT AND VESSEL CERTIFICATION L

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MONTICEUO GENERAL: EI.£CTR1C CO.

AP£ti - S.AN '.OSE VPf:!, //1/-..JSS"* I

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20. Resu lts of Char py V-No tch Impa ct Test s per Par.

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2-18 CHICAGO BRIDGE & IRO:K COMPANY P.O. BOX 13308, MEMPHIS, TENNESSEE 38113 901 947-3111 DATE: March 11, 1969

REFERENCE:

BOILING WATER NUCLEAR REP.CTOR VESSEL 17.167' x 63.167' INS. HDS.

1&'\;UFACTU~F.' S SERIA:i:. KO. B-4697

• lONTICI:LLO PROJECT, ~1ONTICELLO, !-U~N.

G.E. CO. P.O. 205-55582-I CB&I CONTRACT 9-5624

SUBJECT:

WELDING CERTIFICATION TO WHOM IT MAY CONCElli~:

This is to certify that welding of the above referenced vessel was performed in accordance with t.he AS)lE Coc.e,Section III, 1965 Edition, with Addenda through Summer 1966, and General Electric co. Specification 21Alll2 Rev. 5, Paragraph 9.3 and also approved CB&I Co. procedures

,;-J-PS-1 Rev. 2 WPS-18 Rev. 4 WPS-35 Rev. _

WPS-2 Rev. 2 WPS-19 Rev. 3 WPS-36 Rev. 0 WPS-3 Rev. 0 WPS-20 Rev. 0 WPS-37 Rev. l WPS-4 Rev. 2 WPS-21 Rev. 1 WPS-38 Re~,r. 1 WPS-5 Rev. 0 WPS-22 Rev. 1 WPS-39 Re\'. 0 WPS-6 Rev. 2 WPS-23 Rev. 1 WPS-40 Rev. l WPS-7 Rev. 2 WPS-24 Rev. 4 i;-JPS-41 Rev. Q w-PS-8 Rev. 0 WPS-25 Rev. 0 WPS-42 Rev. -

WPS-9 Rev. 0 'vl.TPS-26 Rev. 0 WPS-43 Rev. O WPS-10 *Rev. 0 WPS-27 Rev. 2 WPS-44 Rev. 3 i;\1PS-ll Rev. 0 WPS-28 Rev. 5 WPS-45 Rev. O WPS-12 Rev. 2 WPS-29 Rev. 1 WPS-46 Rev. 0 WPS-13 Rev. 3 WPS-30 Rev. 2 WPS-47  ;'* V. 0 WPS-14 Rev. 2 WPS-31 Rev. 1 WPS-48 r:..::v. O WPS-15 Rev. O WPS-32 Rev. 1 WPS.-49 Rev. 2 vi'PS-16 Rev. 2 WPS-33 Rev. 0 WPS-50 Rav. 2 WPS-17 Rev. 1 WPS-34 Rev. 1 WPS-51 Rev. 0

2-29

• CHICAGO BRIDGE & IRON COl\t!PANY P . 0 . B O X 1 3 3 0 8 , M E M P H I S , T E N N E S S e: e: 3 8 1 1 3*

eo, s.:.7-311, DATE: March 11, 1969

REFERENCE:

BOILING WATER NUCLEAR REACTOR VESSEL 17.167' ID x 63.167' INS. HDS.

l*1ANUFACTURE~' S SERIAL NO. B-4697

.MONTICELLO PROJECT, MONTICELLO, l~I:i'.u-l.

GENERAL ELECTRIC CO. P.O. 205-55582-I CB&I CONTRACT 9-5624

SUBJECT:

FABRICATION TEST PROGRAM CERTIFICATION TO WHOM IT MAY CONCER.°'i:

This is to certify that the fabrication test program was performed for the above referenced vessel in accordance with Attachment "B" Rev. 3 of General Electric Co.

Specification 21Alll2 Rev. 5, Paragraph 2.0 titled:

"Fabrication Test Program" and using specimens cut fro::;

plate of same heat as plate used in the vessel. These specimens were cold formed to CFP-1 Rev. O which correspor.ced to the cold forming performed on the plates of the Reactor Vessel.

Paragraph 2.2.3 of Attachment "B" was complied with by subr:i.itting "For Information Onlv" the test reoorts for 80% T tensile test on Chicago Bridge & Iron Company-letter BBZ-198 dated 8/9/68. The above test reports were compiled by t~e University of Illinois. Chicago Bridge & Iron Company took exception to Paragraph 2.2.3.2 of Attachment "B" and subsequently, agreement was reached with General Electric Co. during the meetings held in San Jose, California, August 22 through August 25, 1966, in the following manner:

Page 10 Item G-1 "The 80% T dia. test specimens from as formed plate r::ay come from rolled plate with girth (category B) sea~ so

• that separate welded on grips are not necessary. T~is interprets differently the plate as regards Paragraph 2.2.3.2 of the GZ S?ecification, but Paul Herbert and .Bud Vancot~

(both were in the meeting when this was discussed) agreed that this was acceptable."

2-30 CHICAGO DRIDGE & IRON COMPANY

SUBJECT:

FABRICATION TEST PROGRAM CERTIFICATION Attachments:

1. CB&I Drawing T-4 Rev. 4, Results of Charpy V-notch impact tests as per Paragraph 2.2.4 of Attachment B.

2. CB&I Drawing T-5 Rev. 3, Results of tensile tests as per Paragraph 2.2.2 of Attachment B.

CHICAGO BRIDGE & IRON COMPA..~Y

-<Ad"",;x.tt-;-lf--/

E. E. VARNUM

NUCLEAR QUALITY ASSURA..~CE ~-~NAGER

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11-26-66 fill NO.

COAUSVILll, PA. DAU, 1543

10. Chice.30 Bridge & Iron Co. CONSIGNEl, TEST CERTIFICATE Mr. G.H. Putman,P.A. Mill OI0fl UO. CUSTOMER P.O.

P.O. Box 277 Saine

• J; .-1 Birmingham, Ala. *35202 43211-1 5624 ,:,'\

MB 112166 E?~ Boyles, Ala.

SPfCIFICATIONS, .

A-533-65 Gr.B Clasa*l Mod.by C.B.& I. Spec. MS~l DTD 8/25/66 Fbx. 80000 Cont.# 9-5624 HND un O. K. HOMOGENITY TUT O* K CHEMICAL ANALYSIS Mfll NO. C MN p s Cu S1 N1 Ci Mo V T, ll .I u.l'a1.n .::,.1 ze 01946 A0998 22 20 L35 1.27 010 008 015 017 ~a ,g 47 49 V.I.

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PHYSICAL PROPERTIES 1 ; SlAI YIIID TINSlll " lj_ONG. IMPACTS

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Plntes healed 16 5-167 *F., eld i hr.pr in*h mini and ,1,ate~ uenched ~ under.

400°F. by !old in,;. in uiter rbr at east -1/2 minutf 3 per r inch f thickne s Affirmed and subscribed before me then ten.pcfed 12f0-125 * °F.

• held l hr. pl-l:r in h min, and tir co led. *' this day 0 ~AR 3 1967 19 Tests fromlheat reate platis str ss re ieve by-~eatin?I with n a rate f --

640F per hl'. to 125-1 75*F. held 50* hr

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?-1111 inspe tion y ago B" idge Iron Co. I I Ck. ~Llt.l We here'.ly certify the above figures are correct as contained in the records of the company .

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4-2 SWIL"!la ry Repor t Monti cello-- NSP Reacto r Vesse l CB&I Contr act 9-5624

~ ' , - ~3,-/

~-,-/ ... - J* **

• Gener al Elect ric P. o. No. 205-55 582-I Reacto r

4-4 OAK BROOK EHGINEEiW-IG SU~.Y~rtY*REPORT--INTRODUCTION SUM.~RY OF THE MON?ICELLO STRESS REPORT The stress analysis for the Monticello Reactor Vessel has been performed in accordance with the General Electric Purchase Specification 21Alll2, Rev. 5 and Section III of the ASME Code. The stress report has been certified by a registered professicnal engineer who is experienced in pressure vessel design.

The following paragraphs summarize the stress results for the various components of the Monticello Reactor Vessel.

For each component, the calculated stress intensities for each stress category, primary membrane stress intensity, local membrane plus bend~ng stres~ intensity and primary plus secondary stress intensity range, are compared with the appropriate Section III, ASME Code allowables. The specified fatigue cycles and Code allowable cycles are given wherever appropriate. This Summary Report is being submitted as required in Paragraph 6.8 of the General Electric Purchase Specificat~on mentioned above .

• s.,b 1ect tiCN':'ICELLO REACTOR VES 3EL n ***

Cont. 9-5624 Dote_ _ th  ;;-.* s,, '-=--:

~

4-6 .

CHICAGO 8Ri0CE g, IRON COMPANY OAK BROOi< EHCINEERINC The basic stress intensities in the main closure flanges and the adjacent top head and cylindrical shell per ASME

.:. ie,Section III, N-414, are as follows: the maximum primary membrane stress intensity in the top hemispherical head is due to preload plus pressure loading at 1250 psi and occurs 4.379 inches above the -flange transition sec-tion. Its magnitude is 28,620 psi (page I-S1-64). Due to the influence of the head to, flange discontinuity it is classified as a local primary membrane stress intensity.

It is seen to be less than 1._1 Sm= 29,370 psi.

The maximum primary plus secondary stress intensity range occurs during the startup transient at the hemispherical head to top flange junction, and has a magnitude of 55,320 psi. The allowa~le stress intensity range in this case is

• 3 Sm= 80,100 psi.

The maximum primary membrane stress intensity in the cy-

• lindrical shell below the shell flange is 29,560 psi (page I-S1-65). This stress intensity is due to the pre-load plus pressure loading at 1250 psi, and is located 15 inches below the cylinder to shell flange junction. As the width of the band in which 1.1 Sm= 29,370 psi is ~x-ceeded is 8.2 inches, and the allowable width is .5 Rt =

11.746 inches, this stress intensity is classified as local. For the location of the above stress intensity band see the attached sketch. The allowable stress inten-sity for local primary membrane stress intensity is 1.5 Sm=

40,050 psi.

The maximum primary plus secondary stress intensity range in the shell flange is 47,110 psi (page I-Sl-67). It Subject )1Qj),:'1"T,?T i Q B"l:'ACTQB VESS-:;I, Cont. 9-5624 Dote _ _ By AEE Sh,_4

_of 34 a .... u .... n,. ** _ Rev.No). _ _ C.::re _ __

4-7 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING occurs during the startup transient and is located on the outside of the shell flange to cylindric al shell junction. The allowable stress intensity range in this case is 3 Sm= 80,100 psi.

It was found that all the requireme nts of the ASME Code,Section III, Par. N-415.1 could be satisfied for the main closure flanges, and therefore no fatigue analysis of the same is required.

5 .. .3.:,

Subject MQNTICEI I,Q--REACTOB VP:-SSEI Cont.

o ..._

9-5624 Oote _ _ Bv AEE

~_._v._~~

_ _ _ _ _~,.

S;,,

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4-13 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENG!NEERING I

SPECIAL STRUCTURES DESIGN POINTS OF HIGHEST STRESSES FIG. 2 P01NT7-.-- ___ po1,vT30 POIN,3

(~ussEcnoN F)

Subjecc * ~ 1 *-**-****

• , ' : * , \

eont. _ _ _ oaie _ _ _ av _ _ _ sht _!.!._ .Jf_3_

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64 sso Checked bv _ _ one _ _ _ Rev No _ _ Oate _ _. Rew No._._ Cate _ _ _ Aw. No. _ _ _ c~te _ _ __

4-18 CHIC.AGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING

SUMMARY

OF STRESS ANALYSIS 3" CRDHSR NOZZLE In the safe end area, the maximum primary plus secondary stress intensity of 44,320 psi occurs at point 3, against an allowable of 48,000 psi at design temperature.

In the nozzle forging, the nozzle vessel junction {point

19) is the highest stressed point. Based on the stress index method, the maximum pressure stress intensity is 88,100 psi. To this is added the thermal stress inten-sity at steady state, which is 38,841 psi, giving a peak stress intensity range of 126,951 psi and an alternating stress ampli~ude of 63,475 psi, which gives an allowable number of cycles of 2000 against an expected 78~ cycles,

• based on the applicable design fatigue curve.

The points referred to above are shown in the sketch on the following page *

• 9-5624 MSM 16 ~* 3.;

Subject MCNTICE! ! Q REACTOR VESSEL Cont. Date-By Sh*

"-*- n.,*., _ _ _ Rev.Mo. _ _ Oa,o _ __

4-19 CHICAGO BRIDGE & IRON COMP.A.HY OAK B~OOK ENGINEERING T

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o e, Oil Oeclced by,_ _ _ Date _ _ _ _ Rev.Mo _ _ _ D11*----Rev,Mo. _ _ _ D o t * - - - Rev.Ho._Da**---

4-25 CHICAGO BRIDGE & IRON COMPANY OAK SROCK ENGINEERING

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e, OB Checked by_ _ Date _ _ _ Rev.Na _ _ Date _ _ _ Rev.No. _ _ oate _ _ _ Rev.No. _ _ Dote - -

4-31 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGli\!EcRING SPECIAL STRUCTURES DESIGN

- RgFuEL 1tJa Po,...,,

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4-32

• CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING

SUMMARY

OF STRESS ANALYSIS FOR STABILIZER BRACKETS The stabilizer brackets were analyzed for two loading conditions per GE Specification Drawing 886D482, Sheet

8. For loading co~dition #1 the bracket stresses were limited to allowable stresses per ASME Code,Section III. For loading condition #2 the bracket stresses were limited to the yield strength of the material.

The bracket design stresses and the corresponding allow-able stresses are as follows:

LOADING CASE 1 Actual Maximum Stresses Pure Shear Stress at Pin Hole= 15,238 psi Bearing Stress at Pin Hole = 21,642 psi Maximum Stress Intensity!

= 14,593 psi At Face of Shell Allowable Stresses Pure Shear Stress = 16,020 psi Bearing Stress = 42,300 psi Maximum Stress Intensity= 26,700 psi Subject MONTICELLO REACTOR VESSEL Cont- 9-5624 Dote _ _ &y AEE Sht.lQ..of -1.,i_

n,.** _ Rev.No. _ _ Date _ __

 



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4-36

• CHICAGO BRIDGE & IRON COMPANY

• su~_-;.Ry OF STRESS ANALYSIS FOR OAK BROOK ENGINEERING DRAIN NOZZLE The Code area replacement requirements for the drain nozzle have been satisfied.

The maximum primary membrane stress intensity is 6716 psi versus the Code allowable of 18,200 psi.

Subject MONTICELLO REACTOR VESSEL Cont. 9- 5 6 2 4 Oote~By JH Sht .2_! of j_L

o a, OB Cnecked by-_ _ _ oa,. _ _ _ _ Rev.No., _ _ _ Date _ _ _ _ Rev.No. _ __..onte _ _ _ Rev.No. _ _ Date - - -

5-2 MONTlCEtLO Mon~icello Rcc:.ctor Vessel, p~.-J..; lV-1

~NE.RAL ELECTRIC CO.

APEO - SAN JOSE iVPF# / (j I I -35ti:,- I I EPslf '.;?-1-1 IV.

4.1 Div:s:::oN OF WORK A significant portion of tile vessel fabrication was f~r=~::::-.cd in t:.e shop, just as would be done for a shop f~~ica~cd v~~s~~-

This work was in accordance wit....,_ t."le ASME Coc:e and G.:::. f~;:ica.~io::

~uality control re~uirements. The balance cf functional require.~ents of t.,;.e vessel.

A ty~ical site assembly area is shown in :'is;ure IV-1 an artist's rendering of t.;.e completed composite reactor a.~d contai:-.rnent vessels is shown in Figure IV-2.

4. 2 S~OP FABRICATION Ai.'ID SUBASSE!;f.3LY WOR.'t(

As much fabrication a.~d subassembly work as possi~le was perfor:ned at C3&I 1 s Bir.ningharn, Alabama, a.~d Greenville, Pen."'l.sylvania, manufacturing plants. The overall job economics favored t.,.is approach because of the convenience of overhead handling equipment, utilization of t....,_e existing shop laoor pool

~d facilities for machining, heat treat.~ent, etc. Restric~io~

on shipping di.~ensioris (not weight) was t.,.e dete=~ni::g fac~or ~::

considering 'how much of t.,;.e vessel assembly work could be pe:.:o=::-.<:!..:i prior to shipment.

5-4 Monticello Reactor Vessel, Fage IV-3 and qualified personnel in accordance with Code specifications.

Certified mill test reports and all quality control measures were reviewed by CB&I engineers to assure compliance with material specifications.

After the plates were marked and flame cut to approximate size, they were pressed to shape on a 6,000-ton hydraulic press, designed by CB&I. Any minor deviation from curvature tolerances found in checking with box templates and sweeps were corrected by sizing the plates on the press. Each plate was then marked and cut to size and edges beveled with semi-automatic cutting torches. To insure proper dimensions and alignment, shop assembled weldments were fit-up and match-marked prior to shipment to: the jobsite for assembly and welding together.

The bottom head was shipped in two sections consisting of (1) the knuckle course of plates with the stub skirt attached and (2) the dollar plate assembly. The dollar plate assembly was predrilled in the shop to accommodate the 121 control rod drive sleeves. The initial holes were drilled to approximately 5 inches in diameter. These holes are large enough to accommodate a boring bar cutting assembly that was used in place for the final boring of the sleeves at the site. Because of the availability of machining equipment, this assembly and predrilling work was performed at CB&I's Greenville, Pennsylvania, plant. The final drilling of the holes was performed in place at the site.

5-5 Monticello Reactor Vessel, Page IV-4 The bottom head knuckle course shop weldment was positioned and two overlay weld metal build-ups were applied (see Figure IV-4) in the two areas where the shroud support was welded to** the bottom head. These weld build-ups were shop machined to the contours shown in Figure IV-4.

The Monticello vessel shell was made up of four rings, approximately ll feet wide. Each ring was made from two formed plates. The half ring sections were temporarily welded toget.~cr and placed on a roller bed. The ring was preheated and the overlay weld metal deposited with automatic equipment similar to that shown in Figure IV-5. All shell f~ttings were shop installed. Postweld

. h e a t treatment was performed and inspection of the overlay weld deposit and insert seams was made after cool-down.

The shell and head flanges were shipped directly to the site as rough machined, non-drilled, seamless forged rings from the Ladish Company plant in CUdahy, Wisconsin. The weld ends were prepared at the forge works (Ladish) for fit-up and welding to the adjacent No. 4 shell ring and top head weldment. -This top head assembly was shipped in one piece. It was welded together from six knuckle plates and a one-piece dollar plate assembly, as shown in Figure IV-6.

The internal _shroud support was completely shop fabricated,

• including preliminary machining, at Greenville and shipped as an integral ring assembly to the site where it was welded in place to the bottom head. The final machining was completed after welding.

5-9 Monticello Reactor vessel, Page IV-8

• t.~e bottom head to No. 1 ring girth seam until the No. 1 to No. 2 girth seam was ready for postweld heat treatment. At that time, the two rings were postweld heat treated simultaneously in the ternporar1 furnace. Steps (b) through (k) used for site subassembly .(Paragraph 4.3.1) were used for assembly in place.

Non-destructive testing methods in the field were the sa.~e as those performed in the shop. Radiography was performed utilizing a 75 to 100 Curie Gamma source with appropriate shieldingo Usage of the source was in accordance with the applicable Federal and State regulations.

• Concurrent with erection of the vessel shell, the vessel top head weldment was fit and welded to the cover flange in the a~sembly yard area. After completion of all the welding, postweld heat treatment and examination steps, the top head was positioned for drilling the 5-1/4-inch diameter hold--down bolt holes, as shown in Figure IV-11. With the cover in this same position, the grooves for the two 1/2-inch diameter stainless "O" ring gaskets were machined with the portable CB&I equipment as depicted in Figure IV-12.

After the No. 1 and No. 2 girth seams were postweld heat treated, the temporary furnace was converted into an air-conditioned and ventilated work room around the bottom head and No. 1 shell ring. A temporary cover was installed above this work area so t.~at

. t h e balance of the vessel could be erected without interfering with

5-11 Monticello Reactor vessel, Page IV-10

• insure complete removal of the TSP solution.

until the effluent conductivity was 5 micro-mho/cm.

The rinsing continued Upon completion of the initial cleaning, the vessel was filled with heated deionized water and tested per the requirements of t."'le ASME Code. Upon completion of the overload pressure test, the vessel head was removed and service gaskets installed. The vessel head was then replaced and a leakage rate test was performed between the double 11 0 11 ring seals at the design pressure.

Upon completion of the hydrostatic test at design pressure, the test caps were removed from the vessel and replaced wit."'l temporary covers. The vessel was once again nigh pressure blasted

. i t h deionized water. After drying the interior surfaces of the vessel, the vessel was sealed to prevent entry of dirt or other foreign materials.

4.4 REACTOR VESSEL QUALITY CONTROL 4.4.l Objective The quality control for the Monticello nuclear reactor was directed by a Quality Control Manager with the assistance of Quality Control coordinators. The primary objective of this group was to coordinate CB&I's many quality connected functions into a system which assured that the reactor ve~sel produced would meet the suality requirements and to document the fact that these quality

• equirements were met.

5-13 Monticello Reactor vessei, Page IV-12

• Each item or piece of material received at the shop or at the site was covered by a Work Order and Traveler *card which listed, in sequence, all of the operations and inspections which that particular item or piece underwento Each operation or inspection was given a unique reference number so that it could be referenced to t.~e report of record. Each operation was referenced to the applicable approved procedure with special notations for witness points or points beyond which further progress was halted until clearance was obtainedo Provision was made for sign-off by the supervisor after the operation was completed, by*the inspector after the inspection was performed, and by the Q.c. coordinator as well as the customer's o~c~

.epresentative after each item or piece was reviewed and accepted.

4o4.4 Documents and Records In addition to the usual records required for presssure vessels built to Section III of the AS.ME Code, a complete thermal history of all parts and a quality control spread sheet of this vessel will be maintained for the specified time period. Written non-destructive test reports were prepared for each radiographic, ultrasonic, magnetic particle and liquid penetrant inspection.

Also, welders' performance qualification certificates and test*

results*are available for review.

The same record, report, inspection or process procedure was

. s e d for similar operations regardless of whether performed in the shop or at the site. Traveler Cards, Thermal History, and Spread

5-14 Monti cello React or vesse l, Page IV-13 Sheet s were initia ted in the shop and were carrie d throug h to t.~e compl etion of the job.

---

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

6-2 INDEPENDENT REVIEW OF STRESS AMLYSIS REPORT I~ accordance with a suggestion by the USAEC Advisory Committee on Reactor Safety (Monticello ACRS Letter, April 13, 1967, AEC Docket #50-263), the Reactor Pressure Vessel Stress Analysis Report was reviewed by independent experts. This s t_udy has been perfonned by Te 1edyne Ma teri a1s Research Division of the Teledyne Company, Waltham, Massachusetts.

Teledyne's summary letter concerning their review is included herewith as Exhibit 6 of this report.

7-5

• -GEHERAL() ELECTRIC NUCL~R ENER~Y :IIVISION REV.

22A5541 2

SH. NO. 4 4.3 Design 4.3. l Themal S1eeve Reactions. The Inconel thennal sleeve 3hown on Drawing 112D1693 will be ins~al1ed with a coid nominal interference of 0.010 inch across the diameter. The effects of the ther.na1 sleeve on the safe end and nozzle shall be con-sidered in the desigr, ar.iilysis. The geometry is shown in Figure 2 and on Drawing 769E367.

4.3.2 Design pressure\~ 12so*psig. Nonnal o~erating pressure is 1111 psig.

4.3.3 Design temp~rature is 575°F, Nonnal operating temperature is 546°F.

~ 4.3.4 Nonnal operating cond:tion pipe reaction loads are !ihown in Figure l. There are no upset. emerge~cy, or fault pipe loads specified for this desion,

- 'l. ~

,r::i"J*

Fe 3,0 kips 19.26 FL 5.7 kips F..

i"-

3.2 ldps Mc 156.0 in.-kips M. 336.0 in.-kips Mz 348.0 in. -kips Loads can be in either direction for all val~es shown.

Figure 1 4.3.5 Seismic loads are included in the pipe reactions.

4.3.6 Corrosion AllO'ttance. All exposed exterior ferritic steel surfaces of pres-sure contarning parts snail have a corrosfon allowance of 0.032 inch in 40 years.

All ferrftfc steel surfaces exposed to reactor coolant shall have a corrosion allowance of 0.063 inch fn 40 years~

4.3.7 Desiqn Life. The design life of this repair shall be not less than 24 months.

If design life 1s extended beyond 24 montt,s, then additional analysis, according to this specification, is required.

7-10 GENERAL() ELECTRIC N~ JCLEAR E*~ERGY Cl VISION IREV, :2A554 l SH.NO. 9 1.0 0.9 o.s N

y 0::

...~

11,J 0.7

.; \I g 0.6 -- -- ~-. -- .

r:

~ o.s I\

llrJ \

t... 0.4

~

~

.... 0.3 1\

0.2

\ \

.125 \

0.1 0 0 0.1 0.2 o.3 o.4 o.s o.6 0.1 o.a

.45 LEAKAGE FLOW VELOCITY (FT/SEC}

FIGURE 4

7-12 GENERAL O ELECTRIC NUCL!Arc E!.. ERGY DIVISION 22A5541 SH. N0.11 AEV, 2 4.5.1.4 Thd temp~rature trans1ent is shown in tabular fonn below:

TEMPERATI.;RE TRANSIENT IFluid Fluid Fluid State

!Temp. Start End of Fluid Vessel Rate Temp. iemp. fluid VelocHy Pressure Notes lOC F/hr 100 540 Water 0 1111 psi g Followed by Step To C 100 100 Water 5 ft/sec 1111 ps1g Followed by Step To

?SO F/hr 260 376 Water 5* ft/sec 1111psig Velocity changes linearly 5 ft/sec to 20 ft/sec 4.5.2 Unstable Flow Cycling. During reactor startup under low power conditions temperatures 1n the top ha1f of the feedwater n~zzle safe end and thermal sleeve shail be assumed to fluctuate over a 250°F temperature range from (100°F to 350°F) as shown on FigurP. 5, for 1: of the operating time, i.e .* 88 hours0.00102 days <br />0.0244 hours <br />1.455026e-4 weeks <br />3.3484e-5 months <br /> per year. This cycling i~ in addition to the temperature cycling in the nozzle defined in Paragraph 4.5.1. _This cyclin*g is due to unstable flow when t~e ,~ed-

~t,,QI ~GI I

,.. ~ . -

-'1'-1111¥1 ,. *. ._._ * ._. v**- ***'- ' - h * *

~=ter f!vw re~-: is tc,o le-.: ~~ k~C?~ th:! hct ~;ion _,._ fl:.:id ~~*::;:t o:.:t o'f !~!

- ***- .JW**-u* *--~* * '-'*

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

and nozzle remain at 100°F during this cycling. The heat transfer coefficient, calculated according to the procedure given in Paragraph 4.4.4.2 at 25: rated feedwater flow, is to be used for the top and bottom for both cold flushing and hot back flow. The transient stresses may be calculated by assuming an axisymmetric model with boundary conditions for the top half of the nozzle.

The stresses due to the top-to-bottom temperature may be upper bounded by assuming that the vessel shell, nozzle forging, and attached piping are rigid and u~ing an equation of the fonn Ea (TToo - TBottom) for the safe end.

where ax

• axial membrane stress in safe end, use upper sign for top, lower sign for bottom E

• Youngs Modulus a

• coefficient of thermal exp,,nsion TTop* mean temperature of top half of safe end TBottom* mean temperature of bottom half of safe end

7-13 GErJ ERAL ( ) ELE CTR I C 22A5541 3H, N~ 12 NUCLEAR ENERGY DIV' llON

.... ---*----***-----*-*--r-- ----t*- -- -

350 - *. -J **-***-***'* - *I- -*. - ---

1--

LI,.,

0 c:: zoo

• ~

l,W i:ij e

LI,.,

150 E. - .:J*** .. -------

I-

< 100

c Q.

....0 so 60 120 180 240 360 nME (SEC)

FIGURES

7-16.

22.AS541 SH. NO.

15 NUCLEAR 1:::t.E:i!G'f OIVISION AEV. 4

.---....-.-.-.----.-.-..**----..-.-.---.--:-:-: ::::=-** .. ::::..-::.-.-.-.--._-.-.-... :=---=-. :-:-:.:::==**

..... ~- -***** ..................... ****** .............. !........... --- . -** ........... :.. ....... _.

(~;;;;;;i;;;;;;;::;:;;;;:;:;i: :~~;:;;:~:i~~:;:;:; ~:;

l: ;: : ; i:: :;:;:i::;::; ;i:~; ~~:

: :: : : : ~::::: :: ::~:: :* ::: :: : i ~:::::: :: :: ::::: ::: : : :::::: :::~:: ::: :: : : : :: :: :: :: : : :.  : : :::: :: : : : : : N N

0 N

C0

>D u

. -*--t:

l:J

~

...u

~

N 0

,-l

~

3 0

f; Ill C0 0

. c.,

5

~

.;&J

,-l

>D 0

C N* "'

• f"'4 C .

f"'4 0

.,, ~rtdil'lilH OtU'OJ'.:> OicMI

• -G-EHlffJ\L ~ ELECiillC NUCLEAR EtlERGY DIVISION Rev.

22A55"1 2 Final SH. NO. 10

--+---....-~---+--:-:. .-:': -;'-----,-:~,-,::**I * *:; I 1,..1.,,;-;....,*;..;..i,;\'\,_p*..._:  !--t-:-/;*~*

i i' _ : , ALLOY STEELS I I  !.

I l I I I L

4 10* 101 102 103 10 105 . 106 10 7 108 109 1010 1011 10 12 CYCLES E

• 30x10 6 ps1 FIGURE 7

8-4 NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A7454 REV 1 SH NO. 3 TABLE OF CONTENTS

1. ABSTRACT

2.

SUMMARY

AND CONCLUSIONS

3. DESIGN REQUIREMENTS

4. ANALYSIS 4.1 Thermal Transient Analysis 4.1.1 Thermal Model 4.1.2 FeedTater Nozzle Heat Transfer Coefficients 4.1.2.1 Cool-Down Transient 4.1.2.2 Beat-Up Transient 4.1.2.3 Normal Operation 4.2 4.1.3 4.1.4 Feedyater Nozzle Annulus Fluid Temperatures Thermal Analysis Results Stress Analysis 4.2.1 Selected Loc.ations for Stress E.valuation 4.2.2 Thermal Stress Analysis 4.2.2.l Selection of Times For Stress Evaluation 4.2.2.2 Thermal Stress Analysis Results 4.2.3 Mechanical Load Stress Analysis 4.2.3.1 Applied Mechanical Loading 4.2.3.2 Mechanical Load Range Calculations 4.2.4 Pressure Stress Analysis 4.2.4.l Pressure Stress Analysis Results 4.2.5 Total Primary Plus Secondary Stress Range 4.2.5.l Thermal Stress Ranges 4.2.5.2 Nozzle End and Thermal Sleeve Load Stress Ranges 4.2.5.3 Pressure Stress Ranges 4.2.5.4 Total P + Q Range 4.2.6 Interference Fit Stresses NEBG.407 A (6/10)

8-5 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 4 BUSINESS GROUP PIEV 1 TABLE OF CONTENTS (Continued) 4.3 Fatigue Analysis 4.3.1 Stress Concentration Factors 4.3.2 Alternating Stress Range 4 .3 .3 Uaago Calculation 4.3.4 High Cyclo Fatigue 4.3.5 Accumulated Fatigue Usage 4.3.6 Total Fatigue Usage

5. RESULTS

6. REFERENCES APPENDIX 10 LISTING OF *~NO' APPENDIX 20 INTERGRANULAR STXESS CORROSION DIDEX CALCULATIONS APPENDIX 30 RECALCULATIONS REQUIRED DUE TO MANUFACTURING DEVIATIONS

• NE8G-a07A (6/10)

8-9 NUCLEAR ENERGY BUSINESS GROUP GENERAL fl ELECTRIC 22A7454 REV l SH NO. 8 W //ACAUON STEEL (SA-50-S - CL.1) - Ori1bal Safe End I\\\ Sc.umoN STEEL CSA-333 - GD.6> - Piping R\_ §  !CARBON STEEL (SA-508 - CL.2) - Nonle V/ /lCARBON STEEL (SA-350 - U-2) - Safe End I !STAINLESS STEEL (SA-351 - CF3) - Thermal Sleeve l03 0 == Potl)T,S FOR. STA.ESS CRL.C:U.L.Ai'1o~s L

• FIGURE 3.1 NllaG 107A NOZZLE, SAFE END, AND THERMAL SLEEVE GBOMETB.Y

R-11 22A7454 10 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP *uv 1

• Loading Nozzle Safe End Loads (Reference 6.1)

Forces in kips Moments in in-kips F F F M )I )I Condition ~ ..::I. ..! ~ ..::I. ..! R(i.n)

Design 1.54 3.15 2.28 387 .6 172.9 324.6 131.6 Nozzle

'A' Dead Wt.

Seismic

- 0.11

!:. 0.29

- 0.63

!:. 2.51 0.15

!:. 2.23 +

11.6 9.3

- 14.1

!:. 158.9

- 11.1

!:. 313.4 131.6 Loads Themal 0.02 0.16 - 0.21 - 12.0 - 12.l - 45.0 Nozz~e Dead Wt. - 0.07 0.18 - 0.04 7.0 2.1 7.3

'B' Seismic  :!:. 2.44  :!:. 1.97  :!:. 0.26  :!:. 376.0  :!:. 106 .3 * + 10.6 131.6 Loads Themal 0.82 - 4.34 1.37 267.2 - 66.7 1.4 Thermal Sleeve Loads (Reference 6.1)

Forces ill kips Moments ill in-kips F F F H M M R(in)

Condition ~ ..::I. ..! ~ ..::I. .J.

Design 2.5 0.6 5.7 2.4 2.0 0 103.0 Dead Wt. 0 - 0.3 - o.s - 1.2 0 -. 0 Seismic  :!:. 2.5  :!:. 0.3  :!:. 1.5  :!:. 1.2  :!:. 2.0 0 103.0

'lliex:mal 0 0 - 1.2 0 0 0 Hydraulic 0 0 - 2.5 0 0 0

• NEBG.a07A (6/80)

8-13 NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A74S4 REV SH NO. ll 1

• Thermal Sleeve Loads Forces in kips Koments in in-kips F

Condition l B -'

De I ign 2.57 3.124 5.7 Service Level 'C' 5.08 5 .39 6.0 Pressure Loads (Reference 6.1)

Design Pressure

• 1.250 psi Service Level 'C' Pressure

• 1,375 psi Loading Sign Convention 1..

\ Sign Convention applies to I both safe end and thermal

.,, ,,,, I sleeve loadings.

X R

NEBG-e07A (6/80)

8-15 NUCLEAR ENERGY BUSINESS GROUP GENERAL fl ELECT.RIC 22A7454 ii.EV 1 SH NO. 14

• • Stress Due to Thermal Sleeve Loads:

P = 2 .57 kip H = 3.124 in-kip F = 5. 7 kip z

m "" 3.124 + 2.57 (15.76) + 5.7 (2.36) "" 57.08 in-kip

= zM = 57,08 51 .98

= 1.1 ksi F

= ~ = 5.7 = 0 .30 ksi A 19.03 Total Stress a~ = 6,4-00 + 11,310 + 120 + 1,100 + 300 = 19,230 psi a = 12,800 psi 0

a = -1,250 psi r

• NEBG-807A (6/10)

8-18 NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A7454 REV 1 SH NO. 17 Section B Design Pressure Stress:

p Di 2t - 1,250 (10.875) 2 {0.531)

• 12. 800 psi ae a~ = = 6,400 psi 2

a * -1.250 psi r

Stress Due To Nozzle Loads:

p = 4.05 kip M = 534.4 in-kip F

z

= 2 .28 kip M = 534.4 + 4.05 (10.22) + 2.28 {0.56) a 577.07 in-kip H 577 *07 aBEID. = Z = 51.98 = 11.102 ~si A

F

=_!=~- 0.12 ksi A 19.03 NEBG-807 A (6/80)

8-20 22A7454 19 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP _ _ _ _ __ _ _ _ , __ _ ____,J REV 1

• Sec-tion B Service Level 'C' Pressure Stress:

= p Di 1.375 (10.875) ae 2t = 2 (0.531) = 14,080.2 psi ae a* = 2

= 7,040 psi a = -1,375 psi r

Stress Due To Nozzle Loads:

p ... 6.44 kip Fz = 4 .61 kip M = 789.3 + 6.44 (10.22) + 4.61 (0.56) = 857.7 ill-kip

! = 857.7 = 16 .501 ksi z 51 .98 F

= -A = 4161 = 0.243 ksi A 19.03 NEBG-407A(6/80)

8-25 22A74S4* 24 GENERAL. ELECTRIC SH NO.

NUCLEAR ENERGY BUSINESS GROUP IIIEV 1 Stress Due to Thenial Sleeve Loads:

P .. 5.08 kips H = 5.39 in-kips F -= 6 .O kip M = S.39 + 5.08 (21.12) + 6.0 (2.36)

• 126.84 in-kip

}! =

• 128.84 Z 51.98 F

= _! = ....!.,.Q_ = 0.316 ksi A 19.03 Total Stress

• a, a

a 0

r

=

=

7,040 + 16,160 + 243 + 2,440 + 316 14,080 psi

-1,375 psi

= 26,199 psi

• NEBG-807A (6/80)

8-27 22A7454 26 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP ~EV l

• Stress Due to Thenial Sleeve Loads:

P = 2.57 kip M = 3.124 in-kip Fz = 5.7 kip H = 3.124 + 2.57 (23.87) + 5.7 (2.36) = 77.93 in-kip M = 1Lll = 1.5 :tsi Z 51.98 F

= _,! = -2.,J_ = 0 .30 ksi A 19.03 Total Stress a~ = 6,400 + 10,680 + 120 + 1,500 + 300 = 19,000 psi

• ae a

r

=

=

1.2,800 psi

-1.2so psi

• NEBG.a07A (6/10)

8-29 22A7454 28 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

- BUSINESS GROUP "EV 1 Stress Due to The1"11lal Sleeve Loads:

P = 5.08 kip X

• 5.39 in-kip F = 6.0 kip K = 5.39 + 5.08 (23.87) + 6.0 (2.36)

• 140.81 in-kip

• ! = 140 .81 "' 2.71 ksi Z 51.98 F

all. = Az = 1::~ 3 = 0.316 ksi Total Stress a6 = 7,040 + 15,820 + 243 + 2,710 + 316 = 26,129 psi a

• 14,080 psi 9

a = -1,375 psi r

NEBG-807 A (6/10)

8-31 NUCLEAR ENERGY BUSINESS GROUP GENERAL f/j ELECTRIC 22A7454 IIIEV l SH NO. 30

• Stress Due to Thermal Sleeve Loads:

P = 2.57 Up F -= 5.7 kip z

M = 3.124 + 2.57 (26.97) + 5.7 (1.8)

• 82.7 in-kip aBEND. = ~ = :;:~9 = 2.12 ksi F

= _! = ..2..i.L = 0.36 k.si ail. A 15.89 Total Stress at = 6.225 + 13,840 + 144 + 2.120 + 360 .. 22.689 psi Ge = 12,450 psi a = -1,250 psi r

• NEBG-a07A (6/10)

8-33 22A7454 32 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP ~EV 1

• Stress Due to Thermal Sleeve Loads:

P = 5.08 kips H = 5.39 in-kips F = 6.0 kip z

M = 5.39 + 5.08 (26.97) + 6.0 (1.8) = 153.2 in-kip M = 153,2 = 3.92 ~s1*

aBEND. = Z 39.09 ~

F

_LQ_

= ~ = = 0.378 ksi A 15.89 Total Stress at = 6,847 + 20,460 + 291 + 3,920 + 378 = 31,896 psi ae = 13,694 psi a = -1,31S psi r

• NEBG-807A (6/10)

8-34 NUCLEAR ENERGY GENERAL fj, ELECTRIC 22A7454 SH NO. 33 BUSINESS GROUP REV 1 Thickness Requirement of Section F Treating the safe end as a 'Nozzle', the safe end thickness adjacent to the attaching pipe shall not be thinner than the greater of the pipe thickness or the quantity t p Smp /Smn ).

Where:

t = Pipe n0111.inal thickness p

smp = Pipe allowable ( Sm) smn = Safe End Allowable (sIll )

For our geometry:

tp = 0.5405 in.

• Smp = 18.1 k3i s = 18.6 ksi lllD.

SAFE END = 0.5855 in.

'llllCKNESS Criteria Met NEBG-a07 A (6/1101

• 8-35 l2A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 34 BUSINESS GROUP ,.EV 1

• Section G Deiign Pressure Stress:

222* (8,505)

"'2,518 psi 2 (0.375) ar = -222 psi Stress Due to Thermal Sleeve Loads:

P = 2.57 kip M = 3.124 in-kip

!

• F * = 5. 7 kip M = 3.124 + 2.57 (13.87)

• 64.47 in-kip

"' !Z = 64

• 47 = 2.89 ksi 22.32 F

ail . . -= AZ = 1~:~6 .,. 0.545 ksi

• 222 psi pressure assumed, twice normal operation

• NEBG-&07 A (6/10)

8-36 22A74S4 NUCLEAR ENER.GY GENERAL@ELECTRIC SH NO. 3S BUSINESS GROUP REV 1 Stress Due To Nozzle Loads:

p = 4.0S kip X = S34.4 in-kip F = 2 .2 8 kip z

The exact amollllt the safe end loads influence the thermal sleeve is unkno,rn.

However, from previous analysis it has been determined that a conservative approach is to ratio the safe end loads to correspond with the moments of inertia.

I Thermal Sleeve 137.93

= = 0.381 I Nozzle 362.33 NOTE: Corrosion not included in calculation (more conservative)

M = S34.4 + 4.0S (4.72) + 2.28 (1.8)

= 0.381 ,!

z = (0 .381)

= SS7.62 in-kip S57,62 22.32

= 9.52 ksi

• F

= 0 .3 81 ~

A

= (0 .381) l~:~: = 0.083 ksi Total Stress

= 1,259 + 2,890 + 545 + 9,S20 + 83 = 14,297 psi ae = 2,S18 psi a * -222 psi r

NEBG-807A (6/80)

8-37 22A7454 36 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP AEV l Section .G

~-*~., J.,evel C' Pressure Stress:*

P*D.

_ _ 1 a:: 333* (8,505)

= 3,776 psi 2t 2 (0.375) a = -333 psi r

Stress Due to Thermal Sleeve Loads:

P = 5.08 kip H = 5.39 in-kip

• F2 = 6_.0 kip M = 5.39 + 5.08 (23.87) = 126.65 in-kip

= M = 126165 = S.675 ksi Z 22 .32

• 333 psi pressure assumed (conservative)

• NEBG-807A (6/90)

8-39 22A7454 NUCLEAR ENERGY GENERALfj ELECTRIC SH NO. 38 BUSINESS GROUP IIIEV 1 Section H Design Pressure Stress:

222* (8,505) a .,. = 2,518 psi e 2 (0.37S) a = -222 psi r

Stress Due to Thermal Sleeve Loads:

P = 2 .57 kip M = 3.124 in-kip I

M = 3.124 + 2.57 (20.12) = 54.84 in-kip

= IZ = li....!!

22.32 = 2 46 ~s1*

• ~

F

= = _i...L

_,! = 0 .545 ks i A 10.46

• 222 psi pressure assumed (conservative)

• NEBG.807 A (6/80)

8-41 22A7454 40 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP PIEV 1

• Section B Service Level 'C' Pressure Stress:

333* (8 .505) -= 3,776 psi 2 (0.375) a = -333 psi r

Stress Due to Thermal Sleeve Loads:

p = S.08 kip M = S.39 in-kip F

z = 6.0 kip M = 5.39 + 5.08 (20.12) -= 107.6 in-kip 107 6 a .. H Z

=

• 22.32

= 4.821 ksi BEND.

F

= ~ = 6.0 = 0.574 ksi A 10.46

• 333 psi pressure assumed (conservative)

• NEB~07A (6/10)

8-43 22A7454 42 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP AEV 1

• Section I Design Pressure Stress:

222* (8,386) = 1,884 psi a =

e 2 ( 0 .494) a = -222 psi r

Stress Due to Thermal Sleeve Loads:

P = 2.57 kip M = 3.124 in-kip

• • M =

F z

3.124 + 2.57 (20.12) =

= 5.7 kip 54.84 in-kip

= zM = .li.,!!

29 07

= 1.887 ksi F

= --!. = 5,7

=

A 13

• 78 0.414 ksi

• 222 psi pressure assumed (conservative)

• NEBG-807A (6/10)

8-44 NUCLEAR ENERGY GENERAL. ELECTRIC l2A74S4 SH NO. 43

---

~-------'

BUSINESS GROUP REV 1

• Stress Due To Nozzle Loads:

p = 4.05 kip H = 534.4 in-kip F = 2.28 kip z

The exact amount the safe end loads influence the thermal sleeve is llllknown.

However, from previous analysis it has been determined that a conservative approach is to ratio the safe end loads to correspond with the moments of inertia.

I Thermal Sleeve 137.93

= = 0 .3 81 I 362 .33 Nozzle NOTE:

• Corrosion not included in calculation (more conservative)

M = 534.4 + 4.05 (8.47) + *2.28 (1.8) = 572.81 in-kip 572,81

= o.381 Mz = co.381) 29.07

= 7.51 ksi F

= 0.381 AZ = (0.381) _Lll = 0 .063 ksi

13. 78 Total Stress

= 942 + 1,887 + 414 + 7,510 + 63 = 10,816 psi ae = 1,884 psi a = -222 psi r

NEBG-807 A (6/80)

8-45 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 4'4 BUSINESS GROUP ,.EV 1

• Section I Service Level 'C' Pressure Stress:

P*D_i

_ = 333* ( 8 .386) a .. .. 2,826 psi e 2t 2 ( 0 .494) a = -333 psi r

Stress Due to Thermal Sleeve Loads:

P -= S.08 kip M = 5.39 in-kip

/

Fz = 6.0 kip M = 5 .39 + 5 .08 (20 .12)

• 107 .6 in-kip

= }I Z

= 107 *6 = 3.702 29 .07 ksi F

~ 6.0

= A = 13. 78

= 0.436 ksi

• 333 psi pressure assumed (conservative)

• NEBG-a07A (6/10)

8-46 22A74S4 45 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP REV 1 Stress Due To Nozzle Loads:

p .. 6.44 kip H a 789.3 in-kip F .. 4 .61 kip z

The exact UIOUD.t the safe end loads influence the thermal sleeve is unkno,rn.

However, from previous analysis it has been determined* that a conservative approach is to ratio the safe end loads to correspond with the moments of inertia.

I Therm.al Sleeve 137.93

= = 0 .3 81 I 362.33 Nozzle NOTE: Corrosion not included in calculation (more conservative)

H = 789.3 t 6.44 (8.47) + 4.61 (1.8) aBEND. = 0 .381

}!

z = (0.381}

=

852,15 852 .15 in-kip 29.07 = 11 .17 ksi F

= 0 .381 ~

= (0.381} ...Lil. = 0.128 ksi a AX. A 13 .78 Total Stress

= 1,413 + 3,702 + 436 + 11,170 + 128 = 16,849 psi a = -333 psi r

NEBG-407 A (6/80)

8-47 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. -46 BUSINESS GROUP PIEV 1

• Section J Design Pressure Stress:

P*D.

__ 1 .. 222* (8.386) a =

• 1,884 psi e 2t 2 (0.494) a = -222 psi r

Stress Due to Thermal Sleeve Loads:

P = 2.57 kip M = 3.124 in-kip F = 5.7 kip z

M = 3.124 + 2.57 (19.33) = 52.81 in-kip 52 81

= }! = * = 1.82 ksi Z 29 .07 F

...! _LL,

= A = = 0 .414 ksi

13. 78

• 222 psi pressure assumed (conservative) *

• NEBG.a07A (6/10)

8-49 NUCLEAR ENERGY BUSINESS GROUP GENERAL fl ELECTRIC 22A7454 JIIEV 1 SH NO. 48

• Section J Service Level 'C' Pressure Stress:

=

P*D

_ _i ... 333* (8 .386)

= 2,826 psi G8 2t 2 (0.494) ae a = "" 1,413 psi 6 2 a = -333 psi r

Stress Due to Thermal Sleeve Loads:

P = 5 .08 kip M = 5.39 in-kip J( = 5.39 + 5.08 (19.33) ~ 103.59 in-kip

- zM =

103,59 = 3 .564 ksi 29.07 F

... ....! = ...L.Q_

= 0.436 ksi A 13. 78

• 333 psi pressure assumed (conservative) *

• NEBG-a07A (6/80)

8-50 22A7454 49 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP REV 1 Stress Due To Nozzle Loads:

p = 6.44 kip H = 789.3 in-kip Fz = 4.61 kip

• The exact amount the safe end loads influence the thermal sleeve is 1U1.kno,ru.

However., from previous analysis it has been. determined that a conservative approach is to ratio the safe end loads to correspond with the moments of inertia.

I Thermal Sleeve 137.93 = 0 .3 81

=

I 362.33 Nozzle

~ Corrosion not included in calculation (more conservative)

M = 789.3 + 6.44 (9.26) + 4.61 (1.8) = 8S7.~4 in-kip

= 0.381 !z = (0.381) 857.24 = 11.24 ksi 29.07 F

aAX. = 0 .381 ...!

A

= (0 .3 81) 1::~i = 0.128 ksi Total Stress

= 1,413 + 3,564 + 436 + 11,240 + 128 = 16,781 psi

= 2,826 psi a = -333 psi r

NEBG-a07A (6/80)

8-51 NUCLEAR ENERGY GENERAL fl ELECTRIC 22A7454 SH NO. 50 BUSINESS GROUP "EV 1 TABLE 3-2 XAXIHUM PRIMARY S'IRESS INTENSITY Pm - Primary Membrane -

All Stresses in ksi p - Primary Bending B

p p

• pm+ Pb **

ll pm! Pb Condition Section --1! Allo-., Allo-.. Jlaterial A 14.0S 26.7 20.48 40.0S SA-508 (CL.2)

B 14 .05 18.1 20.41 27 .15 SA-508 (CL.1)

C 14.05 18.l 20.33 27.15 SA-508 (CL.l)

D 14.05 18.6 20.33 27 .90 SA-350 (LF2)

Design E 14,05 18.6 20.25 27.90 SA-350 (LF2)

Event F 13.70 18.6 23 .94 27 .90 SA-350 (LF2)

G 2.74 18.6 14.52 27.90 sA-3so*-ru-2> --

H 2.74 18.6 14.35 ~

27 .90 SA-350 (LF2)

I 2.11 16.0 11.04 24.0 SA-351 ( CF3)

J 2.11 16.0 11.01 24.0 SA-351 (CF3)

A 15.46 42 ,60 27.73 63 .90 SA-508 (CL.2)

B 15 .46 27.10 27.65 40.65 SA-SOS (CL,l) c* 15 .46 27.10 27.58 40.65 SA-508 (CL.l)

Service D 15.46 27,85 27.58 41.77 SA-350 (LF2)

Level E 15 .46 27. 85 27.Sl 41.77 SA-350 (LF2)

• C' F 15 .07 27.85 33.28 41.77 SA-350 (LF2)

Event G 4.11 27. 85 22.77 41.77 SA-350 (LF2)

B 4.11 27.85 22.33 41.77 SA-350 (LF2)

I 3 .16 19.2 17 .18 28~80 SA-351 (CF3)

] 3.16 19.2 17.11 28.80 SA-351 (CF3)

• P is Sm for Design and the larger of 1.2 Sm or Sy for Service Level C.

mAll o,rab le

• • Pm+ PB is 1.5 S for Design and the larger of 1.8 S or 1.5 S for m m y Allowable Service Level C,

4. ANALYSIS This section provides all the detailed thermal an.d stress analysis required to show an acceptable design for the operating tr&nsients imposed on the nozzle and safe end assembly.

NEBG-407 A (6/10)

8-52 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 51 BUSINESS GROUP 1 REV 4.1 Thermal Transient Analysis, The only feedwater nozzle thermal transient for the vessel operating conditions (Service Levels A and B) is defined in the design specification (Reference 6.1). This transient is also illustrated in Figure 4.1-1 for convenience. In order to simplify the thenial analysis, the feedwater transient was idealized as two separate transients (a heatup and cooldown). These idealized transients are illustrated in Figure 4.1-2.

Notice the step change in temperatures were simulated by steep ramps. This was done to facilitate numerical convergence and results in slightly nonconservative stresses. For more detailed information on the two idealized transients, see Reference 6.1.

4.i.1 Thermal Model. Th.e uisymmetric finite element model of the feed-water nozzle is sho-wn in Figure 4~1.1-1. The model is made up of 2-D axisymmetric isoparametric temperature elements (STIF SS, Reference 6.3). A portion of the vessel wall was modeled as a disc for convenience of analysis since the effect of this approximation on the temperature solutions in the regions of interest is insignificant. The model ends (RPV, thermal sleeve, and safe end) are considered constant for all temperatures. The ther.nal properties used are as follows. Thermal properties are those of approximately 360°F.

Carbon Steel K = 0.03972 BTU/min in°F 3

p =

=

0. 283 lb/ in 0 .1226 BTU/ lbm°F Stainless Steel K ... 0 .01327 BTU/min in°F 3

p =

=

0 .290 lb/ in 0.1191 BTU/lbm°F NEBG-807A (6/80)

8-54 NUCLEAR ENERGY GENERAL fl, ELECTRIC 22A74S4 SH NO. 53 BUSINESS GROUP REV 1.

I

\

8-55 22A7454 NUCLEAR ENERGY. GENERAL. ELECTRIC SH NO. 54 BUSINESS GROUP 4.1.2 Feedwater Nozzle Beat Transfer Coefficients. The heat transfer coe-fficients were evaluated as specified in Appendix 20 of the design specification (Reference 6 .1). The nozzle metal surfaces having unique film heat transfer coefficients are i~ ..1tified in Figure 4.1.1-1. The calculated values for each of these surf.cos follow. Table 4.1.2-1 contains the Tater properties used at the various temperatures analysed.

4.1.2.1 Cool-Down Transient Region 1 ID .. 9.67 in. "" 0.80575 ft.

No Flow Condition (Natural Convection)

For natural convection the film heat transfer equation is as folloTs (Reference 6.4):

hf = 0.14 f (GR Pr) 113

.. 0.14 X

.e.:-11

µ (Pr}

\ 1/3 ,.-1/3

~r

( 2 using water properties at 350°F, and a_ssuming a film temperature diff"erential (AT) of 10°F, obtain

• • NEBG-107A (6/10)

8-56 NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A7454 REV 1 SH NO. 55 TABLE 4.1.2-1 WATER PROPERTIES

~

Yater P:tvoertv ~ w. 250 350 soo 550 1bm 58.8 55.6 49.0 45.9 p 62.0 60.1 3

Ft mu 0.998 1.00 1.01 1.05 1.19 1.31 Ci, lbm °F p.

1bm 0.46 X 10

-3 0.205 X 10

-3 0.158 X 10

-3 0.105 X 10-3 0.71 X 10-4 0.64 X 10-4 Ft Sec K

mu 0.364 0.394 0.396 0.391 0.349 0 *.325 Hr Ft °F p 4.52 1.88 1.45 1.02 0.87 0.93 r

p ll L

IF

...1l!!..

Ft Hr 2 X 10-4 1.649 4 X 0.738 10-4 4.8 X 0.569 10-4 6.9 X 0.378 10

-4 1 X 10-3 0.256 1.1 X 0.230 (Val'aes tu= fJ:tlll Befermce 6.4) 10-3 Where:

p = Density Ci, = Specific Beat ll = Viscosity

r: = Conductivity R = Reynolds No. = (DVp/p.)

e V = Fluid Velocity P

r

= Prandtl No.

3 1 gal = 0.1337 Ft NEBG-a07A (6/80)

8-57 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC _SH NO. 56 BUSINESS GROUP "EV 1 Forced Convection For turbulent flow, the film heat transfer coefficient equation is as follows: (Reference 6.4):

hf O* 023 X R 0.8 p 0.4

= D e r And for 25'1, flow (10<11 flow = 3,720 Gal/min)

(Reference 6.1)

V = .Q = 0.25(3,720)(0.1337)(4)(60) = 14 , 631 Ft/Rr A n (0.80575) 2 At T = 550°F R O.S = 1.25 x 10 5 e

P*r = 0.93 Pro. 4 .= 0.9708

. .. hf = 1125.13 BTU/Rr Ft 2

°F At T = 500°F R =

DVo

= 2.26 x 10 6 R 0.8 = 1.211 x 10 5 e µ e p = 0. 87 Pr0.4 = 0.9465 r

... hf = 1142 .23 BTU/Hr Ft2 op At T = 350°F R = DVo

= 1.734 x 10 6 R 0.8 e

... 9 .so x 10 4

e µ p = 1.015 Pr0.4 = 1.006 r

2 hf = 1100.4 BTU/Hr Ft °F

• NEBG-807A (6/10)

8-58 NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A74S4 REV 1 SH NO. 51 At T a 200°F R =

DVp

= 9.6 x 10 5 R. 0.8 = 6.107 X 10 4 e µ 0 Pr0.4 p

r = 1.88  ; = 1.* 2 853 2

hf = 882.84 BTU/Hr Ft op At T = 100°F DVp 5 R. 0.8 4 R = = 4 .433 x 10 = 3.291 X 10 e .µ 0 p = 4.S2 Pr0.4 = 1. 828 r

2 hf = 62S.27 BTU/Hr Ft °F N£BG-807A (6/80)

8-59 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 58 BUSINESS GROUP PIEV 1 Region 2 ID = 8.38 in. = 0.69833 ft.

No Flow Condition (Natural Convection)

For natural convection, the film heat transfer equation is identical to that of Region 1. Again using water properties at 350°F and a AT of 10°F, obtain hf = 218.44 BTU/Br Ft 2 op Forced Convection For turbulent flow, the film heat transfer equation is as follows:

_(Reference 6.4):

0.8 p 0.4

= 0.023 ID R e r And for 25' flow (101)11, flow s 3,720 Gal/min)

(Reference 6 ~1)

V s 9 = 0.25(3.720)(0.1337)(4)(60)

• 19,478.3 Ft/Br A 2

,r (0.69833)

At T = 550°F 5

R = 2.7098 x 10 6 R 0 *8 = 1.4007 x 10 e e P 0 *4 = 0.9708 r

2 hf s 145S.6 BTU/Hr Ft OF At T = 500°F R

• 2.6076 x 10 6 ae 0

• 8 = 1.3S8 x 10 5 e

0 4 P * = 0.94648 r

2 hf = 1477.75 BTU/Hr Ft °F

• NEBGoa07A (6/110)

8-60 22A7454 59 NUCLEAR ENERGY GENERAL . ELECTRIC SH NO.

BUSINESS GROUP REV 1 At T

• 3S0°F 0 8 5 R

e

• = 1.0989 x 10 P 0. 4 a 1.006 r

2 hf

• 1423.6 BTU/Hr Ft OF At T = 20o°F R = 1.1077 x 10 6 R o.s = 6.8477 x 104 0 e 4 1.2853 P O* =

r 2

hf = 1142.12 BTU/Hr Ft op At T = 100°F P

r 0 4

• = 1.828 R o.s e

= 3.6903 x 10 4

• 2 hf

• 808.92 BTU/Br Ft OF NEBG-a07A (6/10)

8-61 22A7454 60 NUCLEAR ENERGY GENERAL@ELECTRIC SH NO.

BUSINESS GROUP ,.EV 1

• Region 3 ID No Flow Condition a 6 .16 in. a 0 . 51333 ft.

(Natural Convection)

For natural convec** tion, the film heat transfer equation is identical to that of Region 1. Again using water properties at 350°F and a ~T of 10°F, obtain 2 op h

• 218.44 BTU/Hr Ft f

Forced Convection For turbulent flow, the film heat transfer equation is as follows:

(Reference 6.4):

And for 25'11 flow (10~ flow

• 3,720 Gal/min)

(Reference 6.1)

V .,. 9 .. 0,25(3,720)(0.1337)(4) (60)

• 36,047.6 Ft/Hr 2

A fl (0.51333)

At T = 550°F R 0. 8 s 1.792 x 10 5 0

P 0 *4 = 0.9708 r

... hf

• 2533.05 BTU/Br Ft 2 op At T = soo 0 P R O.S = 1.7375 x 10 5 e

0 4 0.9465 P * =

r 2

hf

• 2571.5 BTU/Hr Pt op

• NEBG-107A (6/10)

8-62 NUCLEAR ENERGY GENERAL fj ELECTRIC 22A7454 SH NO. 61 BUSINESS GROUP REV l At T = 350°F 6 R o.s = 1.4057 x 10 5 R8 c 2.7218 x 10 e p 0.4 = 1.006 r

... hf = 2477.4 BTU/Hr Ft 2

OF At T = 200°F R = 1.S069 x 10 6

R 0 *8 = 8.7594 x 10 4

e e p O .4 = 1.285 r

... hf = 1987.5 BTU/Hr Ft 2 OF At T = 100°F P O*

r 4 = 1.828 R

e 0 8

- = 4.7206 x 10 4

2 hf = 1407.67 BTU/Hr Ft OF NEBc:..a07A (6/90)

8-64 NUCLEAR ENERGY GENERAL@, ELECTRIC 22A7454 SH NO. 63 BUSINESS GROUP REV 1 u d

~

'2

-0 e.:.

~

uJ

\Q s....

c.:,

i;.J i:i::

i::r..

.I 0 tfl s....

tCrJ tfl CrJ i::r..

I'll

c E-<

-. I N

-.:r I'll e:

....i::r..

(.!)

NIE G OTA

8-66 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 65 BUSINESS GROUP REV 1 Upper Surface For all feedYater floYs, the film heat transfer coefficient for the upper surface is given in Appendix 20 of the design specification (Reference 6.1).

2 The coefficients given are as follows (hf in BTU/Ft Hr °F);

Section hf at Left End hf at Right End Variation A- B* 400 750 Linear B- c* 750 750 Constant C- D 750 1500 Linear D - E 1500 1500 Constant E - F 1500 500 Linear Lower Surface Since the thermal model does not include the secondary thermal sleeve~

the equivalent film heat transfer coefficient for the outer surface of the

• primary sleeve must be fo1llld. The equivalent heat transfer analysis for the primary sleeve is made up of the following:

a. hf of outer surface secondary sleeve (h1 )

b. hf of inner surface secondary sleeve (h )

2

c. hf of outer surface primary sleeve (h )

3 The conduction through the secondary sleeve will be neglected along Yith any conduction through the Yater. The three modes of heat transfer are illustrated in Figure 4.1.2-2~

• NOTE: This is different from the design specification. However, the slight difference yields higher heat transfer coefficients for these sections and thus is conservative.

NEBG-e07 A (6/80)

• 8-67 22A7454 66 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

~USINESS GROUP "EV 1

• Natural convection will be assumed for the annulus between the two sleeves. Therefore, from Reference 6.4, for natural convection, using properties at T s 3S0°F, obtain BTU Hr Ft 2 op

- - - -,fssu~a.ng tlie !:,,.T from ihe water to the thermal sleeve surface is the saJDe

- i~r both primary and secondary sleeves, obtain -- - ---- _.

q -= h eqT A (!:,.TT)

A = surface area

_1_ =

h eqT

!:,.TT

• TAnnulus - TFeedwater And q = h , A (AT )

eql 1 where:

A = surface area And where:

• NEBG-807 A (6/110)

A = surface area

8-68 NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A7454 REV l SH NO. 67 Using the fact that the heat flow is constant through the heat transfer patp, obtain q = constant Therefore,

=

Recalling that for natural convection,

= = 101.4 ,,,:r1 13 obtain 101.4 Ar 13 (AT)

( t; + 10:./ 3j Ar/

This reduces to yield the following, This was solved by trial and error. The following is a summary of the solutions.

BTU BTU Feedwater h3 = 750 h3 = 1500 Temperature Ft 2

Hr °F Hr Ft*2 OF 5OO°F AT = 18.74°F AT = 19 .2S°F 35O°F AT = 73.35°F AT = 82.PF 2OO°F AT = 124.2°F AT = 141.25°F l0O°F AT = 1S6.9°F AT = 179 .8°F Recalling the equation for the equivalent heat transfer coefficient, obtain

_1_

h eq NEBG-407 A (6/80)

8-70 NUCLEAR ENERGY GENERAL fj ELECTRIC 22A7454 SH NO. 69 BUSINESS GROUP REV 1 4.1.2.2 Heat-Up Transient Region 1 Forced Convection For turbulent flow, the film heat transfer equation is as follows:

{Reference 6.4):

For 25 percent flow, from previous section (Paragraph 4.1.2.1),

V = 14,631 Ft/Hr and for 100 percent flow V = 58,524.1 Ft/Hr At T = 100°F {Flow= 25 percent)

The heat transfer coefficient is identical to that calculated for the Cool-Down transient (Paragraph 4.1.2.1)

BTU At T = 180°F (Flow= 25 percent)

Will assume the heat transfer coefficient to be identical to that calculated at 200°F in the Cool-down. section.

BTU NEB0-.07 A (6/IOJ

8-71 22A7454 SH NO. 70 NUCLEAR ENERGY GEN ERA Lfj ELECTRIC "EV 1 BUSINESS GROUP

• At T = 260°F (Flow

• 25 perce nt) 6 R O.S = 7.391 2 X 104 R

• 1.218 7 x 10 e e

P O*4 = 1.160 5 r

At T = 376°F (Flow a 100 perce nt) 5 Re .= 6.936 x 10 6 R 0 *8 = 2.970 9 x 10 .

e p 0.4 = 1.006 r

• !".EBG-a 07A (6/10)

8-72 NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A7454 REV 1 SH NO. 71 Region 2 m = 8.38 in.

• 0.69833 Ft.

Forced Convection

. For turbulent flow, the film heat transfer equation is as folloTs:

(Reference 6.4):

For 25 percent flow, from previous section (Paragraph 4.1.2.1),

V = 19,478.3 Ft/Hr and for 100 percent flow V = 77,913.8 Ft/Hr At T = 100°F (Flow= 25 perc~nt)

The heat transfer coefficient is identical to that calculated for the Cool-Down transient (Paragraph 4.1.2.1)

BTU 808. 92 Br Pt 2 op At T = 180°F (Flow= 25 percent)

Will assume the heat transfer coefficient to be identical to that calculated at 200°F in the Cool-down section. (Paragraph 4.1.2.1)

BTU hf = 1142 .12 Hr Ft 2 op NEBG-807A (6/10)

8-73 22A7454 72 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP ,.EV l

• At T a 260°F (Flow

• 25 percent) 6 0 *8 4 Re s 1.406 x 10 R = 8.2875 x 10 e

0 4 P *

• 1.1605 r

At T = 376°F (Flow ~ 100 percent) 6 Re0

• 8 = 3.33 x 10 5

R = 8.0 x 10 e

P r

o. 4 = 1.006 BTU 4315. 7 2 Hr Ft °F NEBG-407A (6/10)

8-74 22A7454 NUCLEAR ENERGY _ GENERALfj ELECTRIC SH NO. 73 BUSINESS GROUP REV 1 Region 3 m = 6.16 in. = 0.51333 Ft.

Forced Convection For turbulent flow, the film heat transfer equation is as follows:

(Reference 6.4):

For 25 percent flow, from previous section (Paragraph 4.1.2.1),

V = 36,047.7 Ft/Br and for 100 percent flow V = 144,190.7 Ft/Br At T = l00°F (Flow= 25 percent)

The heat transfer coefficient is identical to that calculated for the Cool-Down transient (Paragraph 4.1.2.1)

BTU

= 1407.67 2 Br Ft °F At T = 180°F (Flow= 25 percent)

Will assume the heat transfer coefficient to be identical to that calculated at 200°F in the Cool-do,rn section (Paragraph 4.1.2.1).

BTU hf = 1987 .5 Br Ft 2 °F NEBGoa07A (6/110)

8-75 22A7454 74 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP IIIEV 1

• At-T = 260°F {Flow

• 25 percent) 6 R 0 *8 1.06 x 10 5 R = 1.913  % 10 e s

0 p O.4 = 1.1605 r

At T = 376°F {Flow = 100 perc*nt) 5 Re = 1.0887  % 10 6 ; Re 0

• 8 = 4.261 x *10 p 0.4 = 1.006 r

BTU hf = 7509.98 2.

llr Ft °F

• NEBG.a07A (6/90)

8-77 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 76

• BUSINESS GROUP "EV 1 The following is a summary of the solutions.

BTU BTU Feedwater h3 = 750 h3

• 1500 Temperature Ft 2 Hr OF Hr Ft 2 op l00°F AT - 1S7°F AT =

180°F 180°F AT -

131°F AT 149°F 260°F AT 104°F '1T  :.: 117.7°F 376°F AT "" S8°F AT = 64.S°F Recalling the equation for the equivalent heat transfer coefficient, obtain

_1_ = L + _____2_____

heq h3 101.39 {AT) 113 The following is a summary of these equivalent heat transfer coefficient calculations.

BTU All h's in 2

Rr Ft OF h

Feedwater Temperature h3 . 750 e

~ = 1500 100°F 200.4 240.4 191.7 228.0 180.9 213 .14 lSS.S 179.04 4.1.2.3 Normal Operation, The heat transfer coefficients used for the normal operation run are the same as those given during the heat up transient (Paragraph 4.1.2.2} when T

• 376°F *

• • NEBG-807 A (6/80)

8-80 NUCLEAR ENERGY GENERAL fl ELECTRIC 22A74S4 SH NO. 79 BUSINESS GROUP REV 1

'I i;

i j

C"*.

~

i FIGURE 4.1.4-1 COOLDOWN TRANSI:mn' NEBG-&07A (6,aO) **

8-82 GENERAL * . ELECTRIC 22A7454 SH NO. 81 NUCLEAR ENERGY BUSINESS GROUP REV l I.

MO~TlCELLO FE~TER t-.OZZLE ST!\ESS A~LYSIS - TEHPEAA1URE PL01 (iIHE~so.o SECi I

I J.

~

FIGURE 4.1.4-3 COOLDOWN TRANSIENT NE9a-a07A (6/80)

8-84 NUCLEAR ENERGY ----- GEN ERAL fj ELE CT RI C 22A7454 REV SH NO. 83 BUSINESS GROUP 1 MCJNT !CELLO FEEDrlATER NOZZLE STRESS AN, .L1515 HEAT-UP TEMPERATURE PL(j1 Tc. '3.o SEC..

N C

FIGtlll 4.1.4-5 NEB~07A (6/80)

8-85 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 84 BUSINESS GROUP AEV l

• MONTlCtLLO FEEDWATEA NOZZLE STRESS ANALYSIS HEA1-UP TEMPERATURE PL01 i:"* /2.0 SEC.

II")

If)

FIGURE 4.1.4-6

• NEBC..07A (6/80)

8-86 NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A745'4 REV J SH NO. 85 1I PIGUU 4 .1

• 4-7 .

NEBG-a07 A (6/SO)

8-87 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 86 BUSINESS (;ROUP REV 1

• 4.2 Stress Analysis, The stress analysis is broken into three separate are~s: thermal stresses, mechanical load stresses, and pressure stresses. The stress intensities for these three cases are then summed up to yield the total primary plus secondary stress intensity. A fatigue analysis is then performed to obtain tho fatigue usage factor for the system.

4.2.1 Selected Locations for Stress Evaluation, The sections sho,rn in Figure 4.2.1-1 are the locations selected for stress evaluation. Finite element stresses are integrated across each section to dottrmine the equivalent membrane, bending, and peak stresses. This Yas done by averaging the stress across the section and linearizing the stress distribution through the section thickness as shown in Figure 4.2.1-2. These calculations were performed using an engineering aid computer program *~NO'. A listing of this program is included in Appendix 10 *

• NEBG-107A (6/10)

8-92 NUCLEAR Ef-JERGV 22A7454 SI-I NO. 91 BUSINESS GROUP REV 1 Location Nod;o l p::iir.t~

1 354. 353. 352 2

3 324, 3,~

13G, i3S, 134, H3, 132 321

.. 1s1. 130, 129. 12s. 121, 126 s .. .,..

• "~, :ts1. 3eo. 379. 378 6 40, 39. 3a. 37. 35 7 so. 19, 78, 11. 16 0

FIGUP.E 4 .2 .2 .1-1 LOCATION3 Us:!D FOR EYALUATING TEUPITRATtJ"n..E DIFFr""~iCE NEBG-907A (6/80)

8-95 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 94 BUSINESS GROUP fllEV 1

• 'B' No%%le Loading P *

(Service Level 'B')

(0.82 2 2 112

+ 4.34 )

• 4.42 Sta tic F

• 1.37.

z Dynamic F = 0.26 z

• NEBG.a07A (6/901

8-96 NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A7454 REV 1 SH NO. 95 Thermal Sleeve Loading ( Service Leve 1 'B,)

p = (F 2 + F 2)1/2 X y

)l = (M 2 + H 2 + K 2,112 X y z p = 0 Static II = 0 F = 3.7 z

2 + 0.32)1/2 p = (2.5 = 2.52 Dynamic ){ = (1.2 2 + 2.02)2 1/2 = 2.333 F = 1.5 z

Therefore, the following will be used to calculate the largest mechanical load range. Note: the dynamic loads are due to seismic loadings only.

Nozzle Loads (Service Level 'B')

P =* 4.42 kip Static M = 275.4 in kip P = +/- 3.136 kip Dynamic )l = +/- 390.9 in kip F

z

= +/- 0.26 kip NEBG-807A (6/80)

• 8-97 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 96 BUSINESS GROUP

• Thermal Sleeve Loads (Service Level 'B')

P =  :!: 2.52 kip Dynamic H =  :!: 2.333 in kip F =  :!: 1.5 kip z

Static F = 3.7 kip z

4.2.3.2 Mechanical Load Range Calculations, The section properties used are found in Table 3-1 of Section 3. Note, these properties include effects of corrosion. A sllll1Jl1ary of the calculations is given in Table 4.2.3.2-1

• NEBG-a07A (6/10)

8-98 NUCLEAR ENERGY -*- GENERAL fl ELECTRIC 22A7454 SH NO. 97 BUSINESS GROUP REV l Section A Noz.z. le Loads P = 4.42 kip P = +/- 3.136 kip M = 275.4 in kip X = +/- 390.9 in kip F = 1.37 kip F z

= +/- 0.26 kip z

M = 275.4 + 4.42 (12.83) + 1.37 (0.56) M = 390.9 + 3.136 (12.83) + 0.26 (0.56)

332.9 in kip = 431 .3 in kip 431.3 a,BEND = }I z

332.9 51.98

= 6.41 ksi 51.98

= 8.3 ksi F z.

a* il.

-

A 1.37 19.03

= 0.072 ksi = 0.014 ksi Thermal Sleeve Loads P = +/- 2.52 kip Fz = 3.7 kip H = +/- 2.333 in kip F

z.

= +/- 1.5 kip K = 3.7 (2.36) = 8.73 in kip H = 2.333 + 2.52 (15.76) + 1.5 (2.36)

45.59 in kip 8.73 = M 45.59 = 0 .877 ksi 0.168 ksi

51.98 a~E'ID = z 51.98 F

c,t =

_LL

= 0.195 ksi a,AI. = J.

A

= ..L.L 19.03

= 0.079 ksi il. A 19.03 Total Stress a,BEND = 6.578 psi = +/- 9,177 psi c,t * .. +/- 93 psi 267 psi AX.

NEBG-807A (6/80)

8-99 22A7454 98 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP IIIEV 1 Section B P

• 4.42 kip p. = +/- 3.136 kip M = 275.4 in kip M *  !,390.9 in kip Fz = 1.37 kip F .. +/- 0.26 kip z

M .. 275.4 + 4.42 (10.22) + 1.37 (0.56) M

• 390.9 + 3.136 (10.22) + 0.26 (0.56)

= 321.34 in kip -= 423 .l in kip 321.34 = 423.1 6 .182 ksi = 8.14 ksi 51.98 51.98 F F 1.37 -z= _Lli.

--1 =

• 0.072 ksi

• 0.014 ksi 19.03 19.03 Thermal Sleeve Loads P = +/- 2.52 kip F

z

= 3.7 kip M = +/- 2.333 in kip F

z

= +/- 1.5 kip M

• 3.7 (2.36)

• 8.73 in kip M = 2.333 + 2.52 (18.31) + 1.5 (2.36)

= 52.17 in kip 8 73 51.98

= 0.168 ksi a, BF.ND

= !Iz = 52,17 51.98 = 1.004 ksi

_LL

• 0.195 :t.si a* -

F

---' a ...L..L .. 0.079 ksi 19 .03 il. A 19 .03 Total Streu a* BEND

• 6,350 psi = +/- 9,144 psi a6

• 267 psi * +/- 93 psi il.

NEBG-a07A (6/90)

8-101 22A7454 100 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

BUSINESS GROUP PIEV 1 Section E NO%%le Loads P

• 4.42 kip P = :t 3.136 kip H = 275.4 in kip K = :t 390.9 in kip F = 1.37 kip F z

= :t 0.26 kip z:

M

• 275.4 + 4.42 (4.72) + 1.37 (0.56) K

• 390.9 + 3.136 (4.72) + 0.26 (0.56)

• 297 .03 in kip = 405

• 8S in kip a~BEND = !Z - =

297,03 Sl.98

= S.715 ksi a~BEND

= !l z - 405.85 51.98

= 7.808 ksi F F

% 1.37 = ...!. = ~- = 0.014 ksi at = -=

• 0.072 ksi a6

(_. il. A Thermal Sleeve Loads 19.03 il.

P = :t A 19.03 2 .52 kip F = 3. 7 kip X = :t 2.333 in kip z

F z

= :t 1.5 kip H = 3.7 (2.36) = 8.73 in kip H

• 2.333 + 2.52 (23.87) + 1.S (2.36)

= 66 .03 in kip 8.73 =  !! 66.03

! = 0.168 ksi = z =

51.98

= 1.27 ksi z 51.98 a~&ID F

F

% J..a1_

19 .03

= 0.195 ksi a6 AI.

= - = ....L.L A

z:

19 .03

= 0.079 ksi Total Stress a6

• 5,883 psi * :t 9,078 psi BEND a6 = 267 psi * :t 93 psi il.

NEBG.807A (6/80)

8-105 22A7454 SH NO. l04 GENER AL. ELECTRIC NUCLEAR ENERGY BUSINESS GROUP AEV 1 Section I Nozzle Loads P = 4.42 kip P =  :!:. 3.136 kip H = 275.4 in kip M ""  :!:. 390.9 in kip F = 1.37 kip P -=  !. 0 .26 kip z z H = 275.4 + 4.42 (8.47) + 1.37 (1.8) M

• 390.9 + 3.136 (8.47) + 0.26 (1.8)

= 315.3 in kip = 417.93 in kip F . F a, AX.

= 0.381 ...! = (0.381) ..Lil= 0.038 ksi A 13.78

-= 0.381 ...! = (0.381)

A 0 26 13.78

= 0.008 k.si Thermal Sleeve Loads P *  !, 2.52 kip F = 3. 7 kip H =  !, 2.333 in kip z  ?

F ""  !, 1.5 kip z

H = 2.333 + 2.52 (20.12)

= 53 .04 in kip M = 53.04 1.83 ksi

a.,BE?-.'D z 29.07 F

-=

% ..LL

• 0 .269 ks i a.,AX. =

F

-z= -1..&.L .. 0.109 ksi 13.78 A 13. 78 Total Stress at = 4,133 psi =  !,7,308psi BF.ND a = 307 psi =  :!:. 117 psi 6AX

• NEBG-a07A (6/90)

8-107 22A7454 SH NO. l06 NUCLEAR ENERGY GENERAL . ELECTRIC 1 INESS GROUP FIEV TABLE 4 .2 .3 .2-1 XAXIMUM MECHANICAL LOAD STRESS INTENSI'IY ( Service Leve 1 'B')

Section Static Stress .1y-i: .... .c Stress*

A 6,845  :!:. 9,270 B 6,617  :!:. 9,237 C 6,385  :!:. 9,204 D 6,385  :!:. 9,204 E 6,150  :!:. 9,171 F 7,721  :!:. 12,110 G S ,504  :!:. 9,888 H 5,787  :!:. 9,665 I 4,440  :!:. ~ ,425

• • 1 (All stress in psi) 4,486  :!:. 7383

• Dynamic stresses are due to seismic only *

~EBG.a07A (6/80)

8-119 NUCLEAR ENERGY GENERAL f/j ELECTRIC 22A7454 SH NO. 118

• BUSINESS GROUP TABLE 4.2-S AEV TBERMAL MEMBRANE PLUS BENDING S'IRESSES (CASES) 1 Boat Up (t = 2.0 seconds) All stresses in psi Inside Outside Section at ae a r ab ae a r

A -1,485 -235 -1,393 -1,009 B 1,521 204 1,521 1,791 C -8, 885 -4,262 8,458 2,412 D -9,129 -4,574 8,694 2,314 E 5,083 -8,4 86 -s,oos -7,446 F 5,694 -4,237 -5,899 -sss

• G B

I 1-

-5,412 22,221 22,531 24,286

• 10,737 25,867 14,531 28,860 4,408

-22,841

-21,391

-24,354 3,684

-7 ,180

-30,069

-30,295 Sections Illustrated in Figure 4.2.1-1 *

• NEBG-807 A (6/90)

NUCLEAR ENERGY .GENERAL. ELECTRIC 22A7454 SH NO.

BUSINESS GROUP REV 1 4.3 Fatigue Analysis. This section provides all the detailed fatigue analysis required to show an acceptable design for the operating transients imposed on the nozzle and safe end assembly. {Service Level 'B' Events) 4.3.1 Stress Concentration Factors Section A The geometry of the outside surface of Section A is illustrated in Figure 4.3-1. To calculate the stress concentration factor, Reference 6.6 will be used.

NOTE:

~ = 3.88 = 5.97 t .65 Using Paragraph A.7.2.6 {Reference 6.6),

D = 2T = 12.12 in d = 2t = 1.3 in X:

0 is found *from Figure A. 7-1, since the scale stops at r/t = 3 .6, th.at value

• is assumed.

X0 = 1.24.

Using Paragraph A.7.2.4, for r <h Solving for K', obtain K' = 1 + 0.24 NEBG-807A (6/80)

8-125 22A7454 124 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO.

• BUSINESS GROUP AEV l Section F The geometry of Section F is illustrated in Figure 4.'!

• To calculate the stress concentration factor, Reference 6.6 will be used. A concentration factor for both the inner 1U1d outer surface will be calculated, however, only the largest will be used.

Assumer= 0, hence K = 4.0.

0 Then using Paragraph A.7.2.4 (Reference 6),

<K' - 1) L (I - l} = l - 90 0

Solving for K', using~ inner = 72.3° and~ ou t er = 75.74°

,. I:'.uu1er = 1.59 X'

outer

= 1.48 NEBG-a07A (6/10)

8-135*

NUCLEAR ENERGY BUSINESS GROUP GENERAL fj ELECTRIC 22A7454 REV 1 SH NO* 134 TABLE 4 .3-5 THERMAL 'SKIN' STRESSES (P + Q + Fl) (CASE 3)

Cool Do,ru ( t = 60. seconds) All stresses in psi Inside Outside Section at ae a r at) ere a-r A 428 983 -189 75 B 702 1,269 -419 219 C -4,004 -867 4,317 467 D -4 ,189 -1,119 4,505 360 E 17,328 -6,934 -6, 859 -15,349 F 20,038 10,184 -15,449 -6,3i1

• G B

I

-2,865 26,786 33,583 28,192 29,469 26,665 1,709

-35,539

-23, 875 9,610

-17,295

-34,231 1 35,206 40,230 -29,417 -37,795 Sections Illustrated in Figure 4.2.1-1 *

• N£BG-807A (6/80)

8-137 fl, ELECTRIC NUCLEAR ENERGY GENERAL 22A7454 SH NO. 136 BUSINESS GROUP REV l TABLE 4.3-7 'IBERMAL 'SKIN' S'IRESSES (P* + Q + F ) ( CASE 5) 1 Heat Up (t = 2 .o seconds) All str -!I in psi Inside Outside Section ert ere er er ti ere er r r A -1,875 -635 _1,352 840 B -2,200 -356 1,341 1,560 C -10,008 -5 ,183 8,597 2,139 D -10,271 -5,520 8,851 2,041 E 8, 83S -9,1S9 -4,209 -8,1S8 F 2,762 -7.988 -8 ,384 -3,018

• G B

I

-8,872 14,115 20,488 7,052 20,018 11,739 3,722

-34,S78

-23 ,383 816

-15,131

-32,975 1 17,903 22,761 _ -30,218 -3S ,591 Sections Illustrated in Figure 4.2.1-1.

NEBG-a07A (6/80)

8-146 NUCLEAR ENERGY BUSINESS GROUP

- GENERAL fj ELECTRIC* 22A7454 SH NO . .* 145 REV 1 APPENDIX 10 LISTING OF 'NONO' 10 REAL NEMSTF'

~~ REAL HEUS'~

30 REAL NUAVGS 40 REAL NUDlS 50 REAL MOMARK 60 REAL MOMENT 70 DIMENSION NELK(25>,CORDX(:7>.STRSl27>,DISAVG(27>,5T~VG(17l,TST~i:~

80& HEUSTR(27>,NUAVGSl27>,NUDIS(27l,MOMARKl27l,NOMENTC27>,~0U~TC99~;

90 PRINT:""

1 00 PRINT: 1111 110 f*RINT:""

120 PRINT:"DO YOU NEED INSTRUCTIONS' 1 30 REAI1: NOYE SO 140 IF(NCYESO.EQ.0) GOTO 10 150 PRI~T:""

160 f'RINT:""

170 PRINT:""

180 PRINT:""

190 PRINT:"THIS PROGRAM LINEARIZES THE STRESSES THROUGH THE" 200 PRINT: 1111 210 PRINT:"THICl(NESS OF A SECTION."

220 f*RINT: 1111 230 PRINT:"INPUT THE LOCATION OF THE POINTS UHERE STRESSES ARE" 240 PRINT:""

250 PRINT:"ACTING, I.E. THE FIRST SURFACE, THE CENTROIDS AND" 260 PRINT: 1111 270 PRINT:"THE SECOND SURFACE,"

280 PRINT:""

290 PRINT:"INPUT STRESSES AT THE FIRST SURFAFE, CENTROIDS AND "

300 PRINT:""

3 l(i PRINT: "THE SECOND SURFACE."

PRINT:""

330 PRINT:" IF N IS THE NUMBER OF ELEi'IENTS THEN TOTAL STRESS AN[1

340 PRINT:""

350 PRINT:"LOCATlON ENTERlES REOUIREr1 ARE H+2, lI.E. 2 SURFACE~"

.360 F'R INT: 111 370 PRINT:"AND H CENTROIDS.>"

38*} PRINT:""

390 PRINT:*THE PROGRAM THEN COMPUTES THE MEMBRANE STRESS BY"

-400 PRINT:""

410 PRINT:"THE EQUIVALENT AREA METHOD."

420 PRINT:""

4 30 PRIHT:"THE BENDING STRESSES ARE CO"PUTED BY LINEARIZING THE" 4 40 PRINT:""

450 PRINT:"KO~ENT ACROSS THE SECTION THICKNESS."

460 PRINT:""

47(* PRINT:"PEAK STRESSES ARE THE TOTAL STRESS MINUS THE" 480 f*RHIT:""

491) PF:l1t7:"I\El'\BRANE AN!: BENDJNG _sn:~sSES AT THE SURFACES."

500 PRINT:.,.

501 PRINT:""

8-147 NUCLEAR ENERGY GENERAL

• ELE CTR I c* 22A7454 BUSINESS GROUP AEV 1

.APPF.NDil. 10 (Continued) so:: PRINT:""

%3 n:,;;:"

510 PRINT:"UHE~ FIN1SHE1 iNFU1 HUK~tR OF E~En~~:s = 0 10 [AiT '

5:V PRINT:""

'5 30 P~:INT:""

540 10 PR I NT : '"'

55 *) F:R!NT:""

560 PRINT:"UANT A LlS7lNG OF THE INPUT STRS & COORDS., (1=YES.O=~Cl" 570 READ: NOYES 580 DO 300 L=l. 999 590 PRINT: II II 600 F'F:INT: II JI 61C NSTRS=O 6~0 f*RHH:" lHF*UT NO. OF ELEM ACROSS THC!: "

630 READ: NEL 640 IF<NEL.LE.O)GOTO 301 6S0 LOCATE=NEL+2 660 KOUNTH >=L 670 IFCKOUNT<L>.EQ.1J GOTO 11 680 f'RINT: ""

690 f'RlNT: 1111 700 PRINT:"USE COCRDS. FROM THE PREVIOUS RUN 7 ? (YES=l ,NO=O>"

-'10 READ: NOYES1

-20 IF(NOYES!.EO.O> GOTO 11 730 GOTO 12

~'40 11 PRINT: "INPUT COORD. LOCATIONS x1,x2, *** ETC~ ~

. 750 READ: (CORDX ,1=1,LOCATE>

760C 770 12 PRINT:" INPUT CORRES STRESSES 11 780 NSTRS=LOCATE 790 READ: (STRS!J).J=1,NSTRS>

800 TDTSTR=O.

G10 TOTIIIS=O.

820 DO 100 K=l, <NSTRS-1 >

830 STAVG(KJ=<STRS(Kl+STRSCK+1))/2.0 840 DISAVG<K>=(CORDX<K+1)-CORDX<K>>

eso TSTR<K>= STAVG(K>*DISAVG(K) 860 TOTSTR=TOTSTR+TSTR<KJ 870 TDTDIS=TOTDIS+DISAUGlK>

880 100 CONTINUE 69*j ~ENSTR=TOTSTR/TOTDIS 90C- T0TNOfl=.0.

• NEBG-a07A (6/aO)

8-148 NUCLEAR ENERGY .--~-G ENER AL fj ELE CTR IC 22A7454 SH NO. 147 BUSINESS GROUP REV 1 APPENDIX 10 (Continued) 910 DO 101 l=l.NSTRS 910 101 NE~STRlI>=STRSIIl-"EMST~

93C* DO 102 1=1, IHSTRS*-1 i 940 NUAVGS=<NEUSTRlll+NEUSTR<I+l 7)/2.0 NUDISCI>= CORDXII+ll-CORDXII) 960 MO~AR"<l>=<CORDX<Il-CORDXlll)+(CORDX(l+ll-CORDXilll/2.

9;"(: HOHENTCI>=MUAVGSCI>*NUDIS~MOHARMII>

960 TOTHOH= TOTMOH+HOMENT(ll 990 102 CONTINUE

, coo SBEND=l6~TOTHOHI/CTOTDISiTOTDIS>

101 0 SBENDl=ABSCSBEND>

1020 SBENI:2=SBEN!11 ti -1

• l 1 021 SBENt1=ABS ( SBEtrn l 1 03*~ IFCSTRSll l.LT.STRS<NSTRSll GO TO 40C

• 040 PEA~S1=STR5ttl-HEMSTR-SBEND1

~ 050 PEAKS2=STRSCNS1~Sl-HEHSTR-SBEND2 1 06*; GO TO 401 _

1 OiO 400 PEAKS1=STRSC1 >-MEMSTR-SBEND2 1 060 PEAKS2=STRSCMSTRS>-MEMSTR-SBEHDI 1 090 401 IFCNOYES.EG.Ol GOTO 20 1100 PRINT:" II

~ i 10 PRINT:"lNPUT STRESSES ARE:"

1120 PRINT:""

. 1130 URITEC06,202)CSTRS(ll,I=1,NSTRS>

1140 PRINT:""

11 so PRINT:""

1160 PRINT:"INPUT COORD. ARE:"

1170 PRINT:""

1 t 80 URITEt06,203llCORDX,I=1,LOCATE>

1190 20 URIT£(06,200l HEMSTR,SBEND.PEAKS1,PEAKS2 1200 300 CONTINUE

• 1210 301 URITE<Oo,204) 1220 200 FORMAT(//,4X,"MEMBRANE STRESS= ",F10.1,4X, 1 23C & *BENDING STRESSES =C+ OR-l ",F10.1.//,4X, 1 240l ,"PEAKS1 =".Ft0.l,4X,"PEAKS2 =",F10.1.//l

~ 250 202 FOR~ATl4X,4F10.1,/)

1 26*:> 203 FORMAT<4X,4F10.3,/)

1 :70 204 FORMAT(//," HAVE A NICE DAY !! ". / / J

' 26C* ST Of*

, no ENr:

NEBG-e07A (6/90)

8-149 NUCLEAR ENERGY BUSINESS GROUP GENERAL fl, ELECTRIC 22A74S4 FIEV 1 SH NO. 148 APPENDIX 20 INTERGRANUI.AR S'nESS CORROSION INDEX CALCULATION The areas requiring an IGSCC stress rule index calculation are Sections C/D and B/I. These sections are at new weld locations.

20.l Sections C and D S.I. = Q + F + RESID S + 0.002E y

Section C is SA-S08 ( Class 1) carbon steel and Sect ion D is SA-3S0 (LF2) carbon steel.

20.1.1 Mechanical Load Stress. The ~echanical load stresses are the result of dead weight, the.nnal, and hydraulic loads during normal operation. The loads are obtained from Reference 6.1, and are as follows.

Nozzle Loads

• Dead Weight F

F X

y (Nozzle 'A' Loads used).

• -0.11 kip

= -0.63 kip ID Ill X

y

= +11.6 in-kip

= -14 .1 in-kip Fz = +0.15 kip Ill

r.

= -11.1 in-kip Thermal (Noule 'B' Loads used)

F X

= +-0.82 kip m

= +267.2 in-kip F = -4.34 kip Ill = +66 .7 in-kip y y F

z

"" +1.37 kip Ill

= +1.4 in-kip

• NEBGo807A (6/10)

A-1 c; 1 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 150

• BUSINESS GROUP "EV 1 20.1.1 (Continued)

Thermal Sleeve Loads Dead Weight plus Hydraulic P = 0.3 kip m = l .2 in-kip F = 3 .O kip z

Thermal F = 1.2 kip z

Section C and D Nozzle Load Stresses Dead Weight m = 21.4 + 0.64(7.47) + 0.15 (0.56) = 26.27 in-kip aBE-ID

.m =

26,27

= 0.506 ksi z 51.98 F

- ...! Q.J,.L ail. - A = 19.03 = 0,008 ksi Thermal m = 275.41 + 4,42(7.47) + 1.37(0.56) = 309.2 in-kip .

309 2 a

BFm>

~z

• 51.98

= S.95 ksi F

=-

= L1L = 0.72 ksi A 19.03

• NEBCr807A (6/110)

8-159 2 7454 NUCLEAR ENERGY G£ N £RA L . ELECTRIC 2A SH NO. 158

• _e_u_s1_N_E_s_s_G_R_o_u_P_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.__"_e_v_1_________....J 20.2.1.2 Pressure Stresses Primary Pressure Stress PD.

= __1 = 111(8.505) ae 2t = 1.26 ksi 2(0.375) ae

= 0.63 ksi 2

a = -0

• 111 ks i r

Secondarv Pressure Stress. The stresses in Table 4.2-9 are the P+Q pressure stresses corrosion not included. The primary stress (corrosion ~ot i~cluded) is expected to be:

a = 0 .93 ksi e

O'~ = 0.465 ksi The actual stress from Table 4.2-9 is given as:

a0 = -1.38 ksi a~= 3.19 ksi Therefore, the secondary pressure stresses are as follows:

a = -2 .31 ksi 0

a~= 2.73 ksi 20.2.1.3 Thermal Stresses. The thermal stresses given in Table 4.2-8 for steady state normal operation are:

a

• 25.12 ksi 0

a~

• 12.08 ksi

• N£BG-807A (6/80)

8-165 22A74S4 164 NUCLEAR ENERGY GENER AL. ELECTRIC SH NO.

• BUSINESS GROUP PlfV l 20.3 Conclusion. All stress indices are less than allowable of 1.0. The calculated indices are as follows.

S. I.

Location (1) (2)

C 1.002 0.965 0.7929 I 0.8701 (1) Based on Stress Intensities (2) Based on Positive Principle Stresses

• NEBG-807A (6/10)

8-1 f. '

22A7454 166 NUCLEAR ENERGY GEN ERA L . , ELE CTR I C SH NO.

• BUSINESS OPERATIONS REV 1 TABLE 30.3.1-1 PRIMARY STRESS ANALYSIS (All stresses in ksi)

After Deviations Before Deviations All owabl es Case Location Pm Pm+B Pm Pm+B Pm Pm+B Design F 13 .96 24.3 8 13.70 23 .94 18.6 27 .90 G 2 .83 14.90 2.74 14.54 18.6 27.90 Service F 15 .36 33. 89 15. 07 33.28 27. 85 41.77 Level C G 4.25 23.37 4.11 22.77 27 .85 41.77

• NEO 107 A (REV. 1 0/11)

z

"' UJ z cc

£ 0 zr Cl) 0

"' mm Cl) )>

~

CD en :n g G>m

D z om C :D

.,, G>

G')

m z:

O. 576 MIN - m 0.125R

a 0.0045 l>

- 0.125R 12.0 5 <<j, l

• m r-m n

-t

x, 9.676 <<j, ----------L 10° MAX n

10. 8 0 <<j, l

0 N m N

9. 390 <<j, < ~

l 8. J 8 <<j, 1

_L ....

-I:'-

Vt

~

4.06 0.50 NOTE: DASHED LINES INDICATE REQUIRED BLENDING FIGURE 30.J.l DEVIATIONS DESCRIBED IN DDR NO. 15139 .... 00 Q\ I 0\

00

8-169 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 168

• BUSINESS GROUP IIIEV l TABLE 30.3.1-2 PRIMARY PLUS SECONDARY STRESS ANALYSIS (All stresses in ksi) r.

After Deviations Before Deviations Location p + Q p + Q* p + Q p + Q* Allowable F 62.3 37 .23 61.68 36.61 SS.8 G 62.82 45.94 62.29 45 .42 SS.8

• Thermal Bending Removed TABLE 30.3.1-3 FATIGUE ANALYSIS After Deviation Before .Deviation

• • Location F

G Fatigue Usage 0.439 0.41 Fatigue Usage 0.279 0.409

• NEBG-807A (6/10)

8-187 22A74S4 186 NUCLEAR ENERGY . GEN ERA L

• ELE CTR I C SH NO

• REV 1

• BUSINESS OPERATIONS

.25R 12.00 t

j_

I

\ B 10.75 ~

l.

8.413 ¢ I

I

...J..... _., .50L 8. 378¢ l

7 8.378 ¢ DETAIL B l

• FIGURE 30.3.4.1- 2 SECTION G DEVIATION AFTER BLENDING NEO 107A (REV. 10/11)