ML21070A122
| ML21070A122 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/11/2021 |
| From: | Xcel Energy, Northern States Power Company, Minnesota |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21070A105 | List:
|
| References | |
| L-MT-21-004 | |
| Download: ML21070A122 (406) | |
Text
{{#Wiki_filter:MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-H REACTOR PRESSURE VESSEL DESIGN
SUMMARY
REPORT Revision 29 MONTICELLO APPENDIX H REACTOR PRESSURE VESSEL DESIGN
SUMMARY
REPORT MONTICELLO NUCLEAR GENERATING PLANT MONTICELLO, MINNESOTA UNIT 1 NOTE: The effects of increasing reactor power to 2004 MWt were evaluated in Task Report T0302, Reactor 01101248 Vessel Integrity - Stress and Fatigue Evaluation, to ensure that the reactor vessel components continue to comply with the existing structural requirements of the ASME Section III Boiler and Pressure Vessel Code. Reference "Task Report T0302, Reactor Vessel Integrity-Stress and Fatigue Evaluation," (Monticello calculation number 11-223), for further information. H-i
MONTICELLO TABLE OF CONTENTS PAGE DESIGN REQUIREMENTS H.1-1 EXHIBITS:
- 1. RPV Purchase Specifications 21A1112 1-1
- 2. Manufacturer's Data Report and Vessel Certification 2-1
- 3. Tensile Tests - Specimens of 80% Plate Thickness 3-1
- 4. Summary Stress Report 4-1
- 5. Vessel Fabrication and Assembly Report 5-~
- 6. Inde~endent Stress Analysis Report 6-1
- 7. Reactor Vessel Design Specification (Repair). 7-1
- 8. Reactor Vessel System Cycling (Stress Report) 8-1
- 9. Reactor Vessel Rapid Cycling (Stress Report) 9-1
.)
H-ii REV 5 12/86
MONTICELLO DESIGN AND FABRICATION REQUIREMENTS The Monticello reactor vessel was designed, fabricated, inspected, ar.d tested in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III,* Nuclear Vessels 1965 Edition and Addenda to and including Summer 1966 Addenda, and the following additions:
- 1. ASME SA533 plate and Inconel material per Summer 1967 Addenda.
- 2. Main closure flange material per Code Case 1332-3.
- 3. Studs and nuts material for main closure flanges per Code Case 1335-2.
- 4. Main closure flange and stud shank transition radius per Code Case 1366.
- 5. Bearing stresses for stabilizer brackets and coefficients of thermal expansion per Winter 1967 Addenda.
- 6. Hagnetic particle and liquid penetrant examination per {.j'j_nter 1966 Addenda.
The date of the contract between the Buyer, General Electric Company, Atomic Power Equipment Department, San Jose, California and the Seller, Chicago Bridge and Iron Company, San Francisco, California, was July 18, 1966. There are no deviations to the formal code throughout the design, fabrication, inspection and testing of the reactor vessel. Design fabrication, inspection, and test requirements in addition to those required by the B&PV code were required by the Buyer's vessel purchase specification 21Al112 (Exhibit 1). These include but are not limited to the following pertinent inspections andlor tests:
- 1. Established specific maximum nil ductility transition temperatures for the main closure flanges and the shell and head materials connecting to these flanges (+lO°F c-."'DT temperature) and elsewhere (+40°F NDT temperature).
- 2. A fabrication test program on vessel shell material which included testing of large size tensile specimens (80% of the vessel wall thickness in diameter) both plain and welded samples. (See Exhibit 3).
- 3. Provisions are made for determining the effects of nuclear radiation upon the reactor vessel structural materials by supplying specimens of the vessel material to be exposed to the core irradiation at the vessel wall inside of the vessel.
Pertinent certifications are contained in Exhibit 2, Manufacturer's Data Report and Vessel Certification, Chicago Bridge & Iron Company. The sUl!'lIlary of results of the detailed stress analysis is contained in Exhibit 4. H.l-l P.E~: '" 12! 8 5
MONTICELLO Plans for the vessel fabrication and assembly were described in Amendment 2 to the FDSAR, "Design Fabrication and Erection of the Reactor Vessel." Actual fabrication and assembly was ~n accord with Section IV of Amendment 2 except for minor modifications as listed in Exhibit 5 of this report. The GE quality control of the reactor vessel was essentially as described by General Electric Quality Control Plan, Section IV of Amendment 2 to the FDSAR, except the Domestic Turnkey Projects organization of General Electric Co. also made an independent QC audit. A detailed seismic analysis of the Reactor Pressure Vessel was prepared by John Blume & Associates and was included in Appendix A along with other seismic analyses. In 1977, repairs were made to the reactor pressure vessel feedwater nozzles and safe ends to minimize damage to the feedwater nozzles due to thermal cycling. The repairs consisted of removing cladding from the nozzle blend radius and bore and the installation of a feedwater sparger interference fit thermal sleeve with a piston ring seal. These design changes invalidated the "Summary of Stress Analysis for the feedwater Nozzles" shown on page 4-14 of Exhibit 4. Details of this repair and design are contained in Exhibit 7. Also in 1977, a design change modified the CRD return line because of its susceptibility to intergranular stress corrosion cracking. The 3" CRD return line and the reactor vessel nozzle safe-end forging were removed and the nozzle was capped using a 4" diameter schedule 120 pipe cap. This design change eliminated the imposed mechanical loading for the nozzle, creating a much less severe condition than the nozzle was originally designed for. As a result of this modification, the "Summary of Stress Analysis for the 3" CRDHSR Nozzle" shown on page 4-18 of Exhibit 4, is invalidated. Details of the modification and new stress analyses are contained in design change 77M069. In 1981, new feedwater nozzle safe ends featuring a tuning fork design with a welded in thermal sleeve were installed and a section of piping upstream of each nozzle was replaced with piping of a different material. These modifications were performed to provide a significant reduction in thermal cycling of the feedwater nozzle area. The new stress analyses that replaced the "Summary of Stress Analysis for the Feedwater Nozzle" shown on page 4-14 of Exhibit 4, are contained in Exhibit 8 and Exhibit 9. In 1984 several modifications were incorporated to provide greater resistance to intergranular stress corrosion cracking. The core differential pressure and standby liquid control safe end was replaced using a safe end of similar design, but with different materials. The new stress analyses are contained in General Electric Stress Report No. 23A4llS, included in Design Change No. 83Z049C. H.1-2 REV 7 12/88
- 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 AS08 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 and 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 rGSCC 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. 84Z068, 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. 23A1627, 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 30l-P-S. 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. 23ASSS3, included as part of Design Change No. 86Z0l6, documents the analysis . H.I-3 REV 7 12/88
THIS PAGE INTENTIONALLY LEFT BLANK H.1-4 REV 7 12/88
MONTICELLO 1-1 EXHIBIT 1 REACTOR PRESSURE VESSEL PURCHASE SPECIFICATION 21A-1112
- REV 4 12/85
1-2 GENERAL. ELECTRIC NUCLEAR ENERGY DIVISION Document No. 2lAll12, Rev. 6 TRANSMITTAL General Electric Class pRoJEcT(s) ____~M~O~N~TI~C~E~L~L~O_________________________________________________ TITLE OF DOCUMENT______~RE~A~CT~O~R~P~RE=S~S~U~RE=_VE~S~S~E:S_______________________________________ TYPE OF ttl PURCHASE SPECIFICATION REPLACES DOCUMENT: [] SYSTEM DESIGN SPECIFICATION DOCUMENT NO . _______________ [] INSTALLATION SPECIFICATION []------------------- PIPING OR COOLING SYSTEM INVOL VED_____________________________________ RESPONSIBLE ENGINEER~_D_R_H_E_IS_I_N_G _______ ISSUED BY JA MAST DATE____ MAR __ ~I i969 REFERENCES MASTER PARTS LIST (MPL) NOS. 21A982l - Stud Tens~oner SPECIFICATIONS____~2.J~AJuO~5~O~-_S~t~d__R~e~q~I!~;.r~em~e~Qwt**~fQ~r~B~e~ac~t~our~Se~nuuri.c~iD~g~T~Q~Q.l~s_ __
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DRAWINGS 107C5305 - Preparation of Nozzles 885D9ll - Bolting 886D482 - Reactor Vessel 117B1550 - 1/4" Tensile Test Specimen l17B1549 - Charpy Impact Speciml~lE678 - Vessel As-Built Dimensions REVISION RECORD REVISED PER (XD., ECN, XXX.)~#~9~6~0~.--~-----~~~~_=~~~~~~-~ SHEETS AFFECTED 16 - 21 and Attachment E REVISION IDENTIFIED WITH~ COMMENTS: DISIiIBIlIIOH NAME MAIL CODE
---- ---- COPIES NAME MAIL CODE COPIES DR HEISING 743 1 RL CALL 743 1 AC DE LOACH 621 1 GRU 742 1 Alexander 366 6+2R Wolf 377 2 Vassar 350 1 Skarpe10s 359 1 Lingafelter 350 1 375 1 591 1 595 1 624 1 711 1 722 1 723 1 743 3 761 1 10/4/68
1-3 GENERAL _ ELECTRIC SPECIAL PROJECT ATOMIC POWER EQUIPMENT DEPARTMENT MONTICELLO SPEC. NO. 21All12 REV. NO. 6 PURCHASE SP ECI FICA TION SH NO. 1 CONT ON SHEET 2 t TITLoE REACTOR PRESSURE VESSEL TABLE OF CONTENTS
~ TITLE PAGE LO SCOPE 2 2.0 RESPONSIBILITY 2 3.0 GENERAL DESCRIPTION 2 4.0 CODES 3 S.O DESIGN REQUIREMENTS 4 6.0 DESIGN ANALYSIS 7 --
I I 7.0 CONSTRUCTION 10 I I 8.0 MATERIALS 14
\
9.0 FABRICATION 19 22 -- 10.0 INSPECTION AND TEST 1l~0 SHIPMENT 32
. ~ : ".
12.0 SUBMITTALS 32 ATTACHMENT A - INSTRUCTION MANUAL, DRAWING & DATA REQUIREMENTS ATTACHMENT B - MATERIAL TESTS AND TEST SPECIMENS ATTACHMENT C - DESIGN ANALYSIS SCHEDULE FOR REACTOR PRESSURE VESSEL FOR MONTICELLO POWER STATION ATTACHMENT D - TEMPERATURE TRANSIENTS ATTACHMENT ! - CERTIFICATION OF DESIGN SPECIFICATION
- rJY*M~
---'y DR REISING 3-'14'1 MAR - ~
IUUEO:J:: A ~T 1~~q
.J # "'7
1-4 GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPEC. NO. 2lAll12 REV. NO 6
- PURCHASE SP ECI FICATION 5 .. NO. 2 CONT ON SMEET 3
- 1. a ..§.£QU 1.1 Thi ** pecification defines the engineeri~~ requirements of the equip~ent specified herein.
1.2 The work done by the Seller in accordance with this specification shall include all necessary design, development, analysiS, drawings, evaluation of materials and fabrication methods, shop fabrication, shipment, field erection, inspection, and testing. 2.0 RESPONSIBILITY The Seller *shall accept full responsibility for his ~ork and for compliance with this specification. Review or approval nf drawings, procedures, data or speci-fications by the Buyer with regard to general design and controlling dimensions does not constitute acceptance of any designs, mater1als or equipment which will not fulfill the functional or performance requirements established by the purchase contract.
. 3. a GENERAL DESCRIPTION 3.1 The reactor vessel will be used as a pressure container supporting the steam generating core in the Monticello Nuclear Power Station to be located near Monticell~Minnesota.
3.2 The equipment to be furnished in accordance with this specification shall be one reactor pressure ves3el assembly with a removable head and nozzles and certain internal support structures, arranged as sholo.n on Drawing 8860482 complete with: 3.2.1 Attachments for thermal insulat~on, vessel and core supports, brackets or legs for li.ftin& and handllllg of the vessel head, and mounts for outside surface thermocouples. 3.2.2 One set of necessary special tools required to remove and replace the reactor vessel head. The set of tools shall include: four hydraulic stud tensioners, stud elongation measuring deVice, stud and nut wrenches, one set of stud thread protectors, three head guide caps, one bush~ng wrench, one stud sling. Stud*tensioners shall be in*accordance w1th Specific... ;..*' 2lA9821 and shall indude a lifting device that properly spaces the tensioners over the bolt circle. ISSUEO: MAR* 1969
1-5 GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPEC, NO, 21A11l2 " .. v, NO, 6 PURCHASE SP ECI FICA TIOH 3 4 S ... .... 0. CONT ON S"EET 3.2.3 ODe set of necessary special tools required to install and remov~ the reactor vessel head seals with manual contact. This set of tools shall include a protective cover for' the ,reactor vessel shell flange seal surface. 3.2.4 Metal boxes for the hand tools. Boxes shall be suitable for hand-ling with a crane and/or fork lift truck. 3.2.5 One lot of reactor vessel material test plate and material test specimens in accordance with Attachment B. 3.2.6 Shipping skids for those portions of the Reac~or Vessel which are shop fabricated. 4'.0 ~ 4.1 The reactor vessel shall be designed, fabricated, ,inspected, tested and stamped in accordance with the American Society of Mechanical Engineers (ASHE), Boiler and Pressure Vessel Code, Section III, applicable requirements for Class A Vessels as defined therein, interpretations of the ASHE Boiler and Pressure Vessel Code, and all laws, rules and regulations of the State of Minnesota in effect on the date of the contract. 4.2 Deviations from the applicable codes or regulations shall be avoided.
'Where a conflict exists among the codes or regulations, the Seller shall bring this to the Buyer's attention. It shall be the responsi-bility of the Seller to obtain resolution and disposition of deviation with the Buyer and other appropriate parties and authorities.
4.3 The intent of this s~ecification is to supplement the requirements of the codes specified herein and to encompass the means ~hereby the design objective is satisfied. 4.4 All standards and material specifications shall be per latest reviSion in effect on the. date of the contract .
- ISSUEC:
MAR - ~ 1969
1-GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPEC . . . 0. 21All12 ~EV. "0* 6 PURCHASE SPECIFICATION S'" NO. 4 co .. ? 0 .. S'"'EE"T" 5 5.0 DESICN RlQUlREMENTS 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: Normal Operating Temperature: 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 ASHE Code, Section III. These pipe reactions shall be used in the detailed stress analysis required by the Code and performed 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 ~nd Reaction The momentary reactions whjch are suddenly applied to each control rod drive housing in the vessel hotto~ head are presented on Drawing 8860482. ISSUEC: MAR - 5 1969
1-7 e GENERAL ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION SPEC. NO. 2lAll12 REV. NO. 6 s .. NO. 5 CONT ON S"EET 6 5.1. 7 Steady State Thermal Conditions Steady state me~G. ;emperatures will be computed by the Buyer for no more than twelve locations on the reactor vessel. The locations will include the head and shell closure flanges, the shell adjacent to the reactor core, the bottom head and major nozzles including the control rod drive nozzles. Temperature gradients through the shell wall adjacent to the portion of the reactor core peak flux zone will be computed by the Buyer and furnished to the Seller. 5.1. 8 Cyclic Loading The thermally induced stresses which result from the transients listed in Attachment D shall be computed for the components listed. .The cyclic stress ranges which result from these and the following conditions shall be evaluated in a fatigue analysis according to the ASME Code Section III. The additional conditions are: a) Zero stress condition b) Isothermal condition at 546°F and 1000 psi inside vessel. c) Isothermal conditions at 70°F and 1000 psi inside vessel 120 cycles. d) tor the closure flanges and bolting - the cold bolt-up condition 120 cycles. 5.1.9 Earthquake Loads . Earthquake loads shall be taken into account in accordance with the criteria and load presented on Drawing 886D482. 5.2 Design Considerations 5.2.1 Design Objective The objective shall be to design and fabricate this reactor vessel to have a useful life of forty years under operating conditions specified by the Buyer. 5.2.2 Reactor Vessel Supports Reactor Vessel supports, internal supports, their attachments and adjacent shell shall be designed to take maximum combined loads including control rod drive reactions, earthquake loads, and jet reaction thrusts as defined on Drawing 8860482. There shall be no gross yielding of the reactor vessel supports causing permanent displacement under these conditions. 5.2.3 Stress Concentrations Care shall be taken in design and fabrication to minimize stress concen-trations at changes in sections or penetrations. Fillet radii shall be equal to at least half the thickness of the**thinner of the *two sections being joined. If reinforcement for openings (except the control rod drive and in-core flux monitor nozzles) requires local vessel shell added thickness, such reinforcement shall extend at least 1-1/2 times the diameter of the opening from the center of the opening. These requirements are not to be construed as a waiver for evaluating the stresses for use in the analysis for cyclic operation. ISSUEO: MAR - 5 1969
1-8 e GEN ERAL ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPEC. NO. 21A1112 "EY. No.6 PURCHASE SPECIFICATION S ... NO. 6 CONT ON SHEET 7 5.2.4 Corrosion Allowance Exterior exposed ferritic surfaces of pressure-containing parts in-cluding heads, shell, flanges and nozzles shall have a minimum corro~ sion allowance of 1/16 inch. The interior surface of carbon or low alloy steel parts exposed to the reactor coolant shall also have a minimum corr08~on allowance of 1/16 inch. If the main closure head is left unclad, its interior surface shall also have a minimum corrosion allowance of 1/16 inch. 5.2.5 Main Closure Seal The reactor pressure vessel main closure seal shall be a double seal designed to have no detectable leakage through the inner or outer member at all operating conditions. These conditions include, but are not limited to: (a) cold hydrostatic pressure test at the design pres-sure, (b) heating to design pressure and temperature at a rate of lOOoP/hr., maximum, (c) operating for extended periods of several months duration at operating conditions, and (d) cooling at a rate of lOOoP/hr., maximum. 5.2.6 Design Stress Design stress values used in the calculations shall be as contained in * . i ASHE Section III and applicable interpretations of ASME Boiler and Pres-sure Vessel Code for materials covered therein. The design stress values for ASHE, Section III calculations for other materials approved by the Buyer in accordance with Paragraph 8.1 of this specification shall be determined per Appendix II, ASHE Code,Section III. 5.2.7 Dimensional Control Seller shall show the method of controlling measuring and maintaining alignment and location of control rod drive penetrations with the vessel and core supports. 5.2.8 The reactor shall be designed to minimize retention pockets and crevices. ISSUEO: MAP. -
~I 1969
1-9 GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION SPEC. NO. 21A1112 "EV. No.6
..... NO.
7 CONT ON 'MEET 8 6.0 DESIGN ANALYSIS 6.1 Requirements the Seller and the Buyer shall perform the design calculations and analyses as required by the applicable Standards and Codes indicated in Section 4.0. The requirements of Article 4, ASHE Code, Section III, shall be fulfilled. The division of responsibility for the analyses shall be in accordance with paragraph 6.1.3. The analysis required shall be performed in two divisions as follows: 6.1.1 Stress Analysis A stress analysis shall be performed in accordance with Section N-430, ASHE Code Section III. Calculations shall be performed in accord-ance with paragraph N-431 to verify that the minimum wall thickness is provided. A detailed stress analysis shall be performed in accordance with paragraph N-432. This analysis shall take into account all combinations of loads in conjunction with metal temp-eratures, as indicated in Section 5.0 above, and Drawing 8860482 within the Design Stress Criteria of ASHE Code Section. III, Article 4. 6.1.2 Analysis for Cyclic Operation An analysts shall be performed in accordance with Section N-4l5 of the ASHE Code, Section III, to determine that the vessel is suit-able for the cyclic loading conditions of paragraph 5.1.8 above. This analysis shall also be performed within the design stress criteria of Section III, Article 4, to establish whether the design objective in paragraph 5.2.1 above is reached: The analysis will be used to determine the adequacy of any required thermal baffling used to control or limit thermal stresses and to place safe operating limits on the cyclic conditions imposed on the vessel where it is reasonable to control them, as in the start-up heating rate and shut-down cooling rate. 6.1.3 Division of Resoonsibilitv The Sell~r and the Buyer shall perform jointly the design analysis required by this speclfication. 6.1.3.1 The seller shall perform calculations to satisfy limits on primary general membrane stress (PM)' primary local membrane stress (P L), primary bending (P ), and secondary membrane plus bending stresses B (except thermal stresses) (Q) from specified steady state conditions. Also included are calculations necessary to reinforce openings per Paragraph N-450, except the calculations necessary to satisfy the cyclic conditions and Paragraph N-45l (b) which are the responsibility of the Buyer. I"j.'/A~' - 5 1969
1-10 GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SP ECI FICA TION SPEC. NO. 2lAll12 REV. NO. 6 s .. NO. 8 CONT ON SHEET 9 6.1.3.2 The Buyer shall perform the transient and steady state thermal analysis and the analysis for cyclic operation on all components requiring such analysis. These analyses will provide the stress categories Q and F of Paragraph N-4l4, ASME Section III. This type of analysis will cover but not necessarily be limited to the following parts of the reactor vessel:
- a. Emergency cooling nozzles (safe end and thermal sleeve)
- b. Feedwater nozzles (safe end and thermal sleeve)
- c. Control rod drive hydraulic system return nozzle (safe end and thermal sleeve)
- d. Vessel Support Skirt
- e. Refueling bellows support skirt
- f. Closure flanges
- g. Bolting
- h. Control rod drive penetration 6.1.3.3 The analyses which are the responsibility of the Buyer but are made with the Sellers assistance, shall be checked and signed by the Buyer.
6.1.3.4 The Seller shall fulfill the requirements of Paragraph 4.0 Codes, and produce the summary report required by Paragraph 6.8. The Buyer shall prepare its portions in suitable form for reproduction. 6.2 Calculation of Stresses The detailed structural analysis required to meet the requirements of 6.1 shall be made for the stresses reSUlting from internal pressure, external and internal loadings, and the effects of steady and fluct-uating temperatures and loads for regions given in 6.3 which involve changes of shape, structural discontinuities, and points of con-centrated loadings. Where dimensions and loading conditions permit, the adequacy of structural elements will be verified by comparison with completely analyzed elements. The c3lculations shall include a complete I"'l1'lIR - 5 1969
1-11 GENERAL _ ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION SPEC. NO. 21All12 REv. NO. 6 SH NO. 9 CONT ON SHEET 10 analysis of stresses under stea~'" ~~ate and transient conditions to determine suitability of the design .lth respect to the allowable stress given in ASHE Code, Section III, and to determine the opera-tional 11mitations. 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 Analyze1 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 5.2.5 above, the relative rotations of the flanges shall be calculated. These rotations shall be used to demonstrate analytically satisfactory seal performance using t~e following assumptions: 6.4.1 The mating 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 O-ring seal locations, the separation between flanges is less than the minimum elastic spring-back of the a-ring. 6.5 Calculations The calculations shall be clear and in sufficient detail to re~it 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:
- ISS'MAR - 5 1969
1-12 CD ELECTRIC GENERAL ATOMIC POWER EQUIPMENT OEPARTMENT PURCHASE SPECIFICATION
"tEe ... 0. 21All12 REV. "0. 6
, .. "0. 10 CO .. T 0" 'HEETll 6.6 Descriptions of Computer Programs If computer programs are used to obtain solutions to design problems, the Seller shall furnish the Buyer the description of each different computer program used. These descriptions shall be furnished with the first issue of the design calculations incorporating such programs.
The computer program description shall include computer type, program capabilities, assumptions, limitations and statement of availability. 6.7 Measurement Reports Measured values of strain, deflections or stresses resulting from tests on models or actual reactor vessels shall be supplied to the Buyer by the Seller. These reports shall include all information necessary to duplicate the conditions required to obtain the results reported. 6.8 Summary Re~ort After completion of the reactor vessel design, the Seller shall furnish the Buyer additional copies of all calculations plus a summary report of results of all computations. Each copy shall be bound in a suitable paper binding and indexed. 7.0 CONSTRUCTION The reactor vessel body including all components which contain pressure including the shell, lower and upper heads shall be made of rolled plate and/or forgings welded with full penetration welds throughout except as noted in 7.3.5. The shell and head flange and nozzles shall be for~ed. 7.1 Shell and Heads 7.1.1 Longitudinal and circumferential weld joints in the reactor vessel shall be oriented so as not to intersect openings or penetrations, wherever practical. Circumferential weld seams should avoid re~ions of highest n~utron flux in the cor~ region, if practic3l. The region of highest neutron flux occurs between the mid-rlane and top of the core. ISSUEO: MAR - 5 1969
1-13 e GENERAL ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION S"EC. NO. 2lAll12 IOEV. NO. 6 5 ... NO. 11 CONT ON S"EET 12 7 .1.2 Bottom Head The section of the bottom head which encompasses the penetrations for the control rod drives and in-core flux monitors shall be either a single forging or dished plate, if practical. If this is not practical and a weldment is used, the orientation of the weld sections shall as far as practical minimize the number of inter-sections of weld seams with penetrations. 7.1.3 Top Head The top head shall be either a single forging or dished plate or shall be fabricated of sections welded together, with the orienta-tion of the weld seams such that no seams intersect openings or penetrations. 7.1.4 Weld Joints Weld joints shall be designed to facilitate a maximum of radiographic examination per the ASHE Boiler and Pressure Vessel Code, Section III,
-paragraph N-624.
7.2 Head Closure 7.2.1 Assembly and Disassembly 7.2.1.1 The head closure shall be designed for removal and reassembly, using 4 or more hydraulic stud tensioners. 7.2.1.2 It shall be the design objective to replace and remove the head within 16 hours elapsed time. Specifically, the cycle shall include placing the head over the studs, tightening the studs to operating bolt-up loads, unbolting and removal of the head over the studs. It is expected that 120 such cycles will be performed during the life of the reactor vessel. 7.2.2 ~ 7.2.2.1 The head seal shall be a double seal with a vent between the seals through which leakage of the inner ring can be detected. The seal vent shall be designed for full design pressure of the reactor vessel. 7.2.2.2 The seal shall be metal O-ring type with pressure equalizin~ vents on 1. D. 7.2.2.3 The grooves for the O-rings shall be placed in the reactor head flan~e. Suitable fasteners shall be provided to hold the O-rings in the grooves during head removal and assembly operations . ISSUEC: MAR - 5 1969
1-14 GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT
,PEC. NO. 2lA1ll2 REV. NO. 6 PURCHASE SP ECI FICATION , .. NO. 12 CONT ON S"EET 13 7.2.2.4 Provisions shall be made for installation of a low pressure leak detection system outside of the second seal, and may be outside of the bolt circle. The provisions shall include a vent through the vessel flange with extended 1" nipple and socket weld fitting and either a shallow groove or other suitable backing to retain a soft asbestos braided packing. There shall be no protruding parts of this low pressure seal beyond the 0.0. of the head and vessel flange.
7.2.3 Bolting 7.2.3.1 Studs shall be used to secure the reactor vessel head. Stud, nut and bushing threads shall be in accordance with Drawing 8850911. 7.2.3.2 The stud bolt hol~s in the reactor vessel flange shall be bushed with removable bushings. Keys shall be provided for each bushing to prevent rotation of the bushings when removing studs. 7.2.3.3 Spherical washers shall be used with the studs to minimize bending of the studs. 7.2.3.4 It shall be possible to remove and replace the head I with the studs install~d. To facilitate head removal and replacement, three special "guide caps shall be provided to couple onto three studs. The lengths of the guiding surfaces of the guide caps shall be staggered so that the shorter of the three guide caps shall extend above the top of the installed studs for a minimum distance of 4 inches. The length of the three guide caps shall be staggered in 3-inch minimum increments. The internal threads of the guide caps shall be similar to the stud nuts threads. The upper end of the guide caps shall be provided with a conical lead-in taper and a horizontal through-hole bored to accommodate a round bar for wrenching. 7.2.3.5 Flange hole, bushing, and stud designs shall be such that the studs stand perpendicular to the flange surface when the studs and bushings are bottomed in the holes to facilitate removal and replacement of vessel head over studs as called for in Paragraph 7.2.3.4. 7.2.3.6 The surface of all threads in the studs, nuts and bushings shall be given a phosphate coating to act as a rust inhibitor and to assist in retaining lubricant on the surfaces. An approved lubri-
~~nt s~ould be applied to the stud threads as soon as possible after coating.
MAR - 0 1969
1-15 e GENERAL ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT S"EC. NO. 21A1112 REV. NO. f, PURCHASE SP ECI FICA nON SH NO. 13 C:ONT ON SHEET 12 7.2.3.7 A stud sling for the main closure studs shall be provided. The
.tud .ling 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 ease 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 7.2.4.1 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. 7.3 Nozzle Ends 7.3.1 The ends of all nozzles other than flanged nozzles shall be prepared for welding in accordance with Drawing I07C5305. 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 B3l.l for establishing the required piping wall thicknesses.
Nozzle safe end wall thickness shall be governed by Drawing I07C5305 and will in general be greater than required by Section III.
- ISSuEO:
MAR - 5 1969
1-16 e ELECTRIC GEN ERAL ATOMIC POWER EQUIPMENT DEPARTMENT
.pce:. NO. 2lAl1l2 REV. N0.6 PURCHASE SPECIFICATION '04 NO. 14 e:ONT ON SHEET 15 7.3.4 o.ta1l. of the transition weld preparation Ihall be lubmitt~d to the Buy.r for approval.
7.3.S Nozzle. of 3" nominal size or .larger shall be full penetration welded to the vessel. Nozzle. les. than 3" nominal size may be partial penetration welded if permitted by ASHE Code, Section III. 7.4 The v.... l top head nozzles shall be provided with 1500 pound weld neck flang ** with small groove facing. Hating 1500 pound flanges with small tongue facing, gaskets and a complete set of studs and nuts shall alsn be provided. The loose flanges for the 6 inch instrument nozzles shall be blind. the remainder shall be weld neck. The flanges and gaskets
.hall be in accordance with ASA Standards B16.5. The threads on studs aDd nuts .hall* be 8-pitch series in accordance with ASA Standard Bl.1.
7 *.5 R.actor Ve.sel Support. 7 * .5.1 Ext.rnal and internal supports shall be provided as an integral part of the reactor vessel. The loca:ion and design of the supports shall be such that stresse. in the reactor ve.sel and supports will be within ASHE Code limit. due to reactions at these supports. The de.ign pressure differential across the core shroud support shall be 100 p.i (higher pressure under the support) occurring at the ves~el ' d.sign temperature. The design of the core shroud support shall also / take into account the restraining effect of the components attached to the .upport and the weight and earthquake loading &s shown on Drawing 8860482. 7.3.3 The drain nozzle shall extend 12 to 16" below thebnttom of tht:. reactor vessel and shall be of the full penetration design. 7.6 External Attachments 7.6.1 Brackets to support insulation shall be provided on the exterior of the reactor vessel 1n accordance with Drawing 8860482 7.6.2 Provisions shall be made for the attachment of thermocouples in mounts on the reactor vessel exterior as specified on Drawing 886D482. 8.0 MATERIALS 8.1 All materials to be used shall be indicated on the Seller's drawings. The Seller shall submit for the Buyer's approval, all material
**lections and material purchasing specifications.
ISSuEO: MAR - 5 1969
1-17 GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION S"EC, NO, 2lAll12 ItEV, NO, 6 So. NO. 15 CONT ON S"EET16 8.2, Records The Seller shall maintain complete recorda showing use of all materials so that it will be possible to relate every c01I1ponent of the finished reactor vessel to the original certification of the material and the fabrication history of the component. The Seller shall prepare a summary of the heat number. chemical composition and mechanical proper-ties for each reactor vessel c01l1ponent. 8.3 Forgings Low alloy steel forgings for pressure parts shall be made in accordance with ASIM AS08 in accordance with ASHE Code Case l332~3. Paragraph 5. Nozzles which are partial penetration welded as specified in 7.3.5 may be nickel-chromium-iron forgings ~ade in accordance with ASME SB-166 modified in accordance with Code Case 1336. The molten steel shall be vacuum treated prior to, or during, the pouring of the ingot in order to remove objectionable gases, particularly hydrogen. 8.4 ~ Plate for pressure parts shall'be in accordance with ASIM AS33, Class I Grade B, Firebox Quality, in accordance with ASHE Code Case, 1339-2
- Plate ingots shall be produced by vacuum degassed pouring.
8.5 Castings The use of castings will be considered by the Buyer but specific Buyer approval shall be required. Castings for pressure parts shall be made in accordance with ASHE SA-356, Grade la, Code Case 1333, Paragraph 1. 8.6 Material for pressure parts shall be selected and worked to produce as fine a grain size as practical. It shall be an objective of the fabrication' technique to retain a grain size of 5 or finer in all material. Grain size shall be determined by the method in ASME El12. 8.7 Heat Treatment Heat treatment of carbon and low alloy steel pressure par'ts shall consist of normalizing and then tempering at not less than l200°F. For section thickness over 3 inches nominal, heat treatment shall consist of accel-erated cooling from the austenitizing temperature to below ,the martensite finish temperature followed by tempering at not less than l200°F to obtain tensile and impact properties c01l1parable to those developed by normalizing and tempering section thickness of less than 3 in. nominal. 8.8 Mechanical Properties The low alloy steel forgings, plate and castings for pressure parts shall be tested in accordance with Paragraph 10.3 and shall have the mechanical properties required therein in addition to those required by the applicable ASKE Sp*ecification* ISSUED: MAR - 5 1969
1-18 GENERALe ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPEC:. NO. 21A11l2 "EV. NO. 6 PURCHASE SPECIFICATION S~ NO. 16 C:ONT ON S"EE' 17 8.9 Studs. Nuts. Bushings. and Washers for Main Vessel Closure 8.9.1 Studs shall conform to ASTM A540, Grades B23 or B24 and ASME Code Case 1335-2, Paragraph 4, Class 3, 4 or 5. 8.9.2 Nuts, bushings and washers shall conform to ASTM A540, Grades B23 or B24, and Code Case 1335-2, Paragraph 4, Class 3, 4, or 5 but to suit the stud material used. It shall be the objective to have a minimum difference in hardness* of 5 Rockwell C points from the stud material. 8.9.3 Hardness and impact properties shall meet the requirements of para-graph 10.3.2.5. 8.10 Cladding Material All internal carbon and low alloy steel surfaces of the reactor vessel including the closure head and closure head flange mating surfaces, shell flange and mating surface, shell, bottom head, nozzles for con-necting stainless steel piping, and internal attachments shall be clad with weld overlay meeting the following requirements: 8.10.1 Weld overlay cladding shall be a minimum of 0.125 inches total thick-ness. The finished surface shall have a composition equivalent to ASTM A371, Type ERJ08 or A240 Type 304 except the carbon content shall not exceed 0.08%. 8.10.2 Cladding in the "as-clad" condition is acceptable, provided the resulting surface finish does not interfere with the ultrasonic and liquid penetrant test requirements. 8.10.3 The sealing surfaces of the reactor vessel head and shell flanges shall be weld overlay clad with austenitic stainless steel which consists of a minimum of two layers and a minimum of 0.25 inch total thickness. The first layer shall be deposited with an analysis equivalent to ASTM A371, Type ERJ09. The second and sub-sequent layers shall have a composition equivalent to ASTM A37l, Type ER308, except the carbon content shall not exceed 0.08% *. Minimum thickness of 1/4 inch shall apply after all machining, including area under groove. 8.11 Attachments 8.11.1 Internal attachments other than the weld clad ferritic attachments shall be annealed stainless steel, Type 304 per ASTM A240 or ASTM A276, or Type F304 per ASIM A182. The core support structure shall be stainless steel clad low alloy or carbon steel, solid nickel-chromium-iron alloy per ASK! SB166, 167, or 168, or annealed stainless steel, Type 304 per ASIK A240 or ASTM A276, or TypeF304 per ASIM A182 . ISSUIi:O: l~hJ( -~. tCO*'g oJ
1-19 GENERAL fj ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION SPEC. NO. 21A1112 REV. NO. 6
... NO. 17 CONT ON SHEET 18 8.11.2 External attachments to the reactor veasel shall be of the same material as the reactor vessel base material. or shall be of a material which has mechanical and impact properties compatible with the base material. tlhere welds must be made to the attach-ments in the field. the ~aterial selected shall not require rre-heat or post-weld heat treatment.
8.12 Nozzle Safe ~ds and Flanges 8.12.1 Nozzle ends for austenitic p.j.pe shall be ."Snf A336 *. Class F8 or F8m; A240. Type 304. or Type 316; or A376. Type 304 or Type 316 solution heat treated stainless steel. depending upon the mating pipe material selected by the Buyer. Nozzle ends for carbon steel pipe shall be ASTM AlOS. Grade II. forgings except phosphorous content shall be 0.035% Max. and sulphur 0.040% Max; ASTM A508 Class I; or ASTM AS16 Grade 70. Proportions shall be as shown on Drawing 107C530S. 8.l2.2Standard flanges for flange~ nozzles and separate mating flanges shall be ASTM AI82. Grade F304. stainless steel
- li.l2.3 Studs for standard flanges shall be SAl93, Grade B7. Nu'ts for standard flanges shall be SAl94, Grade 2H.
8.13 Pipes and tubes shall be ASTM Al13, A249. A312, A376, solution heat treated, Grade TP304 or TP3l6; or Al40. Type 304 plate welded and radiographed in accordance with ASHE Code. Section III. Paragraph N624. 8.14 Miscellaneous bolting material shall be subject to the Buyer's approval. 8.15 Weld Electrodes and Rods 8.15.1 Material for weld electrodes and rods shall be selected from ASlli> A233, A298, A3l6, A37l or eqUivalent for other processes and reported to the Buyer for approval. 8.15.2 All austenitic stainless steel welds and weld cladding shall con-tain controlled amounts of ferrite, confirmed by quantitative tests. The procedures for control of. and testing for the ferrite content of welds and weld cladding shall be submitted to the Buyer for approval. The acceptance standard for quantitative tests shall be either % Cr - 1.9 x % Ri, or 5% ferrite mint=um *
- IssuEO:
MAR - 5 1969
1-20 e GENERAL ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPEC. NO. 2lAll12 REV. NO. 6 PURCHASE SP ECI FICA TIOH S ... NO. 18 CONT ON S"EET 19 8.16 Alternate t-!aterials The Seller shall be free to suggest alternate materials during prepar-ation of detailed drawings and shall bring such alternates to the attention of the Buyer, but shall not make substitutions without approval of the Buyer. Request shall include: 8.16.1 Re.. on for substitution. 8.16.2 Identification of the component or parts involved. 8.16.3 Either the complete material specification similar to AS1lf for each type and form of proposed material, or the information as follows: a) Type of Service (Structural, High/Low Pressure, Temperature, Weldable) b) Manufactured Form (Pipe, Plate, Tube, Bar, Bolting) c) Size, Thickness Limits d) Alloy Grades (C-Steel, Alloy Steel, Stainless Steel Designations) e) Steel-Making Process (Open Hearth, Basic Electric) f) Forming Process (Hot Forged, Hot/Cold Rolled, Drawn, Seamless Welded,' Cast) g) Heat Treatment, Stress Relief Parameters h) Type, Location and Number of Mechanical Tests (Tensile, Bend Homogeniety, Hydrostatic) i) Mechanical Property Acceptance Limits j) Chemical Composition Acceptance Limits k) Ins~ection Requirements such as: Radiography,- Liquid Penet rant, Magnetic particle, Ultrasonic Including Acceptance Limits.
- 1) Surface Finish Acceptance Limits 8.16.4 Allowable Stresses (If not an ASME Material) 8.16.5 For major pressure parts, additional information-will be required regarding details of previous applications of the material, impact strength, NDT temperature, micro-structure variations, creep, stress rupture, hardness, radiation damage, welding, forming, corrosion and temperature effects as applicable for engineering evalu~tion of the application and as required for code purposes.
ISSUEO: MAR - 5 1969
1-21 GENERALe ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION SPEC:. NO. 21A1112 "EV. NO. 6
- 50. "0. 19 C:ONT ON S"EET 20 9.0 FABRICATION 9.1 Procedures 9.1.1 The Seller shall submit for the Buyer's approval, all of the following procedures and procedure specifications:
9.1.1.1 Heat treatment procedures for all thermal processes exceeding BOO°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 farming and cladding. 9.1.1. 3 Welding and weld repair procedures including tempor'ary welda as required in accordance with the ASME Code, Section IX, Paragraphs Q-I0 and 11, and QN-IO and 11, Section III, Paragraph N-S40. 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 9.2.1 Stainless steel and carbon steel shall be cut to size or shaped by machining, shearing or thermal cutting .
- IJ'ofAR - a 1969
1-22 PURCHASE SPECIFICATION GENERAL e ELECTRIC ATOMIC POWER EQUIPMENT OEPARTMENT SPEC. NO. So. NO. 2lAlll2 20 CON,. REV. NO. ON SHEE" 6 21 9.2.2 Thermal cutting of stainl.ss steel shall be followed by the removal of approximately 1/32" depth from the cut surface. Thermal cutting of carbon steel shall be followed by the removal of oxides. 9.3 Welding 9.3.1 The reactor vessel base material pre-heat and interpass temperature shall be as specified in the welding procedures. but in no case leiS than lOO*F, except weld overlay pre-heat which shall be no less than 200*F. Pre-heat temperature shall be maintained after weldiag until start of post~weld heat treatment. Pre-heating techniques shall be such as to ensure that the full thickness of the weld joint preparation and adjacent base material is at the specified temperature for the dis tance of "T" or two inches. wh ichever 1. greater. where "T" is the material thickness. 9.3.2 When atainless ateel or nickel-chrOMium-iron alloy is welded to itself or to each other, no pre-heat is required. except when the heat-affected zone reaches ferritic base material as in the cases of welding to buttered nozzle ends or cladding. When the buttering or cladding is less than 1/4 inch thick. pre-heat to at least 200*r is required. followed by post-weld heat treatment except that sub-sequent welding to cladding greater than 1/8 inch thick may be done without preheat if the specific welding procedure is qualif~ed to show that the heat affected zone does not reach the base metal. 9.3.l All surfaces (to be welded) shall be free of cavities or protrUSions which may interfere with the welding procedure. 9.3.4 Pre-heat. welding and post-weld treatment shall be planned and con-ducted to minimize undue distortion or warping of the parts and preclude cracking. 9.3.5 Machined surfaces and threads shall be protected against weld splatter. 9.3.6 Stainless steel welds shall be cleaned with stainless steel wool or stainless steel brushes before adding the next bead and follow-ing the final bead to facilitate inspection. The light oxide dis-coloration which forms on the weld surface need not be removed. 9.3.7 Welds shall be cleaned of slag and flu:.. :.,\,,'*ween passes and following the final deposit.
""MAR - 5 1969
1-23 GENERAL fa ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT
EC. NO. 2lAll12 IOIEy. NO. E PURCHASE SPECIFICATION , .. NO. 21 CONT ON ."EET 22 9.3.8 Any cracks, blow holes, or other defects which appear on the surface of weld beads shall be removed by machining, chipping, grinding, or arc gouging. Austenitic weld repairs, if arc gouged shall be followed by grinding. Austenitic welds shall not be peened; ferritic welds may be peened under controlled conditions after approval by the Buyer.
9.3.9 Wide welds to overcome poor fit are not permissible. Poor fits shall be remedied by suitable means such as regrooving. and approved by the Buyer. Except for small cavities, the Seller shall not correct a plate edge de-ficiency unless approved by the Buyer. The Buyer may require radiography or other methods of examination of welds used to correct plate edge de-ficiencies. 9.3.10 Post-weld heat treatment temperature shall be l150°F +25°F -SO°F. Inter-stage post-weld heat treatment holding time shall be 15 minutes minimum. Final post-weld heat treatment holding time shall be one hour per inch of thickness, minimum. 9.4 Repair of Defects Repair procedures shall be prepared for the repair of all defects. Major de-fects shall require prior approval by the Buyer and may require witnessing by the Buyer's representative. Major repair is defined as (1) a repair .to material other than weld metal which requires an excavation greater than 3/8 inch deep or 10 percent of the wall thickness, whichever is less; (2) the repair of any*cracks, other than crater cracks, in any material or weld metal; and (3) the repair of any defect which is indicative of either a fundamental material problem or a process out of control. A minor repair is defined as all other repairs. 9.S Cleaning 9.5.1 Interior Surfaces After the Seller has completed all other work, the interior surfaces of the reactor vessel shall be thoroughly cleaned to be visibly free of lubricants, weld splatter, chips, embedded iron particles and other foreign materials. A preferred method for cleaning and rinsing is use of high pressure water blasting equipment for these operations given in Paragraph 10.8. To main-tain cleanliness of the mterior of the vessel and head during drying and sealing, the personnel required to enter the vessel or head should wear clean cloth shoe covers and clean clothes. The vessel shall be sealed to prevent entry *of dirt or foreign materials. Seals used on nozzle ends and flange faces shall not alter weld preparations or sealing surfaces. 9.5.2 Exterior Surfaces Exterior carbon steel surfaces shall be cleaned of oil and grease after which mill scale, rust scale, and other foreign matter shall be thoroughly removed by such means as sandblasting as specified by the Buyer. All sur-faces shall be brushed or air cleaned to remove all traces of sand or grit
- IS.UEO:
PlAt( - ~ 1~69
1-24 PURCHASE SPECIFICATION GENERAL e ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT S"ltc. NO. SH NO. 2lAll12 22 CONT "EV. "0. ON SHEET 6 23 10.0 *INSPECTION ~~D TEST 10.1 General The Seller shall submit for the Buyer's approval. the following inspection and test procedures: 10.1.1 Ultrasonic Examination Procedure for the followin~: 10.1.1.1 1. Forgings
- 2. Plate
*3. Welds
- 4. Weld build-ups
- 5. Cladding
- 6. Tubular Products 10.1.2 Magnetic Particle Examination Procedures for the Following:
10.1.2.1 Carbon steel & low alloy steel forgings 10.1.2.2 Carbon steel & low alloy steel welds 10.1.2.3 Weld build-ups 10.1.2.4 Bolting 10.1.2.5 Carbon steel and low alloy steel tubular products 10.1.2.6 Carbon steel and low alloy steel castings 10.1.2.7 Edge preparations of carbon steel and low alloy steel materials. 10.1.3 Liguirl Penetrant Examination Procedures for the Following: 10.1.3.1 Austenitic Forgings 10.1.3.2 Austenitic welds 10.1.3.3 Austenitic weld bu1ldu~ 10.1.3.4 Cladding 10.1.3.5 Austenitic tubular products 10.1.3.6 Austenitic castings 10.1.3.7 Edge preparations of austenitic materials 10.1.4 Radiographic examination procedures for welds. castings, for each type of radiographic source above and below 2 Mev. 10.1. 5 Hydrostatic Examination Procedures 10.1.6 Leak Check Procedures
- 1-25 GEN ERAL __ ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT
,PEC. NO. 2lAlll2 REV. No.6 PURCHASE SPECIFICATION , .. NO. 23. CONT ON 'MEET 24 10.1.7 Hethods, proces~.S and equipment to be used in .stablishin~
"a.-built" dimensions and alignment co"' . . . :. which are not generally used in a typical industrial sh~p.
10.2 Definitions 10.2.1 "As-Fabricated" Specimens "As-fabricated" specimens are mechanical test specimens taken from carbon and low alloy steel forgings and plates used in the vessel fabrication from each heat and heat treatment lot and from welds between base m4terial made by each weldin~ procedure used and in a thickness equal to or g~eater than the thickest weld made with each procedure. Coupons for "as-fabricated" specimens shall be taken from the forgings or plates following all hot working or forming and all heat treatment except post-weld heat treatment. These coupons shall then be subjected to a post-weld heat treatment equivalent to the treatments which the parts it represents will receive in the completed vessel. This shall consist of holding the coupon at the post-weld heat treatment temperature for a time equal to or greater than the longest accumulated time any part it repre-sents shall be at the post~weld heat treatment temperature. 10.2.2 "1/4T x r' Location The "1/4T x 1'" location of specimens is defined as a location within the material no closer than "1/4T" from one quenched surface, and no closer than "T" from any other quenched edge, where "T" is the nominal thickness of the material. 10.2.3 NIL-Ductility Transition (NDT) Temperature The nil-ductility transition (NDT) temperature is defined as the temperature at which a specimen is broken in a series of tests in which duplicate no-break performance occurs at a temperature 10°F higher, when tested in accordance with ASTM E208. 10.2.4 Impact-Transition Curve A curve representing breaking energy vs. temperature from at least twelve Type A Charpy-V specimens, tested in accordance with ASTM A370, except each specimen tested at a different temperature. The temperature range of testing shall establish the upper plateau, the transition region, and the lower plateau. Each plateau shall be determined by at least one, but not more than two points. The remain ng specimens shall be used to develop the transition region. The lower plateau need not be developed if it occurs below -80 r. o 10.2.S A "lot of material" consists of all material from one heat (one melt) in a heat treatment furnace.
"'UEe:
MAR - 5 1969
1-26 PURCHASE SPECIFICATION GEN ERAL e ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPite:. NO. S ... NO. 2lAll12 24 'ONT NEV. NO. ON S"EE," 6 25 10.3 Material Mechanical Tests 10.3.1 Mechanical Properties 10.3.1.1 Impact properties of all carbon and low alloy steel used in the main closure flanges and the shell and head materials connecting to these flanges shall meet the requirements of the ASME Code, Section III, Paragraph N-330 at a temperature no higher than 10°F. In addition, this material shall have an NOT temperature no higher than 10°F as determined per ASIM E208. 10.3.1. 2 Impact properties of all other "as-fabricated" carbon and low alloy steel pressure containing material and the vessel support skirt material shall meet the requirements of the ASME Code, Section III, N-330 at a temperature no higher than 40°F. In addition, this material shall have an NOT temperature no higher than 40°F as determined per ASIM E208. The actual NOT temperature of all material opposite the center of the active fuel of the core as indicated on Drawing 886D482 shall be determined. 10.3.1.3 Tensile test properties of all materials shall. ~e inspected and tested to meet the requirement of the applicable ASME Code or ASTM specification. 10.3.1.4 Test data shall be reported to the Buyer. 10.3.2 Required Number and Specimen Location The number and location of tensile and impact test specimens required shall be per ASME Code, Section III, N-3l3.2 and the following depending on the form of the material. The following tests may be integrated with the tests required by the ASME Code and ASTM Specification wherever possible. 10.3.2.1 ~essel Flange and Head Flange Forgings Tangential specimens, as-fabricated, shall be taken from locations per ASME Code,* Section III, N-313.2 (d) (2). A total of at least 2 tensile, 6 Charpy-V impact and 4 drop weight specimens shall be tested for each flange from which 1 tensile, 3 Charpy-V impact and 4 drop weight specimens shall be located approximately 180 0 from the other specimens. The materia~
.hall meet the requirements of Paragraph 10.3.1.
I"MAR -;) 19S;]
e 1-27 GENERAL ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SPECIFICATION S"EC. NO. 21A1112 REv. NO. 6 5" NO. 25 CONT ON . " EET 26 10.3.2.2 Lov-Alloy Steel Nozzle Forgings Specimens. as fabricated, shall be taken from locations per ASHE 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 10.3.1. 10.3.2.3 In addition to the tests required by the ASHE Boiler and Pressure Vessel Code, longitudinal specimens (parallel to the primary rolling direction), as-fabricated, shall be taken from the 1/4T 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 ASHE Code,. Section III. N-313.2 (d). Castings 1000 lb. weight and under shall have a total of 1 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 0 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 in the core area. The material shall meet the require-ments of paragraph 10.3.1 *
- ' 'MAR - 5 196CJ
1-28 PUl~CHASE SP ECI FICA TION GENERAL _ ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT S"EC. NO. 2lAl112 5.. NO. 26 REV. NO. 6 CONT ON 5"EE"'27 10.3.2.S Studa, Nuts. Bushings and Wsshers for Main Vessel Closure Hardness tests shall be made on all main vessel closure b~lting to demonstrate that heat treatment has been performed. Studs, nuts and bushings shall be hardness tested individually. One sample from each lot of washers shall be hardness tested. Impact tests required by ASME Code. Section III. paragraph N-330 shall meet the Code requirements at a temperature no higher than 10°F. In addition to the magnetic particle or liquid penetrant accep-tance standards specified in ASME Code. Section tIl. paragraph N-325, axial defects of less than thread depth shall be investi-gated to determine their nature. Any cracks or sharply defined linear indications are unacceptable. 10.4 Welded Base Material - Mechanical Tests 10.4.1 AS~~ Code Weld Test Plates The Seller shall prepare and test weld coupons of Category A and B joints in accordance with ASHE Code. Section III. N-713. The impact test temperatures shall be determined in accordance with paragraph 10.3.1 of this specification. In addition to the tas~ required by the Code. 6 d~op weight specimens shall be taken from the 1/4T x T location from these plates and, if different welding procedures are used. from plates for base material to base material welds of Category D joints as defined in ASME Code. Section III. N-461. Two each of the drop weight specimens shall represent the the base metal. heat affected zone and weld metal. The specimens shall meet the requirements of paragraph 10.3.1.2. Additional drop weight specimens shall be required in accordance with paragraph 10.3.1.2 if the welding procedure is to be applied 1n the area o~posite the core. 10.4.2 One of the test plates of Category A or B required in 10.4.1 above shall be selected by the Buyer for rhe fabrication tests required in Attachment B. Paragraph 2. The Seller shall perform all required tests and reports. These te~ts are for information only, but time is of the essence and the tests should be performed and results reported as early as practical. 10.4.3 The Seller shall prepare and ship. but not test. Surveillance Test Program material and specimens in accordance with Attachment B, Paragrapb 3. 10.4.4 Flange Forging Weld Test Plate In the event the vessel and head flanges are made by welding two or more forged segments, the Seller shall prepare a weld test plate from the forging material. Impact and tensile specimens I'S~~' _ 0 1969
1-29 GENERALe ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPEC. NO. 2lAll12 REV. NO. 6 PURCHASE SPECIFICATION s~ .... 0. 27 CONT ON SWEET 28 ahall be prepared and tested. The specimens shall be pr\.:I"~_ ".~ from material in the weld-heat-affected zone and from the weld metal. Test results shall meet the requirements of paragraph 10.3.1. 10.5 Ultrasonic Inspection 10.5.1 Ultrasonic inspection of plate and forged material shall be per-formed in accordance with ASHE Code, Section III, except that ASHE Case Interpretation 1338-2, Alternate 2 shall not be accep-table, and the plate material testing shall be a 100 percent volu-metric inspection and shall be performed after forming and heat treatment. The following acceptance criteria shall apply in addition to Code requirements. A defect which causes any echo indication that exceeds 50 per cent of the indication from the calibration standard and that is continuous during movement of, the transducer more than 3 inches in any direction shall be unacceptable. A chart shall be maintained of defects with 50 per cent or greater loss of back reflection. 10.5.2 Prior to connecting any attachment, support or bracket, except insul-ation and thermocouple brackets, to the interior or exterior of plate portions of the vessel by means other than'groove welds 'below the pla*te, the plate shall be ultrasonically inspected. The plate shall be inspected to a depth at least equal to the thickness of the part being joined, and over the entire area of the subsequent connection plus a band all around this area of width equal to half the thickness of the part being joined. The inspection shall be in accordance with ASHE Code, Section III, Paragraph N-321, using longitudinal ~ave technique. The surface shall be 100 per cent inspected with the transverse interval being no greater than 90 per cent of the crystal diameter. 10.5.2.1 Reference Standard The Sell~r shall prepare a reference standard which consists of a flat bottom hole having a diameter equal to one-quarter of the thickness of the part being joined or 1/4 inch diameter whichever is greater. The bottom of the hole shall be one thickness of the part being joined below the plate surface. This reference standard shall be used for calibration purposes. 10.5.2.2 Acceptance Standards Any d~f~ct which produces a trace line pattern equal to or in exc~~~ of the app,opri3t~ r~fer~nc~ standard shall be unac~ep tablt>.
- ' . ' - 5 1969
e 1-30 GENERAL ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT SPEC. NO. 2lAl1l2 RE*'. NO
- 6
... URCHASE SPECIFICATION S .. NO. 28 CONT ON EET 29 10.5.3 Th. . .iD cloaure stud, nut, bushinl and wa.her material shall be ultrasonically teated follovina h.. t treaement aad rough machining to l2~ rm. or better finiah using both longitudinal and shear wave techniques. Longitudinal wave examination shall be performed on 100% of the cylindrical surface, and in additi~n on stud material from both ends of each stud. The longitudinal wave transducer shall have a maximum diameter of 1/2 inch. Shear wave examination shall
- b. performed on 100% of the outer cylindrical surface in both axial and circumferential directions.
10.5.3.1 Reference Standards The Seller shall prepare a reference standard of the same material, thickness and curvature as the part being examined. The reference standard shall contain calibration features as follows:
- 1) Longitudinal Wave-Radial Scan: 1/2 inch diameter flat-bottom hole having a depth equal to 10% of the material thickness.
- 2) longitUdinal Wave-En~ Scan: Flat-bottom hole with ar.ea equal /'
to 1% of stud cross-section or 1/4 inch diameter, whichever is smaller, having a depth of 1/2 inch.
- 3) Shea~ Wave: Square bottomed notches 1 inch long and 3% of the part thickness in depth, both ax~al and circumferential.
10.5.3.2 Acceptance Standards Any defect which produces a trac~ line pattern (echo indication) greater than the indication f~om the applicable calibration feature shall be unacceptable. A distance-amplitude curve may be used for the lonaitudinal wave examination. The curve may be a line established by plugaing the hole and examining it from both sides of the material. For end examination of studs the curve may be established for half the stud length and applied to an examination from each end to the center. ISSUEO: MAR - ~ 1969
1-31 GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT
'PEC. NO. 21A1ll2 ""Y. NO. 6 PURCHASE SPECIFICATION s .. NO. 29 CONT ON SHEET jQ
- 10. (, Claddin .
10.6.1 Ultrasonic Insp~ction - Cladding General 10.6.1.1 The cladding bond shall be "tested with the transducer on the clad side using a suitable couplant. The entire clad surface shall be inepected at intervals 1.4 times the base material thickness. but not greater than 12 inches. transverse to the direction of welding. 10.6.1.2 Reference Standard Th. Seller shall prepare a reference standard which consists of a flat bottom groove tn typical clad plate. The Iroove shall
- b. 0.35 tnch eaxt.ua width by ~t leaat one crystal diameter lOftg. parallel to the direction of welding. The groove shall be foreed by machining the baae .etal within 1/32" of the cladding interface and etched with nitric acid to remove excess ferritic material frOll the interface. This reference standard ahall be used for calibration purposes
- 10.6.1.3 Acceptance Standards Cladding which produces a trace line pattern equal to or in axc.as of the appropriate Reference Standard shall be unacceptable if a continuous pattern occurs during IIOvement of the transducer acre than three inches in any direction or if one or more patterns occur during IIOveeent of the transqucer les8 than" one inch in any direction from the boundary of anyone pattern.
10.6.2 Liquid Penetrant Inspection - Cladding General 10.6.2.1 All clad areas and clad repairs shall be liquid penetrant inspected per ASHE Code, Section Ill, N-627. The following indications shall constitute unacceptable defects and sh~ll be repaired.
+0.6.2.2 Any crack-like indications or incomplete fusion.
10.b.2.3 Linearly-disposed spot indications of 4 or more spots spaced 1/4 inch or less from edge to edge
---. 0;
~he indication.
10.6.2.4 Spot indications which are indicative of defects greater than 1/32 inch deep as revealed by bleed-out . I"j;iAYi - 5 1969 I
1-32 GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT OEPARTMENT S"EC. NO. 21Al1l2 PURCHASE SPECIFICATION S ... NO. 3~ CONT ON SHEET 31 10.6.3 Ultra.onic Inspection - Cladding Special Areas 10.6.3.1 The flange seal surfaces shall be inspected for bond to the flanges as per 10.6.1 except that the inspectlon shall be'over 100 per cent of the area. Prior to final machining the volume 1/8 inch above and below the surfaces on which the double seals will seat shall be 100 percent inspected for defect using longi-tudinal wave technique. The acceptance criteria shall be that any defect which produces a trace line pattern equal to or in excess of a 1/16 inch flat bottom hole ~ay be unacceptable. 10.6.3.2 The final machined surfaces on which the double seals seat shall be inspected by surface w~ve technique. Any defect producing a signal greater than the signal produced by the 0.002 inch deep by 1/8 inch long spark machined groove in a reference standard which the Seller shall furnish may be cause for rejection. 10.6.4 Liquid P~netrant Inspection - Cladding SpeCial Areas 10.6.4.1 The area of the flange seal surfaces on which the double seals seat shall be liquid penetrant inspected per ASHE Code, Section III, N-627, except that any indication of any type shall be unacceptable. 10.6.5 Magnetic Particle Inspection - Plate Material 10.6.5.1 Both internal and external surfaces of all low alloy steel plate material shall be magnetic particle inspected per AS~E Code, Section III. Paragraph N-626 following forming and heat tre~tment. The acceptance standard of ASHE Code, Section Ill, Paragraph N-625.S shall apply. 10.6.6 Openings in Pressure Parts 10.6.6.1 The entire surface of all openings for partial penetration nozzles, regardless of size, except for the seal leak detection connection, shall be examined in accordance with ASME Code, Section III, N-S13. 10.6.6.2 The entire surface of the finished stud holes in the head flange and the holes in the vessel flange prior to tapping shall be examined by the methods of ASHE Code, Section III, N-513. Any indication of cracks or linear indications shall be reported to the Buyer for information. Any crack or linear indication may be subject to removal and "~rlir i f required. 10.7 Welds 10.7.1 Radiographs 10.7.1.1 Gamma rays shall not be used unless approved by the Buyer. 10.7.1.2 Films shall be suitably marked to identify the weld. Film identi-fication markings shall coincide with the detail drawing markings f or each weld.
'i=lAH' - 5 1969
e ELECTRIC 1-33 GEM ERll ATOMIC POWER EQUIPMENT DEPARTMENT PURCHASE SP ECI FICATION
'P&C:. NO. 21A1112 REV. NO. 6
... NO. 31 C:ONT ON 'HEET 32 10.S Hydrostatic Tests 10.S.1 Code Test Immediately prior to hydrostatic testing, all interior surfaces of the vessel and head that will contact water during hydrostatic testing shall be thoroughly cleaned. Cleaning and degreasing*shall be by the use of high pressure (greater than 5000 psi) deionized water containing 500 ppm by weight of TSP for water blasting all internal surfaces. These sur-faces shall be subsequently water blasted with deionized water (no ad-ditives). The vessel shall be filled with deionized water for hydro-static testing. The method of heating the vessel is subject to approval by the Seller.
Defini tions : Deionized water Conductivity 2 micro-mho/cm Solids 10 ppm max Chlorides 1 ppm max Fluorides 1 ppm max Sulfides 1 ppm max TSP - Reagent grade per American Chemical Society Specification for tri-sodium-phosphate CAUTION: Special care shall be taken to thoroughly water blast rinse with deion-ized water crevice areas such as between the botto~ head and stub tubes and behind welded-in thermal sleeves directly following cleaning with TSP solutions until effluent conductivity is less than 5 micro-mho/em. After completion of fabrication but prior to shipment, while the vessel is supported on its normal supports, the reactor vessel shall be pressure-tested in accordance with the ASHE Boiler and Pressure Vessel Code, Sec-tion III, Paragraph N-7l4. Reactor vessel material temperature shall be
*at least 100°F. In no case, however, shall the *water temperature be higher than 200°F. Suitable gasket material instead of metal "0" rings may be used for this test.
10.S.2 Second Hydrostatic Test Following the Code test, the vessel shall be hydrostatically tested at de-sign pressure with new metal "0" rings. This test shall demonstrate that the head seal meets the sealing requirements. Relative displacement and rotation of the head closure flanges during this test shall be measured in at least four places and reported to the Buyer. The measurements shall be made prior to stud tightening and at 250 psi intervals from zero psi to the design pressure. 10.9 The placing of the head, tightening the studs to operating bolt-up loads, un-bolting and removal of the head over the studs shall be demonstrated. The elapsed times for each step shall be recorded. 10.10 Final inspection after hydrostatic tes.t per ASHE Code, Section III, N-61S shall include seal surfaces and the nozzle weld preparation
- 1969
1-34 GENERAL. ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT S"EC:. "0. 2lAll12 ~EV . . . 0'. 6 PURCHASE SPECIFICATION SH NO. 32 C:ONT ON .... EE.,. 33,
- 11. a SSIPMDT 11.1 5. . 11 Parts Small t loose pieces, including bolting, tools, ga&kets, etc., shall be adequately crated or boxed for protection during shipment. Parts subject to rusting shall be suitably protected. All pieces shall be marked with the equipment piece number or mark specified by the Buyer.
11.2 Shipping Weishts and Dimensions Eatimated shipping weights and overall clearance dimensions of all major pi~ces to be shipped to the erection .ite shall be shown on the drawings when submitted to the Buyer for approvaL 11.3 Shipping Skids Shipping skids for composents shall be designed to support the coaponents adequately and securely during shipment to the erection site and to account for the means of movement lifting, and posi-tioning to be provided by the Seller at the erection site. 12.0 SUBMITTALS 12.1 Tabulation (For Information Only) Fabrication, qualification and inspection procedures, reports processes, and calculations are tabulated below (all of which require submittal to the Buyer in quantities as shown on Attachment A). This tabulation shall in no way be construed as being complete or limiting th~ do~uments necessary to meet the requirements of this specification. Heat treatment procedure Forming and bending procedure Welding and weld repair procedure specification Repair procedures Cleaning and preserving procedures Ferrite content or Ni/Cr ratio control procedure Ultrasonic examination procedure Magnetic particle examination procedure Liquid penetrant examination procedure Radiographic examination procedure Hydrostatic examination procedure Leak Check Procedure Measurement reports SUllllllary reports "As-built" dimensions and alignment checks procedures. ISSUEC: MAR - 5 1969
1-35 GENERAlO ELECTRIC ATOMIC PowER EQUIPMENT DEPARTMENT I I SPEC. NO. 2lAl1l2 REV. No.6 PURCHASE SPECIFICATION 33 S ........ 0. NT ON S""EET J4 12.1 (Continued) Design analysis calculations Material purchase specifications Material selections Thread Lubricant Specifications 12.2 The following shall be submitted in accordance with Attachment A: 12.2.1 Drawings 12.2.1.1 Outline Drawings - A drawing depicting the outline of the reactor vessel indicating over-all dimensions, location and size of nozzles, location of supports, shipping and operating weights. 12.2.1. 2 Assembly Drawings - A section drawing depicting the arrangement of the functional parts, parts list and material designation. 12.2.1.3 Detail Drawings - Drawings for details of construction such as weld preparations, surface finishes, finished dimensions, nozzles lifting attachments, insulation attachments, ;hermocouple pads, flanges and supports. 12.2.1.4 Drawings for Approval - Outline, assembly and detail drawings shall be submitted for approval. The detail drawings sub~itted shall be for design details enumerated in 12.2.1.3 which are required for coordination with piping and structure and design details which are at variance with the code or the requirements of this specification. 12.2.1. 5 Controlling Location Arrangement Drawings One or more drawings shall be devoted exclusively to outline dimensions such that mating components designed and supplied by others such as piping, anchor bolts, instruments, etc .. may be procuted for an exact fit with the reactor vessel asse~~ly. These drawings shall show reference to the controlling detuil drawings and show over-all dimensions and locat ions or: :-~;Inor vessel. 12.2.1.6 Drawings to be Certified - Outline, Assembly and Detail drawings for design coordination shall, upon .completion of the design. be certified to be correct with no further changes required. No alterations may be made to the design after certification without the approval .~~ the Buyer *.
1-36 GENERAL _ ELECTRIC ATOMIC PCWER EQUIPMENT DEPARTMENT S"e:c. NO 2lAll12 Re:V NO 6 PURCHASE 51) ECI FICA TION s ....... 0. 3.'5 CONT ON S"'EC~ F 12.2.2.5 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 "essel 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. - ultr.asonic, radiography and hydrostatic). A list of the records shall be submitted to the Buyer on co~pletion 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 Re ... iew or appro ... al of drawings, procedures, data, Or sloOecificalions by .he buyer wilh regord I:) general design and cOI.trolling oi ... en.ions does not c:onstitut .. acceptance of ony designs, mot.rlols. or equipment which will not full.! I the functionol or p .. ,formance r.Qui'~m **"s eSlab I i shed by .hi s speci Ii cation and the purchose contract. Doc ..... ent s and drawings sub ... itted shall be b lack Iine and of n Qual i.y whi ch will produce readable-prints when microfi Imrod (35 mm) and blown bock on a con ..... n"onal 18. 24 prin,er ... iewer sueh as F "mac 200 or Itee. Send 011 documents and drawings to L.L. Kleinhesselink, CE, APED, (wilh eopy of "ansml"al,o .l.PED Buyer, es ,"d,:a * .,> 4- 3O"IIIIN: 'HIS DlMEN510N MAI'<< rU#NI5NW j
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(4) Core diff.rential pre.,ure and liquid control (5) lee ire Inlet (6) lee ire Outlet (7) Steam Outlet CIlN CB&I is responsible for the complete analysis of :J"~ O\N (1) All DOzll ** not .pecifled above C'l (2) Stabilizer bracket. o ::11:1
- s 111 (3) lnaulatton bracket.
" ~
(4) Head lifting l.ugs (5) Shroud he~d and dryer guide support go (6) Feedwater sparger support (7) Any other internal vessel attachments ..., At4.p ",'
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TEMPERATURE TRAl, .ENTS No. Fluid Fluid Fluid State Vessel Part of Temp. Start End of Fluid Vessel Cvcl-- Ra
- - T
--- - T,
-- ... - Fluid
-- - Veloci - , P
-------- N------
Recirc. Outlet 200 100 F/hr 100 546 Water 25 ft/sec Saturated I 1000 F/hr 546 370 Water 25 ft/sec Saturated Followed by 200 100 F/hr 370 100 5 Step 546 (Step 546 (Step Water 5 ft/sec 1000 psig 26 seconds duration to 130) from '130) (ren~~T at 130°F Recirc. Inlet 200 100 F/hr 100 546 Water 32 ft/sec Saturated I Nozzle 200 a 546 546 Water 32 ft/sec Followed by a 90 90 Water 10 ft/sec 170 psig Steam Outlet 532 100 F/hr 100 546 Steam 5 ft/sec Saturated Condensing Steam Nozzle in nozzle I 531 100 F/hr 546 346 Steam 5 ft/sec Saturated Followed by I 1000 F/hr 346 296 Water 14 ft/sec Saturated Followed by 100 F/hr 296 100 Water 0 Saturated I I 1 1000 F/hr 546 370 Steam 25 ft/sec Saturated Followed by 100 F/hr 370 100 Steam - a Saturated I Water Feedwater a 376 376 Water 10 ft/sec 1100 pslg Steady State 546° Nozzle Water in Vessel 1~"0 100 F/hr 100 546 Water 0 1100 psig Followed by Step Tol 0 100 100 Water 5 ft/sec UOO psig Followed by Step To 250 F/hr 260 376 Water 5. ft/sec 1100 psig N Core Spray 250 100 F/hr 100 546 Water 0 Saturated E Nozzle N 3: 250 546 Water 0 1000 psig Followed by (Steam > 0 80 ** 80 Water 20 ft/sec o psig in Thermal Sleeve
- a ~
Annulus) < c.J1 ~ ~ Veloctty changes linearly 5 ft/sec to 20 ft/sec ~ ** l.,rater reaches this temperature in 15 seconds ATTACHMENT D .... I Sht. 'I Cont. on 2-. CJ1 o
- TEMPERATURE 'It{A
- ENTS
- No. Fluid Fluid Fluid State Vessel Part of Temp. Start End of Fluid Vessel Cvcl
- R
-- - - .-. - T- - - luid Veloci P ._----- N Jet Pump 200 1000 F/hr 546** 370 Water 0 Saturated Followed by Instrument 100 F/hr 370 100 0 Saturated Nozzles CRD Hydraulic 200 0 45° 45° Water 15 ft/sec 1000 psig Steady State 546° Return Nozz e Water in vessel 0
Core Diff & 10 0 80 80 Water 15 ft/sec 1000 psig Nozzle at 546°** Liquid Control *** *** *** Isothermal at Nozzle Start 200 1000 F/hr 546** 370 Water 0 Saturated Followed by 100 F/hr 370 100 Water 0 Saturated 2 Inch In- 200 100 F/hr 330 100 Water 0 Saturated Nozzle at 546° 180-strument thermal at Start Nozzle 200 1000 F/hr 546** 370 Water 0 Saturated Followed by I 100 F/hr 370 100 0 Saturated I Core Support 200 100 F/h.r 100 546 Water 5 ft/sec* Saturated Structure 199 100 F/hr 546 346 Water 5 ft/sec* Saturated Followed by 1000 F/hr 346 296 Water 5 ft/sec* Saturated Followed by 100 F/hr 296 100 Water 5 ft/sec* Saturated 1 1000 F/hr 546 370 Water 5 ft/sec* Saturated Followed by 100 F/hr 370 100 N Water is on all sides of Core Support structure and on inside surface of Reactor Vessel for I both the above transients. E
~
~
5 Step 546 546 Water See Recirc 1000 psig 26 seconds duration N l .. Step to 130 Step from 130 Outlet of l30°F I
~
~ ~
- Water velocity above the support plate, on the vertical surface of the shroud cylinders, and on the underside of the ~
support plate is essentially zero and natural convection heat transfer coefficient may be used. The 5 ft/sec velocity
~
c.Jl is directed against the vessel bottom head but by using natural convection heat transfer coefficients a conservative cD analysis should result. c:r>> U:I
** Water reaches this temperature at a fluid temperature rate of lOOF/hr.
*A* See 886D482, Sht. 1 for location of liquid control flow. .....
I Sht. 2 Cont. on ~ U1 ATTACHMENT D .....
TEMPERATURE ,TRM LENT~S " i No. Fluid Fluid Fluid State Vessel Part of Temp. Start End of Fluid Vessel Rat T . Fluid P N Cvcl- - - - T- - - --- . ----- Veloc1 - -- - - -- - I I, Closure Flanges 200 100 F/hr 100 546 Steam 0 Saturated Condensing Steam I
& Adjacent Shell Heat Transfer and Refueling 191} 100 F/hr 546 350 Steam ,Saturated Followed by Bellows Support Flooding with water at 330*F Free Followed by Skirt 100 F/hr 300 150 Water Conv. Saturated r- '
1 1000 F/hr 546 375 Steam Free Saturated Followed by 100 F/hr 375 100 Steam Conv. Saturated Bottom Head'& 200 100 F/hr 100 546 Water 5 ft/sec Saturated Support Skirt 199 100 F/hr 546 375 Water 5 ft/sec Saturated Followed by 300 F/hr 375 330 Water 5 ft/sec Saturated Followed by 100 F/hr 330 100 Water 5 ft/sec Saturated
, 1 1000 F/hr 546 370 Water 5 ft/sec Saturated Followed by 100 F/hr 370 100 Water 5 ft/sec Saturated Control Rod I
370 0 50* 50* Water
- 1000 psig Penetration As-Drive Penetra- sembly at 546*
*Heat transfer coetficients through a thermal 10 ft/sec
,tion Periphera at Start sleeve within the housing are: outside Location and assembly Central Loca- (a) h = 75 Btu/hr ft 2 *F above stub ,tube tion (b) h = 193 Btu/hr ft 2 °F at stub tube (c) h = 40 Btu/hr ft 2 °F below stub tube NOTE: For the purposes of demonstrating for other parts of the vessel applicable exception from Detailed Stress Analysis according to Paragraphs N-4l5.l an~ N-45l of the ASHE Code Section III, the following values may N be used.
~
I-' (a) Total design pressure cycles from atmospheric pressure to operating pressure and back to I-' N atmospheric pressure is 200 cycles. ~ (b) TIle number of significant pressure fluctuations (200 psi full range) during normal opera- :;u ~ fD tion is 280 . <: .:J' (c) The number of major temperature fluctuations is 400 I-' cD al -to ATTACIINENT D Sht ) Cont. on Final ..... I U1 t-:)
1-53 GENERAL e ATOMIC POWER EQUIPMENT DEPARTMENT ELECTRIC SFECIAL PROJECT MONTICELLO SPEC. N021A1112 REV. No.1 I PURCHASE SPECIFICATION ad s .... 0. 1* CO .. T 0" S"EET F T1T~E REACTOR PRESSURE VESSEL ATTACHMENT E CERTIFICATION OF DESIGN SPECIFICATION AS TO COMPLIANCE WITH THE REQUIREMENT OF THE ASHE BOILER AND PRESSURE VESSEL CODE SECTION III NUCLEAR VESSELS This Specification 21A1112, Rev. 6, lists for the Monticello Nuclear Power Station the purchase specification, specification control drawings, and supplementary speci-
'fications which comprise the Design Specifications required by Paragraph N-14l of the ASHE Boiler and Pressure Vessel Code, Section III, Nuclear Vessels.
This certification is issued in order that design and fabrication of the reactor pressure vessels identified by General Electric Company Purchase Order 205-55582-1 may proceed in accordance with the requirements of Section III. Pursuant to Paragraph N-140 of Section III, this certification is solely for the purpose of complying with the requirements of ASHE Boiler and Pressure Vessel Code, Section III, and: is not to be construed as involving, modifying or changing con-tractual relations or legal liabilities
- CERTIFIED BY 03 l. zR ~
Registered Professional Engineer DATE STATE ____~C~a~l~i~f~o~rn~i~a~____________ BRANCH Mechanical NO. 13540 REFERENCE DRAWINGS AND DOCUMENTS NUMBER DESCRIPTION 886D482, Rev. ** Reactor Vessel Specification Control 8850911, Rev. 2 Vessel Flange Bolting 107C5305, Rev. 2 Nozzle End Preparation 21A982l, Rev. 0 Stud Tensioners 117B1549, Rev. 2 Charpy Impact Specimen ll7B1550, Rev. 2 1/4" Tensile Test Specimen 21Al050, Rev. 0 Reactor Servicing Tools Attachment B, Rev. 3 Material Test and Test Specimens Attachment 0, Rev. 1 Temperature Transients 731E678, Sht. 1, Rev
- 0 Vessel As-Built Dimensions 731E678, Sht. 2, Rev. 0 Vessel As-Built
,r Dimensions Sht. II Rev. II Sht.
.::.:.:.::..:.....;:.11 Rev. II 1 5 5
- ~
* ;t 2 10. 6 3 7 4 8 3 M~R I*" DR HEISING JA MAST
MONTICELLO 2-1 EXHIBIT 2 MANUFACTURER'S DATA REPORT AND VESSEL CERTIFICATION
- REV 4 12/85
2-2 e\
- r PRESSURE VESSEL REPORT MANUFACI'URER I S DATA REPORT AND VESSEL CERTIFICATION e)
L ..J MONTICEllO GENERAL: EL.ECTRIC CO. APEo- SP-N ~OSE VPF~ / 111-..J.5"r- I EPJ: .:1-1-1 - \,
.1-I
2-3
- PRESSURE VESSEL RECORD MANUFACTURER'S DATA REPORT AND CERTIFICATIONS BOILING WATER NUCLEAR REACTOR VESSEL 17.167' x 63.167' INS. HOS.
MANUFACTURER'S SERIAL NO. B-4697. MONTICELLO PROJECT, MONTICELLO, MINN. G.E. CO. P.O. 205-55582-1 CB&I CONTRACT 9-5624 TABLE OF CONTENTS
- 1. Manufacturer's Data Report
- 2. Hydrostatic Testing Certification
- 3. Radiographic Testing Certification 4* Ultrasonic Testing Certification
- 5. Liquid Penetrant Testing Certification
- 6. Magnetic Particle Testing Certification
- 7. Final Cleaning Certification
- 8. Welding Certification
- 9. Weld Repair Certification
- 10. Heat Treatment Certification
- 11. Weld Rod and Wire Certification
- 12. Cladding Carbon Content Certification
- 13. Welder Qualifications Certification
- 14. Parkerizing Certification 15 .. Material Certification 160 Material Identification
2-4
- 17. Nameplate Photograph
- 18. Fabrication Test Program Certification
- 19. Results of Tensile Tests per Par. 2.2.2 Attachment B
- 20. Results of Charpy V-Notch Impact Tests per Par. 2.2.4 Attachment B
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2-9
2-10 SUBJECT - ASME Code, Section III Hydrostatic Test PROJECT - Mpnticello Nuclear Reactor 9-5624 Type of Vessel - Boiling Water Nuclear Reactor Vessel Material - A533 Class I Grade Firebox Quality Vessel Height - 46'-0" + 1 1/2" T.L. Vessel Diameter - .17'-2-3/8" Base Metal I.D. Shell Thickness - 5 1/16" Min. Design Pressure - 1250 psig at bottom of vessel Design Temperature - 575 0 F Hydrostatic Test Pressure - 1563 psig at bottom of vessel Design Code - ASME Code, Section III, 1965 edition including Summer 1966 addenda, and the following additions or exceptions: A.--Details governed by analysis only.
- 1. Main closure flange configuration.
- 2. Configuration of the skirt attachment knuckle.
B.--Application of Code Revisions not covered by the Summer 1966 addenda.
- 1. ASME SA533 plate material Summer 1967 addenda.
- 2. Inconel material per Summer 1967 addenda
- 3. Main closure flange material per code case 1332-3.
- 4. Studs and nuts material for main closure flanges per code case 1335-2.
- 5. Main closure flange stud shank transition radius per code case 1366.
- 6. Bearing stresses for stabilizer brackets.per Winter 1967 addenda.
- 7. Coefficients of Thermal Expansion per Winter 1967 addenda.
- 8. Magnetic particle and liquid penetrant examination per Winter 1966 addenda.
Design Specificationa - Certified G.E. Specification No. 21Alll2, Rev. 5 Vessel Manufactured by - Chicago Bridge , Iron Company Vessel Manufactured for - General Electric Co~any The above vessel was hydrostatically tested according to the rules of ASHE Coda, Section III, Paragraph N-7l2. No detectable defects were found. The ves.el was built and* inspected according to the rules of ASHE Code Section III (see above), and complies with the manufacturer's drawings. The vessel is completed except for the certification of the stress report and completion of the final ASME Code, Section III inspections. Siqned --~1M~~~"~~~--------- Inspector DATE ( e
2-11 P~ECT Monticello Nuclear Reactor 9-5624 SUBJECT LEAKAGE TEST The leakaqe test of the qaskets in the vessel .head closure flanqes was performed as specified in p~cedure OHT-l, Rev.!!. The test was performed in conjuction with the hydrostatic test at design pressure per cus-tomer's specification, 2lAln2, Rev. ~ Paragraph 10.8.2.
. ~S No .iqnificant leakaqe was found in the inner or outer g** kets of the vessel closure flanges .
- Siqned DATE
2-12 CHICAGO BRIDGE & IRON COMPANY P. O. BOX '330B, MEMPHIS, TENNESSEE 38"3 90' 947-3'" DATE: MARCH 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' ID x 63.167 INS. HDS. MANUFACTURER'S SERIAL NO. B-4697 MONTICELLO PROJECT, MONTICELLO, MINN. GENERAL ELECTRIC CO. P.O. 205-55582-I CB&I CONTRACT 9-5624
SUBJECT:
HYDROSTATIC TESTING CERTIFICATION TO WHOM IT MAY CONCERN: This is to certify that the above referenced vessel was hydrostatically tested in accordance with the ASME Code, Section III, 1965 Edition, with addenda through Summer 1966 and General Electric Co. Specification 21All12, Rev. 5, Paragraph 10.8 and also approved" CB&I procedures CHT-l Revision I and DHT I Revision 3. (included in Monticello Project Manual Vc-lume II) *. Furthermore, no significant leakage ,was detected in the. inner or outer gaskets of the vessel and closure head flanges. See attachments. CHICAGO BRIDGE & IRON COMP&~Y
~~dhA'UtA?V E *E
- VARNUM NUCLEAR QUALITY ASSURANCE MANAGER
2-13
- CHICAGO BRIDGE & IRON COMPANY P. O. BOX '3308, MEMPHIS, TENNESSEE 38"3 DATE: March 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' x 63.167' INS. HDS. MANUFACTURER'S SERIAL NO. B-4697 MONTICELLO PROJECT, MONTICELLO, MINN. GE CO. P.O. 205-55582-I CB&I CONTRACT 9-5624
SUBJECT:
RADIOGRAPHIC TESTING CERTIFICATION
- TO WHOM IT MAY CONCERN:
This i~ to certify that radiograp~ic_examinations for the above referenced vessel were performed in accordance wi th t."e ASME Code, Section III, 1965 Edition, with Addenda t."rough Summer 1966 and General Electric Co. Specification 21A1112, Revision 5, Paragraph 10.7 and also approved CB&I Co. and/or suppliers Procedures RTP-l Rev.2, and advanced products radio-graphic procedure of finished welds, Rev. 2 (above procedures included in Monticello Project Manual Volume II). CHICAGO BRIDGE & IRON COMP~~Y
/,'/..//
VVl'V tfl.'Vrt,t?-,)lV E
- E. VAEl.~UM NUCLEAR QUALITY ASSURANCE
~~AGER
2-14 CHICAGO BRIDGE & IRON COMPANY P.O. BOX , 3 3 0 a, M EM PH IS, ,. E NN E SSE E 3 a, 1 3 90' 947-3'" DATE: MARCH 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' x 63.167' INS. HDS. MANUFACTURER'S SERIAL NO. B-4697 MONTICELLO PROJECT, MONTICELLO,MINNESOTA G.E. Co. P.O. 205-55582-I CB&I CONTRACT 9-5624
SUBJECT:
ULTRASONIC TESTING CERTIFICATION TO WHOM IT MAY CONCERN: This is to certifv that ultrasonic examinations for th~ above referenced vessel were performed in accordance with the ASME Code, Section III, 1965 Edition, with Addenda through Summer 1966, and General Electric Co. Specification 2lAll12 Rev. 5, Paragraphs 10.5 and 10.6, and also approved CB&I Co. and/or suppliers procedures HT-IOl Rev. 2, LE-2 Rev. a, QCP~89 w/addendum Rev.O, TT-2 Rev.l, UTP-l Rev.3, UTP-2 Rev.l (same as 9Q-63 Rev.l), UTP-3 Rev.O, UTP-4 Rev.O, UTP-5 Rev.l, UTP-6 Rev.l, UTP-7 Rev.l, UTP-8 Rev.O, UTP-IO Rev.O and UT-7l8777 Rev. a (above procedures included in Monticello Project Manual Volume II)
- CHICAGO BRIDGE & IRON CO~~&~Y
~;:;V:t"-r..tv-n'L E.E. VARNUM NUCLEAR QUALITY ASSURA.~CE ~~~AGER
2-15
- CHICAGO BRIDGE & IRON COMPANY P. O. 60X 13306, MEMPHIS, TENNESSEE 36113 901 947-3'"
DATE: March 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' x 63.167' INS.HDS. M&~UFACTURER'S SERIAL NO. B-4697 MONTICELLO PROJECT, MONTICELLO, MINNESOTA GE CO. P.O. 205-55582-I CB&I CONTRACT 9-5624
SUBJECT:
LIQUID PENETRANT TESTING CERTIFICATION I'. TO WHOM IT MAY CONCERN: This is to certify that liquid ~enetrant examinations for the above referenced vessel were performed in accordance with the ASME Code, Section III, 1965 Edition, with Addenda through Summer 1966, and General Electric Co. Specification 21All12 Rev. 5, Paragraph 10.6, and also approved CB&I Co. and/or suppliers procedures LE-4 Rev.O, PTP-l Rev.3, PT-7l877 Rev.O, advanced products liquid penetrant procedure Rev.l and TT-4 Rev.l (above procedures included in Monticello Project Manual Volume II). CHICAGO BRIDGE & IRON COMPANY
?{1!a,""~~(.t-rn-EoEo VARNUM NUCLEAR QUALITY ASSUR&~CE ~~AGER
2-16 CHICAGO BRIDGE & IRON COMPANY P. O. sox 13306. MEMPHIS. TENNESSEE 36"3 90' 947-3'" DATE: March 11, 1969
REFERENCE:
BOILING WATE.R NUCLEAR REACTOR VESSEL 17.167' x 63.167' INS. HDS. MANUFACTURER'S SERIAL NO. B-4697 l-lONTICELLO PROJECT, MONTICELLO, MINN. G.E. CO. P.O. 205-55582-1 CB&I CONTRACT 9-5624
SUBJECT:
MAGNETIC PARTICLE CERTIFICATION TO WHOM IT MAY CONCERN: This is to certify that maqnetic particle examinations for the above referenced vessel were performed in accordance with the ASME Code, Section III, 1965 Edition, wit~ Addenda through Summer 1966, and General Electric Co. Specification 2lAl1l2 Rev. 5, Paragraph 10.6 and also approved CB&I Co. and/or supplier's procedures LE-3 Rev. 0, MTP-1 Rev. 4, MTP-2 Rev. 1 (also known as Ladish l43-M), NDT-M-1 Rev. 0 and TT-3 Rev. 1 (above procedures included in Monticello Project Manual Volume II). CHICAGO BRIDGE & IRON COMP~~
~/U)NvY'~
E. E. VARNUM NUCLEAR QUALITY ASSUR&~CE ~~AGER
2-17 '. CHICAGO BRIDGE & IRON COMPANY P. O. BOX 13308, MEMPHIS, TENNESSEE 38113 901 947-3"1 DATE: March 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' x 63.167' INS. HDS. MANUFACTURER'S SERIAL NO. B-4697 MONTICELLO PROJECT, MONTICELLO, MINN. G.E. CO. P.O. 20S-55582-1 CB&I CONTRACT 9-5624
SUBJECT:
FINAL CLEANING CERTIFICATION TO WHOM IT MAY CONCERN: This is to certify that the final cleaning of the ,above referenced vessel was performed in accordance with the ASHE Code, Section III, 1965 Edition, with Addenda through S~ner 1966, and General Electric Co. Specification 21All12 Rev.S Paragraph 10.8 and also approved CB&I Co. Procedure CP-4 Rev.l (above procedure included in Monticello Project Manual Volume II). CHICAGO BRIDGE & IRON COMP&~Y
$tf!laMV~
E.E. VARNUM NUCLEAR QUALITY ASSURANCE ~~AGER
2-18 CHICAGO BRIDGE & IRO:K COMPANY P. O. BOX 13308, MEMPHIS, TENNESSEE 38113 901947-3111 DATE: March 11, 1969
REFERENCE:
BOILI:-JG \'lATER NUCLEAR REP.CTOR VESSEL 17.167' x 63.167' INS. HDS.
- 1Ai'\;UFACTU~F. 'S SERIA:i:. KO. B-4697
- 'lONTICI:LLO PROJECT, ~10NTICELLO, ~.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 the AS)lE Coc.e, Section III, 1965 Edition, with Addenc.a through S~~~er 1966, and General Electric Co. Specification 21All12 Rev. 5, Paragraph 9.3 and also approved CB&I Co. procedures \4J"'PS-1 Rev. 2 WPS-18 Rev. 4 WPS-35 ~ev. _ ~'lPS- 2 Rev. 2 WPS-19 Rev. 3 lt1PS-36 ~ev. 0 WPS-3 Rev. a WPS-20 Rev. a WPS-37 ~ev. 1 WPS-4 Rev. 2 WPS-21 Rev. 1 HPS-38 Re~J'. 1 WPS-5 Rev. a ~'lPS- 22 Rev. 1 11PS-39 Re\'. 0 WPS-6 Rev. 2 ~vPS-23 Rev. 1 ~'I"PS- 4a Rev. 1 ~.;rPS-7 Rev. 2 WPS-24 Rev. 4 loJPS-41 Rev. Q w"'PS-8 Rev. a WPS-25 Rev. a WPS-42 Rev. - WPS-9 Rev. a 'VI.TPS-26 Rev. a WPS-43 P.ev. 0 WPS-10 *Rev. a WPS-27 Rev. 2 w"PS-44 Rev. 3 \-iTPS-11 Rev. a WPS-28 Rev. 5 1;vPS-45 Rev. a WPS-12 Rev. 2 WPS-29 Rev. 1 vJPS-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. a a
.0' WPS-15 Rev. WPS-32 Rev. 1 WPS.-49 ReV. 2 vlPS-16 Rev. 2 WPS-33 Rev. a vlPS- 5 a Rev. 2 WPS-17 Rev. 1 WPS-34 Rev. 1 WPS-51 Rev. a
2-19 CHICAGO BRIDGE & IRON COMPANY
SUBJECT:
WELDING CERTIFICATION WPS-52 Rev. 1 WPS-64 Rev. 0 WPS-53 Rev *. 0 WPS-66 Rev. 0 WPS-55 Rev. 0 WPS-68 Rev. 1 WPS-56 Rev .. 0 WPS-69 Rev. 0 W7S-58 Rev. 0 WPS-73 Rev. 1 WPS-59 Rev. 0 WPS-74 Rev. 1 WPS-60 Rev. 0 WPS-75 Rev. 0 WPS-63 Rev. 0 WPS-77 Rev. 0 (above procedures included in Monticello Project Manual Volume II) * (. CHICAGO BRIDGE & IRON CO~P~~Y v4-P;t..~l.-~f/J?l../ E. E. VA&~UH NUCLEAR QUALITY ASSU~~CE Y~~AGER
2-20 CHICAGO BRIDGE & IRON COMPANY P. O. BOX 13308. MEMPHIS. TENNESSEE 39113 DATE: March 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' x 63.167' INS. HDS.
~~UFACTURER'S SERIAL NO. B-4697 MONTICELLO PROJECT, MONTICELLO, MINN.
G.E. CO. P.O. 205-55582-I CB&I CONTRACT 9-5624
SUBJECT:
_ WELD REPAIR CERTIFICATION TO WHOM IT MAY CONCERN: This is to certify that any \~eld repair done on the above referenced vessel was performed in accordance with the ASME Code, Section III, 1965 Edition, with Addenda through Sununer 1966, and General Electric Co. Specification21All12 Rev. 5, Paragraphs 9.3 and 9.4 and also approved CB&I Co. procedures GRP-l Rev. a, GRP-2 Rev. 1, GRP-3 Rev. 3, GRP-4 Rev. a, GRP-5 Rev. 3, GRP-6 Rev. 0 and GRP-7 Rev. a (above procedures included in Monticello Project Manual Volume II). CHICAGO BRIDGE & IRON CO~~A.~Y
/ ..
t* f;., !!;t.1---,-at. 4 'L. E. E
- VARl.'JUM NUCLEAR QUALITY ASSURA.~CE l-lANAGER
2-21
- CHICAGO BRIDGE & IRON CO?v:PANY P. O. 90X '3309, MEMPHIS, TENNESSEE 39"3 90' 947-3'"
DATE: March 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.1671 x 63.167 1 INS. HDS. MANUFACTURER1S SERIAL NO. 8-4697 MONTICELLO PROJECT, MONTICELLO, MINN. G.E. CO. P. O. 205-55582-1
- C8&I CONTRACT 9-5624
SUBJECT:
HEAT TREATMENT CERTIFICATION TO WHOM IT MAY CONCERN: This is to certify that the heat treatment performed on the above referenced vessel was done in accordance with the ASME Code, Section III, 1965 Edition, with Addenda through Summer, 1966; General Electric Company Specification 21All12 Revision 5, Paragraphs 8.0 and 9.0 and approved CB&I Company Procedures HTP-l Revision 1, HTP-2 Revision 1, HTP-3 Revision 1, HTP-4 Revision 1 and HTP-5 Revision O. Performance of heat treatment of material by suppliers is certified in the mill test reports. This work wa~ performed in accordance with one or more of the following procedures: HLA-1 Revision 1, HT-71S777 Revision 0, LE-l Addendum 1 Revision 0, LE-S Revision 3, LE-6 Revision 0, LE-7 Revision 1, LE-8 Revision 1, LE-9 Revision 1, LE-12 Revision 0, LS-1 Revision 2, "LS-2 Revision 0, TS-l Revision 0, TS-2 Revision 0, TS-3 Revision 0, TT-l Revision 1, TT-5 Revision 1, TT-6 Revision 1, CA-l Revision 1 and CA-2 Revision 1 (all of above C8&I Company and suppliers proceduTes included in Monticello Project Manual Volume II). CHICAGO BRIDGE & IRON COMPANY 6~!k7u~nt/ E. E. VARNUM NUCLEAR QUALITY ASSURANCE MANAGER
2-22 CHICAGO BRIDGE & IRON COMPANY P. O. BOX 13309. MEMPHIS. TENNESSEE 39"3 901 947-3'" DATE: March 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' ID x 63.167' INS. HDS. MANUFACTURER'S SERIAL NO. B-4697 HONTICELLO PROJECT, MONTICELLO, ~lINN. GENERAL ELECTRIC cq. P.O. 205-555S2-1 CB&I CONTRACT 9-5624
SUBJECT:
WELD ROD AND WIRE CERTIFICATION TO WHOM IT MAY CONCERN: This is to certify that all weld rod and wire usee ~~. t~e fabrication of the above referenced vessel was within the acceptable limits of the As~m Code, Section III, 1965 Editio~, with Addenda through Summer 1966, and General Electric Co. Specification 21All15 Rev. 5, Paragraph S.15 and a~so approved Chicago Bridge & Iron Company procedures which are included in Monticello Project Manual Volume II. The types of weld rod and wire used in this contract were as follows: SA3l6-E-S01S G SA233-E-701S SA29S-E-30S-l5 SA37l-ER-309ELC SA29S-E-309-l5 SA37l-ER-30SL SA29S-E-30SL-15 SA29S-E-30SL SA29S-E-308
2-23 CIIICACO BRIDCE & IRON COMPA~Y
SUBJECT:
WELD ROD JU~D WIRE CERTIFICATION SA298-E-309 SB29S-Inco 182 SB304-Inco 82 SB304-ERNiCr-3 SB29S-ENiCrFe-3 Linde 40 w/l% Ni or equal.
- CHICAGO BRIDGE &. IRON C01*!p~rY
2-24 CHICAGO BRIDGE & IRON COMPANY P. O. BOX 13308, MEMPHIS, TENNESSEE 3B113 g01 947-:31" DATE: r-tARCH 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.1671 10 X 63.167" INS. HDS. MANUFACTURER'S SERIAL NO. B-4697 MONTICELLO PROJECT, MONTICELLO, MINN. GE~I ERA L ELECTRIC CO. P.O. 20 5- 555 a2- I CB&I CONTRACT 9-5624
SUBJECT:
CLADDING CARBON CONTENT CERTIFICATION TO WHOM IT MAY CONCERN: This is to certify that the cladding has been accepted by General Electric Company as meeting the intent of Specification 21All12 Revision 5, Paragraph 8.10, and of the contract. Furthermore, the carbon content of the cladding does not exceed 0.08% as specified in the above General Electric Company Specification. CHICAGO BRIDGE & IRON COMPANY
! . ~.//
't/if/,(jl-llv?-C,-(,vn'v E
- E
- VARNUt4 NUCLEAR QUALITY ASSURANCE MANAGER
2-25
- CHICAGO BRIDGE & IRON COMPANY 1". O. sox 13308, MEMPHIS, TENNESSEE 39113 SOC1 947-3111 DATE: Harch 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' ID x 63.167' INS. HDS. MANUFACTURER'S SERIAL ~O. B-4697 MONTICELLO PROJECT, MONTICELLO, HINN. GENERAL ELECTRIC CO. P.O. 205-55582-1 CB&I CONTRACT 9-5624
SUBJECT:
WELDER QUALIFICATION CERTIFICATIO:~ TO w"HOM IT MAY CONCERN: This is to certify that all welder qualifications for bo~h shep and field fabrication of the above referenced vessel were performed in accordance with the ASHE Code, Sec t ; 0 r: s I I! and I X , 1965 Edition, with Addenda through Summer 1966, and *General Electric Specification 21All12 Rev. 5, Paragraph 9.0. Copies of these qualification records are on file with Chicago Bridge & Iron Company and will be furnished to General Electric Co. upon written request. CHICAGO BRIDGE & IRON COMPA~Y
.:~//
,,/ (-1,/ti./l...-n..t-vr/L E. E. VARl.~UM NUCLEAR QUALITY ASSURANCE Y~NAGER
2-26 CI-iICAGO BRIDGE & IRON COMPANY P. O. BOX 13306, MEMPHIS, TENNESSEE 36113 90' 947-3111 DATE: March 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' ID x 63.167 INS. HDS.
~~NUFACTURER'S SERIAL NO. B-4697
~10NTICELLO PROJECT, Zvl0NTICELLO, r*lnm.
GENERAL ELECTRIC CO. P.O. 205-55582-I CB&I CONTRACT 9-5624
SUBJECT:
PARKERIZING CERTIFICATION TO WHOM IT MAY CONCERN: This is to certify that the Parkerizing performed on equi?~e~t supplied with the above ,referenced vessel was done in accordance with federal Specification TT-C-490 dated !-larch 30, 1961, titled: "Cleaning Methods and Pretrea~"'le~-:' of Ferrous Surfaces for Organic Coatings", this procedure was approved by General Electric Co. for use on this contract, November 28, 1967. Process was perfo~ed for Chicago Bridge & Iron Co. by Hayes Aircraft Corporation, Birmingham, Alab~a. CHICAGO BRIDGE & IRO~ COMPA.,"JY U/dh,.-~f..:.t.nz.-, E. E. VARNU~l NUCLEAR QUALITY ASSU~'\JCE ~~'\JAGER
2-27
- CI-fICAGO BRIDGE & IRON COMPANY P. O. BOX 13308, MEMPHIS, TENNESSEE 38"3 DATE; March 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' ID x 63.167' INS. HDS. MANUFACTURER'S SERIAL NO. B-4697 MONTICELLO PROJECT, l*10NTICELLO, MINN. GENERAL ELECTRIC CO. P.O. 20S-55582-1 CB&I CONTRACT 9-5624
SUBJECT:
MATERIAL CERTIFICATION TO WHOM IT MAY CONCERN: This is to certify that the material in.the above referenced vessel is in accordance with the ASZ*lE Code, Section III, 1965 Editiori, with Addenda through Summer 1966, anc General Electric Co. Specification 21A1112 Rev. 5, Paragraph 8.0 anc also approved Chicago Bridge & Iron Company procedures MS-l Rev. 0, MS-2 Rev. 1, 1-1S-3 Rev. 3, lvlS-4 Rev. 3, MS-5 Rev. 1, MS-6 Rev. 1. MS-7Rev. 2, M5-B Rev. 0, MS-9 Rev. 0, MS-10 Rev. 0, 115-11 Rev. 2, MS-12 Rev. 0, MS-13 Rev. 2, MS-14 Rev. 0, MS-15 Rev. 2, MS-16 Rev. 1, MS-17 Rev. 1 and HS-MISC.-l Rev. 1 (above procedures included in Monticello Project Manual Vol~T.e I). CHICAGO BRIDGE & IRON COMPANY
~/dl./Yt.'v~
E. E. VARNUM NUCL~~ QUALITY A5SU~~CE ~~~AGER
2-28 CORE SPR"',/~IIIIS*Z. ASOS~lr 5AF'E. 1tNO ... MS,-,,,, INCONE\.... THEttl'/\A.~ 5L££.YE- .... ~I' TP 30~
- II'I~TR:U""'ENT ... Tu..."\N NOZ'Z..-M!;i-13 S81'-(.,
SAt--E £NO-M~*'" ~ C.~F8-
2-29
- CHICAGO BRIDGE & IRON COj\tlPANY P.O. BOX , 3 3 0 a. M E MPH IS. TEN N E SSE E 3 a, , 3*
eo, S47-3'" DATE: March 11, 1969
REFERENCE:
BOILING WATER NUCLEAR REACTOR VESSEL 17.167' ID x 63.167' INS. HDS. i'1ANUFACTURE~' S SERIAL NO. B-4697 1'10NTICELLO PROJECT, HONTICELLO, l~I~u-l. GENERAL ELECTRIC CO. P.O. 205-55582-1 CB&I CONTRACT 9-5624
SUBJECT:
FABRICATION TEST PROG~~ 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 21All12 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-l Rev. 0 which correspo~ded to the cold forming performed on the plates of the Reac~or Vessel. Paragraph 2.2.3 of Attachment "B" was complied with by SUbmitting "For Information Only" the test reports for 80% T tensile test on Chicago Bridge & Iron Company letter BB~-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 t~e meetings held in San Jose, California, August 22 through August 25, 1966, in the following manner: Page 10 Item G-l "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 GE Specification, 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 COMPAlSY
SUBJECT:
FABRICATION TEST PROGRAM CERTIFICATION Attachments:
- 1. CB&I Drawing T-4 Rev. 4, Results of Charpy V-notc~
impact tests as per Paragraph 2.2.4 of Attac~~ent B.
- 2. CB&I Drawing T-S Rev. 3, Results of tensile tests as per Paragraph 2.2.2 of Attac~~ent B.
CHICAGO BRIDGE & IRON COMP&~Y
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MONTICELLO 3-1 EXHIBIT 3 TENSILE TESTS SPECIMENS OF 80 PERCENT PLATE THICKNESS
- REV 4 12/85
3-2 Tensile Tests Specimens of 80 Percent Plate Thickness Nuclear Reactor Vessel GE/APED for Northern States Power Company Monticello, Minnesota. for CHICAGO BRIDGE AND IRON COMPANY Contract 9 - 5624
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;" T' 1 8 : l-.3SJ- I Richard N. Wright, IIi Professional Engineer i .'- ;. :', -!l- .~- / New York License 035399 Department of Civil Engineering University of Illinois Urbana, 111 inois January 1968
3-3
- I. Introduction Test procedures and results are described for tests conducted for Chicago Bridge and Iron Company of six 80 percent thickness tension specimens for Nuclear Reactor Vessel GE/APED for Northern States Power Company, Monticello, Minnesota. The tests were conducted at the same time as similar tests for Nuclear Reactor Vessel GE/APED for Central Vermont Power Company, Vernon Dam~ Vermont. Facilities of the Department of Civil Engineering, University of Illinois, Urbana, Illinois were used in accord with a Memorandum Agreement for Commercial Tests between the Department and the Chicago Bridge and Iron Company. R. N. Wright, Associate Professor of Civi 1 Engineering, supervised the testing. G. K. Sinnamon, Professor of Civi 1 Engineering, and V. J. McDonald, Associat"e Professor of Civi 1 Engineering and Principal Research Engineer, participated in planning and conduct of the tests.
Instrumentation and test procedures are described in the following section. Test results are described in the last section.
- 2. Instrumentation and Test Procedures Test specimens were prepared by Chicago Bridge and Iron Company.
Dimensions are given in their drawing T6 Rev 1 for Contract No. 9-5624 which is reproduced here as Fig. 1. Gage diameter was 4-5/16 in. and gage length 13 in. Specimens were delivered to the Department of Civil Engineering on September 11, 1967 and stored in Talbot Laboratory adjacent to the testing machine until tested during the week of January 8, 1968. Test procedures conformed with ASTM E8 66. Test temperatures ranged from 70 to 75 degrees Fahrenheit
- 3-4 The testing machine used was the University of Illinois 3,000,000 lb. capacity universal testing machine. Loads were recorded from the load indicator of the machine. Ames dial indicators with .0001 in. divisions were used in measuring elongations from zero load to approximately 1 percent strain; and Ames dial indicators with .001 in. divisions were used from zero load to maximum load. Initial gage lengths and elongations after rupture
-were measured using a steel scale with .01 in. divisions. Initial gage diameters and diameters after rupture were measured using micrometers with
.001 in. divisions.
Figure 2 shows a specimen in place in the testing machine with the dial indicators supported by a split ring and angle device. Diametrically opposed, spring loaded, gage points fit holes 1/16 in. diameter by 1/8 in. deep drilled into the specimens to support the rings. A third gage point between the-diametrical ones prevents rocking of the ring. The angles attached to the rings hold the two .001 in. division Ames dials at 4 1/4 in. from the axis of the specimen and the two .0001 in. division Ames dials at 6 1/4 in. from the axis of the specimen. Indicators are pulled by copper wires at-tached to a similar split ring and angle device. Figure 3 shows indicators in place during loading and the television cameras used to read elongations. Figure 4 shows recording of load from the load measuring system of the testing machine and recording of elongations from closed circuit television receivers. During the first two of the six tests, SR-4 electrical resistance
~train gages were used to measure strain prior to yield in order to check upon the accuracy of the dial indicator system. SR-4 gages showed slightly greater strain during the first loading increment (50 to 100 kips) than the dial indicators; thence to yield essentially identical changes in strain were
3-5 ~ recorded by the two procedures. The discrepancy in the initial increments is attributable to reseating of the gage points supporting the dial indicators following the reversal _& strain direction during the preliminary steps of loading desc*ribed below. 'Only strains obtained from elongation measurements with the dial indicators are reported here. Preliminary steps of the testing consisted of centering the specimen in the upper head of the machine, recording initial elongations at zero load, fastening the lower head of the specimen and loading to 100 kips to set the grips, reducing load to 50 kips and reading elongations which were used as base values in reducing stress-strain data. Elastic range loading began with increase of load to 100 kips and reading of elongation, followed by increase of load and elongation reading in 100 kip increments. Upon noticeable yielding, ~ a slow deformation rate was maintained, load and elongation were recorded at intervals of approximately .01 in. elongation until pronounced strain hardening at an elongation of about. 1 in. Then load was reduced and the .0001 in. dial indicators were removed. Continuous deformation was resumed; load and elonga-tion were recorded at intervals of about .05 in. elongation until maximum load was observed. Dial indicators were then removed from the specimen and it was deformed to rupture. In the first two tests somewhat fewer readings were made in the post yield range. In the inelastic range to maximum load, strain rate did not exceed .01 in./in./minute. In the elastic range stress
.rate did not exceed 10,000 psi/minute.
In the first test the. lower ring came loose twice during the post-yield range of testing. The deformed gage length was measured to 0.01 in. accuracy after the dial indicat'ors were removed at maximum load. This measure- ~ ment provided a basis for computing strains from dial readings for the majority of the post yield region; a small region of uncertainty is shown by dotted lines in Fig. 5. The cause of the loosing of the lower ring was improper spring
3-6 loading of the gage points. One more loosening of the lower ring occurred in the process of obtaining proper adjustment. It was in the second test. Fig. 6, at a stress of 81.9 ksi. It was determined that dial readings were not substantially affected by the loosening and replacement.
- 3. Test Results Test results are summarized in Table I. Shown for each specimen are: stress-strain curves, Fig. 5 through 10; photographs of the two fracture surfaces, Fig. II through 16; and photographs of the broken specimens with fracture surfaces fitted together, Fig. 17 through 22. Specimens are identified by the numbers provided by Chicago Bridge and Iron Company and an "OT NO.'I assigned by the writer to facilitate identification of individual tests and specimens.
Table I shows that, test'results meet tensile requirements of ASTM A 533, Grade B, Class I steel. The elongation in 13 in. is not directly comparable to the standard elongation in 2 in. for a 1/2 in. diameter specimen. Larger elongation is observed because the gage length is only three times the diameter for these specimens. If, however, an additional '4-5/16 in. of gage length were considered to be present and to elongate by the 10 percent uni-form strain typical of Fig. 5 through 10, the elongation in 17-1/4 in. would in every instance exceed 20 percent. Welded specimens, denoted by T3-2X, showed substantially the same properties as the unwelded. It is apparent in Fig. 20 through 22.that fracture of the welded specimens was ductile and occurred in the base metal well away from the weld. A clear indication of weld yielding at a stress of 67 ksi ap-pears in Fig. 10. Possible weld yielding at 64 ksi is suggested by the stress-strain curve shown in Fig. 9; for the stress-strain curve of Fig. 8, yielding of base metal and weld appears to have occurred simultaneously.
3-7 TABLE 1.
SUMMARY
OF TEST RESULTS Specimens Yield a Tensile Elongation Reduction 5624 Strength Strength in 13 inches in Area A 0998 2 ksi ksi percent percent T3-2 OT 63.0 84.4 28.3 61.2 aT 2 62.0 84.2 25.8 61.1 aT 11 61.3 83. 1 23.7 62.4 T3-2X aT 5 65.8 85.6 25.9 61.0 OT 8 63.7 83.6 26.7 60.9 OT 9 64 .. 8 84.5 25.8 58.9 aYield strength at 0.2 percent offset.
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3-24 FIGURE 19. Fractured ~pecimen, 5624 T3-2 A0998 2, OT 11
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, , . LU[(ENS S'i't:L:L COMPANY At'i'i:
PURCHASER. CO... ,ESVILlE. PA. o ... n, 11-26-66
- 10. ChIc aBo Bridge & Iron Co. CONSIGNEf, TEST CERTIFICATE
~~. G.H. Putman,P.A. '. Mill olon UO. CUSTOMER r.o.
P.O. Box 277 Same
'J: .-'l BirmIngham, Ala. '35202 43211-1 5624 ..,-'\ Boyles, A1 MB 112166 E!~
SPECifiCATIONS, . . . A-533-65 Gr.B Class'1 Mod.by C.B.& I. Spec. MS-l DTD 8/25/66 Fbx. 80000 Cont.D 9-5624 liND un O. K. HOMOGfNI1Y un O.K. CHEMICAL ANALYSIS MElT NO. C MH P S Cu SI NI Ci Mo V TI AI. ~ ~a...!.n ~!Ze I. C1946 22 L35 010 015 ,g 47 V.I. ~.Stee T#7 BU8 A0998 20 1.27 008 017 ~a II II I II 49 Tin B/18 i PHYSICAL PROPERTIES Sl .... WlflD TlNSlll " 1 ONG. IMP ...CTS MflT NO. flol flol " I.A. DESCRIPTION t",. liDO liDO IN 2
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MONTICELLO 4-1 EXHIBIT 4 SU~Y STRESS REPORT '. REV 4 12/85
4-2 S~'1\ary Report Monticellc--NSP Reactor Vessel CB&I Contract 9-5624
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4-3 CHICAGO BRIDGE & IRON COMPAN':' OAK BROOK ENGINEERING
- 5UMMARY REPORT INDEX MON~ICELLO NUCLEAR REACTOR Page 2 Introduction 3 Main Closure Flange (5-1*)
3 Main Closure Flange (T-l*) 7 Skirt Junction (5-2) 7 Skirt Junction (T-2) 8 Shroud Support (5-3) 8 Shroud Support (T-3) 12 Feedwater Nozzle (5-4) 12 Feedwater Nozzle (T-4) 14 Control Rod Drive Penetration in Bottom Diameter Dollar Plate(5-5) 14 Control Rod Drive Penetration in Bottom Diameter Dollar Plate(T-5) 16 3" ¢ Control Rod Drive Hyd. Return Nozzle (5-6) 16 3" ¢ Control Rod Drive Hyd. Return Nozzle (T-6) 18 Core Spray and Flooding Nozzle (8" ¢) (5-7) 18 Core spray and Flooding Nozzle (S" ¢) (T-7) 20 Recirculation Inlet Nozzle (12" ¢) (5-8) 20 Recirculation Inlet Nozzle (12" ¢) (T-G) 22 Recirculation Outlet Nozzle (36 x 28) (5-9) 22 Recirculation Outlet Nozzle (36 x 28) (T-9)
- 24 18" ¢ 5team Outlet Nozzle (5-10) . I 24 18" ¢ Steam Outlet Nozzle (T-10) 26 Cor~ Differential Pressure and Liquid Control (S~ll) 26 Core Differential Pressure and Liquid Control (T-ll) 26 Instrumentation Nozzle (6" ¢) (5-12) 26 Vent (4" ¢) (5-13) 27 Instrumentation Nozzles (4" ¢) (5-14) 27 Jet Pump Instrumentation Nozzles (4" ¢) (5-15) 28 Refueling Bellows (5-16) 28 Refueling Bellows (T-16) 30 Stabilizer Brackets (5-17) 32 Brackets (5-18) 33 Main Shell 5tress Analysis (5-19) 34 Miscellaneous Stress Analysis (5-20)
* "51" ~n . d-~cates a stress analysis and "Tl" indicates a tr.ermal analysis.
S.. bi.ct _ _Monticello ProJ' ect - Reactor _ _ _ _ _.........;~~.;;..;;.....;;..._....:.;.;:;.::.;:.:.;:.=._ CO"', 9 - d- p2. .
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4-4 CHICA.GO Bi(.OGE 0. IRa", COMPAl'i';' OAK BROOK EHGINEEiWoIG SU~MV~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 21All12, 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 S~~ary Report is being submitted as required in Paragraph 6.8 of the General Electric Purchase Specificat~on mentioned above .
- Sublect :'iC~':'ICELLO REACTOR VES 3EL Cont. 9-5624 Dote_ay 0;';
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4-5 CHICAGO BRIDGE & IRON COMPANY
SUMMARY
OF STRESS ANALYSIS FOR OAK BROOK ENGIN E ERII'IG MAIN CLOSURE FLANGES The stress analysis for the main closure flanges and the studs was conducted in accordance with requirements of GE D~sign Specification No. 21All12, Rev. 5, dated 9-30-68. Temperature distributions used in this analy-sis are presented in Section Tl of the stress report. The minimum required stud cross-sectional area, per ASME Code, Section III, Article I-12, was found to be 1177.14 square inches (page I-Sl-7). This was based on an allow-able stress Sm = 36,325 psi at 575°F. The actual cross-sectional area, provided by 64 studs with 5-1/16 inch shank diameter and 7/16" extensiometerhole, is 1278.61 square inches. The average and maximum stud service stresses (per ASME III - N-4l6.l) were found to be criti-cal during startup at 270 minutes into the transient, with their respective magnitudes being 47,929 psi and 89,824 psi (page I-Sl-84). The average stud temperature at this time is 340°F. Allowable stresses at this tem-perature are 2 Sm = 79,280 psi for the average stress and 3 S~ = 118,920' psi for ~~e maximum stress. The stud fatigue anlaysis was performed in accordance with Par. N-4l6.2 of ASME Code, Section III. The peak stress intensity ranges were computed at the root of the thread using a fatigue strength reduction factor of
- 4. The cumulative usage factor was found to be 0.5637 which is well within the allowable of 1.0 (page I-Sl-100).
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4-6. CHICAGO BRiDGE 8. IRON COMPANY OAK BROO/( ENGINEERING The basic stress intensities in the main closure flanges and the adjacent top head and cylindrical shell per ASME
<:. ie, Section III, N-4l4, 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-Sl-64) 0 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 o*f 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-Sl-6S). 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 .S/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 Subjec:t MO'i\'1"TC-PT i Q Bt:'ACTOB VESS-::T, Cant. D ... u .... 9-5624 Dote _ _ Sy M""t. AEE Sht 4 of 34 _ Rev.Nr)o_C,:'e _ __
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 cylindrical shell junction. The allowable stress intensity range in this case is 3 Sm = 80,100 psi. It was found that all the requirements of the ASME Code, Section III, Par. N-41S.1 could be satisfied for the main closure flanges, and therefore no fatigue analysis of the same is required. 9-5624 Oote_By AEE Subject MONTICET,T,O--REACTOB V';:'SSEI, Cont.
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4-8 CnlCAGO BRiOG~ !. IRON COM~ANY OAK Si\OO!( ENGINEeRINC
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4-9 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSIS FOR SUPPORT SKIRT AND ITS JUNCTION TO BOTTOM HEAD Using the data contained in the contract specifications and temperatures calculated in Section T2, the stress analysis has been done in Section S2. The maximum value of seismic stress along the support skirt is 2623 psi. The maximum local membrane and bending primary stress intensity occurs at the inside point of the skirt-to-bottom head junction and has a value of 10,625 psi. The Code allowable at design temperature for the sum of all primary stresses is 1.5 Sm = 40,000 psi. The value of the maximum range of primary plus secondary stress intensities, which also occurs at inside point of the junction, is 55,580 psi. The Code limits this range to 3 Sm = 80,000 psi. The same point is also most critical from a fatigue standpoint. The conservatively calculated value of the fatigue usage factor at this point is 0.40. MONTICELLO S"bject _ ______ REACTOR _ _ _ _VESSEL _ _ _ _ _ _ _ _ Cont9-5624
4-10 CHICAGO BRIDGE & IRON COMPANY OAK aROOK ENGINEERING
SUMMARY
OF RESULTS - STRESS ANALYSIS OF SHROUD SUPPORT The static analysis of Subsection C of Section S3 indi-cates that al~ ~o~a requirements for the secondary mem-brane plus bending combined with local membrane, local and general primary membrane and also primary bending stress intensities, have been met. The maximum secondary membrane plus bending combined with local membrane stress intensities of 23,970 psi occurs at point 5 of the main shell (see page S3-57), and 26,099 psi occurs at point 17 of the shroud 0 Both of these stresses are below allow-ables of 3 Sm = 80,000 psi and 3 Sm = 70,000 psi respec-tively (see Figure 1). Local membrane stress intensities of 15,555 psi at Section 7-8 and 1?,575 psi at Section 17-18 are also within the allowable limits of 1.5 Sm = 40,000 psi and 1.5 Sm = 34,950 psi. Primary bending plus membrane and general primary membrane stress intensities are 26,740 psi at point b and 26,585 psi at Section a-b on the main shell which are below the allowable of 1.5 Sm = 40,000 psi and Sm = 26,700 psi. This also holds for internals with maxi-mum primary membrane stress = 7,910 psi at point.c and primary general membrane stress = 7,750 psi at Section c-d for Inconel material for which allowables are 34,950 psi and 23,300 psi respectively (see Figure 1). Subsection C of Section S3 also shows that the stilts which support the shroud will not buckle under the most critical compressive load *
- Subject MQNTICELLO REACTOR VESSEL Cont.9-5624 Dat._8y,....::::.JT.:.-_Sht~of~
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4-11 CHICAGO BRIDGE & IRON COMPANY Stress intensity calculations of Subsection E of Section S3 OAK BROOK ENGINE ERIHG show that maximum range of secondary membrane plus bending stress intensity for carbon steel is 62,100 psi at point 13, which is less than the allowable of 3 Sm = 80,000 psi, and Inconel material for the internals is 58,924 psi at point 30, also within the allowable of 3 Sm = 70,000 psi (see Figure 2). Stress analysis of the jet pump baffle plate was performed at the junctions to the main shell and shroud support. Re-sults are listed* under Subsection F for static loading and transients considered. The results in the static analysis show that local membrane stress intensity' is 17,327 psi at Section 3-4 and secondary membrane plus bending combined with local membrane stress intensity is 19,229 psi at point 3 within the allowables of 1.5 Sm = 34,950 psi and 3 Sm = 70,000 psi respectively. Results of loading plus transients stress analysis show maximum range of stress intensity of 63,656 psi at point 3 which is also b~low the allowable of 3 Sm = 70,000 psi (see Figure 2). Fatigue analysis performed under Subsection G shows a per-missible number of 4000 cycles and the usage factor of
.064 for point 7 based on stress results of Subsection E and 20,000 cycles and the usage factor of .012 for pOint 3 based on Subsection F (see Figure 2).
Sl.Ibject MONTTCET.I,O REaCTOR ¥ESSEI, Co"t.9-5624 Date_By JT Cneeked b~ Dote _ _ _ Re ... No Dote _ _ _ Re ... No. _ _ Dote _ _ _ Rev.No. _ _ 00'" _ __
4-12 CHICAGO BRIOGE & IRON COMPANY OAK BROOK ENGINEERING SPECIAL STRUCTURES OESIGN POINTS OF HIGHEST STRESSES F'IG. I pOINi /7
/
- Subject M* ,.:.,' i :' ir L\ (~ I? ": ~, -., -.. .,; \: h. ;:/7\ Cont. _ _ _ Dat. _ _ _ By _ _ _ Sht --.l:Q. of _'>_"1_ _
~ ~
e4 SSO Chf!Cklld bv o~ ____ Rw No _ _ DatI _ _ Rw No. _ _ DltI _ _ _ Rw. No. _ _ _ 0.,, _ _ __
4-13 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENG:NEERING SPECIAL STRUCTURES DESIGN I POINTS OF HIGHEST STRESSES fiG. 2-POINT7-.-........ ~__ pOINr30 Subjecc ".
~ I* , ;: . > \
CMWt. _ _ _ Date _ _ _ Bv _ _ _ Sht ~;,f _3_4_ _ 64 SSO Checked bV _ _ Oate _ _ _ Rev NO _ _ D8te~ R., NG. _____ D8te _ _ _ Rw. No. _ _ _ C~te _ _ __
4-14 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSIS FOR FEEDWATER NOZZLE Using the loadings contained in the contract specification and temperatures calculated in Section T4, the stress analysis has been performed in accordance with Article 4 of Section III of the ASME Code. The area replacement requirements of Article 4 have been satisfied. The calculated maximum general membrane stress intensity for the safe end i's 18,200 psi (page 54-10) com-pared to the allowable, at 575°F, of 18,200 psi. For the nozzle forging, the calculated maximum general primary membrane stress intensity is 14,218 p~i compared to the allowable of 26,700 psi
- The maximum local membrane and bending stress intensity due to design pressure plus nozzle loads is 22,5.80 psi (page 54-10) at section AA on the attached sketch. The allowable stress intensity is 105 Sm = 27,300 psi.
The maximum ranges of primary plus secondary s~ress inten-sity are 26,540 psi on the inside of section DD and 59,600 psi on the inside of section CC for the safe end and noz-zle forging respectively. (See pages 54-35 and 54-33.) The Code allowable ranges are 56,040 for the safe end ma-terial and 80,100 psi for the forging. The allowable number of fatigue stress cycles is 1760 ver-sus a specified number of 15000 Subject MONTTCET.T.Q REACTOR VESSEI. Cont. 9-5624 Date_By JH
- "'.. CI... "I....j ..... 1'\ .... 0 ...... . "'_6. 0 ...... "' ...
4-15 CHICAGO BRIDGE & IRON COMPANY OAK BFcOOK ENGINE ERING
.._4. 83CJS.: IJ80 I I i
~\~
0-'
-I I 1
1 I I 1 I i FEfl>WATER NOri:1.E M4A rH~V #4b Subject t<1flN r Ie tiL () 5
),13
' _ 0f3~
Go 6.08 eneeked b",_ _ Date _ _-_'_R.".No, _ _ Date,_ _ _ Re".No._Date _ _- Re".Mo._Oof. _ _
4-16 CHICAGO BRIDGE & IRON COMPANY OAK- BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSIS eRn PENETRATIONS The maximum primary plus secondary stress intensity in the stainless steel housing is 38,408 psi compared to an allowable of 3 Sm (= 52,800 psi), at point 1. The maxi-mum value for the Inconel stub tube is 58,351 psi at point 2 against an allowable of 3 Sm (= 60,000 psi). The maximum alternating stress intensity occurs at point
- 3. This value is 79,634 psi. The allowable number of cycles from the applicable design fatigue curve is 2900 against 370 specified cycles.
The points referred to above are shown in the sketch on the following page. The sketch shows an outermost pene-tration which is found to be more critical than the center penetration.
- Subject MONTTCEI,t,O REACTOR VESSEL Cont. 9-5624 Dote_By MSM ShtJ:.!.of~
4-17 e'" CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGIHEEiW~G CR::C PENS. TRATIONS au T E. i? HO!:T G, EO M E. TR'I I
~-- - I~ - - -AREA- -" INCONEL I
i
'STU'B-TUt3E iI i
I
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I II
- 2. ~------- '----- - r---LVe-L-:D-.....:----*-~
, ARE4 : I
~~---- - - --- --t I CARBo'"
E."TEEL V e. s s s: L.. WALL.. e* _-r--L_-__-_-_-__-_____"__~ A Sublect MONTICgLLQ REaCTOR "PR )Jeer CO"toe2-S-b~Lj Date By M~ M Sh, 15 of~ GO 140& enecked by Dote _ _ _ Re"oNo _ _ _ oote_ _ _ Re"oNoo Dote _ _ _ RevoHoo_ 00'. - _ _
4-18 CHICAGO BRIDGE 8. IRON COMPANY OAK BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSIS 3 n 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 782 cycles,
- based on the applicable design fatigue curve.
The points referred to above are shown in the sketch on the following page *
- Subject MONTTCET.T.O REACTOR VESSET.
. "" .... to' .. "_6_ n ..... _ _ _ Rev.No._Oo,,, _ __
4-19 CHICAGO BRIDGE & IRON COMPANY OAK B~OOK ENGINEERING T '%. II At o;;~ CRl'HSR NOci!:LE. I I I I I I II 10
*! 1"1-i W E.L..'b L..INe.
I I J Subject MON"'f'IC.e.Ll.O REAl.TOR VE~iL e~oa Oeclced by,:-_ _ Date _ _ _ Re".No _ _ _ Date_ _ _ _ _ Re".No. _ _ Dote _ _ _ Re.,.Ho._Date _ __
4-20 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSIS for CORE SPRAY NOZZLE Using the data contained in the contract specifications and temperatures calculated in Section T7, the stress analysis has been done in Section S7. The calculated maximum general membrane primary stress inten-sity, for the safe end, is 14,050 psi (page S7-7) compared to the allowable, at 575°F design temprature, of 23,300 psi. For the nozzle, the calculated maximum general membrane primary stress intensity is 12,550 psi (page S7-8) and the allowable at design temperature is 26,700 psi
- The maximum local membrane and bending primary stress intensities are 24,783 psi, 2,656 psi, and 11,123 psi for the safe end, sleeve and nozzle respectively, (page S7-86). The allowables for corresponding materials in the same order are 35,000 psi, 23,700 psi and 40,000 psi."
The maximum range of primary plus secondary stress intensities are 28,262 psi, 29,445 psi and 8,157 psi at points 8, 22 and 31 respectively, (page S7-92). These points are located on the safe end, sleeve, and nozzle, in that order; with respective allowables of 70,000 psi, 47,400 psi and 80,000 psi.
~he most critical point from the fatigue standpoint is point 11.
The conservatively calculated value of the fatigue usage factor at this point is 0.52.
- " .ect
~oints ing page.)
" _MONTICELLO referred to above are shown on the sketch of the follow-
_ _ _ _ _ _REACTOR _ _ _ _VESSEL _ _ _ _ _ _ _ Cont:'.-0 5624 3/3/;:;9_ Do,._'By, KM_ _ 18 3. S b Sh'_~f--=:"- 1'\ ..... D *** ~_ 1"\_ ** D *** ~_
"'_... n_._
4-21 CHICAGO BRIDGE & IRON COMPANY OAK aROOK ENGINEERING POltI'T 8--~ PO\NT l \
......_ _ _ POI",T ~\
~ol ... T zz. .,
CoRE
".... ... ---_ .... 0 *** w_ II.... W... . _Date _ _ _ Rev.Ho._Oate _ __
4-22 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERIHG
SUMMARY
OF STRESS ANALYSIS for RECIRCULATION INLET NOZZLE Using the data contained in the contract specifications and temperatures calculated in Section T8, the stress analysis has been done in Section S8. The calculated maximum general membrane primary stress inten-sity, for the safe end, is 12,900 psi (page S8-4) compared to the allowable, at 575°F design temperature, of 15,800 psi. For the nozzle, the calculated ~imumgeneral membrane pri-mary stress intensity is 16,600 psi (page 58-5) and the allow-able at design temperature is 26,700 psi. '. The maximum local membrane and bending primary stress intensi-ties are 17,107 psi (page 58-9) and 23,488 psi (page 58-11 for the safe end and nozzle respectively. The allowables for corresponding materials in the same order are 23,700 psi and 40,000 psi. The maximum range of primary plus secondary stress intensities are 33,670 psi (page S8-30), 47,830 psi (page S8-33) and 43,550 psi (page 58-31) at points 4, 9 and 6 respectively. These points are located on the safe end, sleeve and nozzle, in that order; with respective allowables of 47,900 psi, 47,900 psi and 80,000 psi. The most critical point from the fatigue standpoint is point 8. The calculated value of the fatigue usage factor at this point is approximately 0003. (Points referred to above are shown on the sketch of the follow-
- ing page.)
MONTICELLO REACTOR VESSEL S"bject _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Cont. 9-5624 KM Dote_By,...!::..!.-_ s"', _"0 " 3.; n w'~ 0 .... 0" ** I)"** 0", ..
4-23 OtICAGO BRIDGE & IRON COMPANY OAK SROOK ENGINEERING n
""'-t-- ?0 ,"'T 4-
-Ri.<::.\;'~~~~TlaN- '~L~T- Nci~,-6.-
Mc:=>>,..:t.\\<:"~,-\....O ~EJ'-c.\C)~ '.J~Se.\... CO"t.~-5b*1.4 Dote 314./b"By Subject k M Sht-1l:.ofl.L
,... .' "_A. D *** w ...
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4-24 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSIS FOR RECIRCULATION OUTLET Using the loadings contained in the contract specification and temperatures calculated in Section T9, the stress analysis has been performed in accordance with Article 4 of Section III of the ASME Code. The area replacement requirements of Article 4 have been satisfied. The calculated maximum general membrane stress intensity for the safe end is 13,806 psi (page S9-36) com-pared to the allowable, at 575°F, of 15,800 psi. For the nozzle forging, the calculated maximum general primary membrane stress intensity is 12,261 psi compared to the allowable of 26,700 psi. The max~um local membrane and bending stress intensity due to design pressure plus nozzle loads is 14,777 psi at section AA on the attached sketch (page S9-37). The allowable stress intensity is 1.5 Sm = 23,700 psi. The maximum ranges of primary plus secondary stress inten-sity are 26,540 psi on the inside of section BB and 36,700 psi on the inside of*the same section at the safe end-nozzle forging junction. See page S9-l9 of the report. The Code allowable ranges are 47,400 for the safe end ma-terial and 80,100 psi for the forging. The Code allowable number of fatigue stress cycles for the maximum stress amplitude is 41,720 compared with the 400 cycles specified *
*Subi.ct_....M
...Q~N T...IIolIICr.aE"""L""L.:Q"-'RE=A~C.=.T~Q;.:;R~VE:.=S;:.::S;:.::E=.:L=-_ _ _ Cont. 9 - 5 624 Dot._By JH
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os Ch.ck.d b'l,_ _ _ Dat. _ _ _ R.".No*_ _ _ Dat., _ _ _ R.".No. _ _ Dot. _ _ _ Rey.No. _ _ Defe _ __
4-25 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING i i- /2. q l" __ 11~_230 i : I~ 0"._
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Subiec' _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ con'. _ _ _ Da'!l1!M By jij Sht 23 ,,1 3.;
- e. os Checked by~_Date_...;,.,._Re ... Na _ _ Date _ _ _ _ Re ... No. Da'e R.... N o . _ Oote _ _
4-26 CHICAGO BRIDGE 8. IRON COMPANY OAK BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSIS 18" STEAM OUTLET NOZZLE The maximum primary plus secondary stress intensity in the safe end is 11,924 psi at point 1 against an allow-able of 57,450 psi (3 Sm at design temperature)
- In the nozzle forging, the nozzle-vessel junction is the highest stressed point (point 13). Based on the stress index method, the maximum pressure .stress intensity is 88,100 psi. To this is added the maximum thermal stress intensity of 19,592 psi and the additional stress inten-sity due to .the pipe reactions at the point which is 10,919 psi, giving a total peak stress intensity of 118,621 psi and an alternating st~ess amplitude Salt of 59,310 psi. This gives an allowable number of cycles of 2500 which is more than the expected 532 cycles.
The points referred to above are shown in the sketch on the following pageo
4-27 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING s.,. E A M I S'l f OVTLE-r NO=-<it.I..E n C*
- j J 2
TI II I I I I
~ 10
- II 1'2-11 Subject MONTIC.ELLO 'REAC.JPR VI=.~$EL Cont. _ _ _ _ Dot.31~lb!)Sy M.$t1 Sht~of~
GO 640B Checked by~ _ _ Dote _ _ _ Rev.No _ _ _ Dote _ _ _ Rev.No. _ _ Dot. _ _ _ RevoNoo_ Date _ __
4-2c CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSES for CORE DIFFERENTIAL PRESSURE AND L~>c":'-~~ CONTROL NOZZLE, HEAD COOLING SPRAY AND INSTRUMENTATION NOZZLES, VENT NOZZLE, INSTRUMENTATION NOZZLES, JET PUMP INSTRUMENTATION NOZZLES, DRAIN NOZ ZLE , HIGH PRESSURE SEAL LEAK DETECTOR NOZZLE and LOW PRESSURE SEAL LEAK DETECTOR NOZZLE The maximum primary membrane stress intensity for the core differential pressure and liquid control nozzle is 6076 psi (page Sll-6), compared to ~e allowable, 15,800 psi
- This nozzle has been exempted from fatigue analysis in ac-cordance with the rules of Par. N-4l5.l of Section III, ASME Code.
The maximum primary membrane stress intensity for _the head cooling spray and instrumentation nozzles is 2849 psi (page S12-7), compared to the allowable, 15,800 *psi. This nozzle has been exempted from fatigue analysis in accord-ance with the rules of Par. N-4lS.l of Section III, ASME Code. The maximum primary membr~ne stress intensity for the vent nozzle is 2501 psi (page S13-6), compared to the allowable, 1~,800 psi. This nozzle has been exempted from fatigue analysis in accordance with the rules of Par. N-415.1 of Section III, ASME Code * - ... r_ .... "'" Subi.c:t _ _M~O.l.llN....
~t.. __ t. .... ..1 .....
T.... .......Sil.l Sil-'EiW..... I"",C"",E.....I"""'fIoilo.O",--,R~Ew.Ag,:CIoo.T~O~R-.lIVE
"'.a . . D ...... _ ".a_
L ___ Co"t. D ...... .. 9- 5 6 2 4 Dote_By JH
"'_a_
Sht 26 of 34
4-29 CHICAGO BRIDGE 8. IRON COMPANY OAK BROOK ENGiNEERiNG The maximum primary membrane stress intensity for the jet pump instrumentation nozzles is 10,489 psi (page S15-6) , compared to the allowable, 15,800 psi. This nozzle has been exempted from fatigue analysis in accordance with the rules of Par. N-4l5.l of Section III, ASME Code. The maximum primary membrane stress intensity for the in-strumentation nozzles is 5796 psi (page S14-6), compared to the allowable, 15,800 psi. This nozzle has been ex-empted from fatigue analysis in accordance with the rules of Par. N-4l5.l of Section II,I, ASM.B Code.
.\/
....E Subi.ct_......Io:lM~Q/.I,;NuT~I...:C ....
L... ....Ac.C~T..l.Qu;R~V::JE~S~S~E""'L'--_ _ _ Cont. 9 - 5624 Date_By JH L~Q"-'R~E s,,!'--=....::.~t ~ n", ** n,. ** R .... ,Nn, Ont ..
4-30' CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSIS FOR REFUELING BELLOWS USl.n~ the da.ta contained in the contract specifications and the temperatures calculated in Section T16, following is a summary of the stress analysis which is found in Sect~on S16. The calculated maximum general membrane primary stress intensity for the refueling bellow skirt is 3547 psi occuring during refueling at a point midway on piece #3. This is compared to the allowable at 70°F of 23,300 psi. (See Page I-S16-AIB of the S tress Report.) The maximum local membrane and bending stress intensity is 10,310 psi occuring at the inside face of part #2 at the junction to part #1. This occurs during refueling *. This is compared to . t h e allowable stress" intensity of 34,950' psi.at 70°F. (See Table 1, Page I-S16-1 of Section S16.) The maximum range of primary pl~s secondary stress intensity is 51,734 psi and occurs in the cooldown-steady state cycle at the inside face of the junction of part #3 and part #2. The maximum allowable stress intensity at 545°F is 59,070 psi. (See Pages I-S16-1 and I-S16-10.) The most critical point from a fatigue standpoint is at the junc-tion of part ~2 and #3 during the cooldown-steady state cycle. The fatigue usage factor at this point is .67. (See Page I-S16-2 of Section S 16. ) (Points referred to above are shown on the sketch on the following page. ) Subject Monticello Reactor Vessel
4-31 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGIi'JEs:RING SPECIAL STRUCTURES DESIGN [~; L.bOu)'5 =5t<rt5r Po, . .,., 6c:r~ ~Cn::::6* lC"'-'<:j O f'"1 rc
~" I<J'! ; ry Poi.n" A p,., 3/5J.i7 ~i.
H.... Po
- I I
POII.1i " 101 310 ~, ii Po jJ!.}/ c. R.~ fr; ~~ 51, 731 fSc f~fli ~ I 1 f Ollte _ _ _ R8\I No _ _ Data _ _ RIIY No. _ _ Date _ _ _ RIIY. No. _ _ _ Date _ _ __
4-32
- CHICAGO BRIDGE & IRON COMPANY OAK SROOK 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 l~ited 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 Intensi tYI
= 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 Cant. 9-5624 Dote_ByAEE
"'_a_ D *** w. n"** _ Rev.No. _ _ Dat,, _ __
4-33 CHICAGO BRIDGE & IRON COMPANY LOADING CASE 2 OAK BROOK ENGINEERING Actual Maximum Stresses Pure Shear Stress at Pin Hole = 19,551 psi Bearing Stress at Pin Hole = 27,767 psi Maximum Stress Intensi ty ~
= 26,854 psi At Face of Shell ' f Allowable Stresses Pure Shear Stress = 21,150 psi Bearing Stress = 42,300 psi
= 42,300 Maximum Stress Intensity psi I
Subject MONTICELLO REACTOR VESSEL n""** D ... w. "' ...
4-34 CHICAGO BRIDGE & IRON COMPANY OAK BROOK ENGINEERING
SUMMARY
OF STRESS ANALYSIS FOR BRACKETS Brackets were analyzed for loadings specified on GE Drawing 886D482, Sheet 7, Rev. 5. All brackets, except stabilizer brackets and steam dryer supports*, are included in this section. The brackets are subject to primary membrane and bending stresses for which the allowable stress intensity is 1.5 Sm in accordance with Fig. N-414, Section III of the ASME Code. The maximum stress intensities for the various brackets are as follows: BRACKET STRESS INTENSITY Core Spray 10,007 psi Guide Rod 18,982 psi Feedwater Sparger 314 psi Dryer Hold Down 9,052 psi Shell Insulation Support 2,500 psi Head Insulation Support 1,590 psi Upper Surveillance Bracket 8,350 psi Lower Surveillance Bracket 5,078 psi The allowable stress intensity for all brackets, except the guide rod brackets and the head insulation supports is 24,060 psi at 545°F. The allowable stress intensity for the guide rod brackets is 30,000 psi at 120°F and the head insulation supports 28,725 psi at 575°F.
- The steam dryer supports have been reanalyzed for the replacement steam dryer.
(Reference EC 14214) Subject Monticello Reactor Vessel Cont. 9-5624 Date By AEE Sht 32 of 34 GO 640B Checked by Date Rev. No. Date Rev. No. Date Rev.No. Date Revision 28
4-35 CHICAGO BRIDGE & IRON COMPANY
SUMMARY
OF STRESS ANALYSIS FOR OAK BROOK ENGINEERING TOP HEAD AND CYLINDRICAL SHELL In this section the maximum stress intensities in the top head and cylindrical shell due to combined loadings were computed. The loadings considered were: stud pre-load, internal pressure, dead weight of vessel and con-tents, insulation weight, horizontal and vertical seismic forces, horizontal jet reactions, stabilizer rod reac-tions, local bracket and nozzle reactions, refueling bel-lows support loads, and the'thermal loads. The maximum stress intensities and their locations were found to be as follows: the maximum general primary membrane stress intensity occurs in that portion of the cylindrical shell which is removed from gross structural
. discontinuities. Its magnitude is 26,375 psi which is within the allowable value of 1 Sm = 26,700 psi. The maximum local membrane stress intensity of 29,610 psi occurs at 15 inches below the bottom of the shell flange hub. This stress intensity was found to be local in extent and is less than the allowable value of 1.5 Sm =
40,050 psi. The maximum range of stress intensity for primary plus secondary stresses has a magnitude of 55,320 psi which is well within the allowable of 3 Sm = 80,100 psi. The location of this stress intensity is at the top of the hub of the head flange. MONTICELLO REACTOR VESSEL AEE
"" .... ""II ,
Subject CO",. 9-5624 Da'e_By S"'~=;~
- ) 6608 Checked bv,_ _ _ _ Date _ _ _ _ .. R..... N ... n". ** 0 ..... _ "'_... 0 ..... .. .. f"'III .....
4-36
- CHICAGO BRIDGE 8. IRON COMPANY
*SU~_~Y OF STRESS &~ALYSIS FOR OAK BROOK ENGIN E ERING 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 Cant. 9 - 5 6 2 4 Date~By JH Sht.2.! af J..L
- 06408 enecked by Oat. R.v.No. Date' _ _ _ _ Rev.No. _ _ _ Oate _ _ _ Rev.Na._Oate _ __
MONTICELLO 5-1 EXHIBIT 5-VESSEL FABRICATION AND ASSE}ffiLY REPORT REV 4 12/85
5-2 MONT1CEttO Mon~icello Rcc:.ctor Vessel, P~"J"; 1"-1 ~NE.RAL ELECTRIC CO. APe:O - SAN JOSE IV. 4.1 DIV:S:;:ON OF 'HORK A significant portion of the vessel fabrication was f~r=~==.cd in ~.e shop, just as would be done for a shop f~~~ca~=d v~~s=~. This work was in accordance wit."'" t."-le ASME Coc:e and G.:::. f~=ica.~io::
~uality control re~uirements. The balance of functio::al A
requira~ents ty~ical of t.~e vessel. site assembly area is shown in Fisure IV-l an artist1s rendering of t.~e completed composite reactor ~~d contai::rnent vessels is shown in Figure IV-2. 4.2 S~OP FABRICATION AI.'1D SUBASSE!;f.3LY viOR.'t( As much fabrication ~~d subassembly work as possible was perfo~ed at C3&IIS Bi~ngharn, Alabama, ~~d Greenv~lle, P~~sylvania, manufacturing plants. The overall job econo~cs favored t.~is approaCh because of the convenience of overhead handling equipment, utilization of t.""'e existinS shop labor pool
~d facilities for machining, heat treat.~ent, etc. Restric~io::
on shipping dL~ensioris (not weiSht) was t.""'e dete==~ni::s fac~or ~:: considering how much of t."l.e vessel assen-.bly work could be perfo=::-.ec. prior to shipment.
5-3 Monticello Reactor vessef, Page IV-2 An effort was made to clear completed shell rings 18 feet diameter by 10 feet 11-1/2 inches long and weighing 140,000 pounds. Although these rings could be barged to Minneapolis, the interconnecting railroad (Minneapolis, Northfield and Southern Railway) could not move the shipment from Port Cargill to the Great Northern Cedar Lake yard interchange. The Great Northern services the Monticello area. Overland truck handling clearance checks were also unsuccessful in finding an open route to the Great Northern Railroad: therefore, half ring sections were shipped from the shop. Figure IV-3 shows the shop assembled pieces for ~~e Monticello reactor yessel
- A l7-foot diameter support skirt extension with leveling devices attached was shipped in one piece to the site. It joined
.the stub skirt on the vessel bottom head to the long skirt constructed with the drywell. as shown in Figure IV-2.
A cold forming procedure was utilized to press the bottom head, shell, and top head plates. All plate material was detailed to the maximum length and width dimensions ~~at could be delivered from eastern mills and properly handled by the fabricating facility. The shell plates were purchased in the quenched and tempered condition and cold formed utilizing approved procedures. Before starting fabrication, all plates were inspected for size, thickness, surface condition and the mill stamps properly 411bentified. Ultrasonic testing of material was done by trained
5-4 Monticello Reactor Vessel, ~age 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,eee-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 ~th 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, ~ennsylvania, plant. The final drilling of the holes was performed in place at the site.
5-5 e 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 11 feet wide. Each ring was made from two formed plates. The half ring sections were temporarily welded toge~~cr and placed on a roller bed. The ring was preheated and the overlay weld metal deposited with automatic equipment similar to that shown e in Figure IV-s. All shell fittings were shop installed. Postweld heat 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. e 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.
5-6 e Monticello Reactor Vessel, Page IV-S The stud bolts, washers, and gaskets were shipped directly from General Electric qualified manufacturers to the jobsite storeroom. 4.3 SITE SUBASSEMBLY AND ERECTION Site subassembly of the" reactor vessel started about three months after work began on the containment vessel. Erection of the reactor bottom head followed the completion of the leak rate test of the drywell. The bottom head and stub skirt was welded to the reactor support skirt which was attached to the drywell prior to the leak rate test. Unlike the case for determining the maximum size of e",;l subassemblies at the shop, weight of the lifts or derrick capacity dictated the subassemblies that could be made at the site. The closure seams between subassembled sections were made in place. The postweld heat treatment zones were established by the location of penetrations with respect to circumferential weld joints and the adherence to safe thermal gradients through adjacent vessel materials. Methods of achieving the machined surface requirements, drilling and tapping and boring operations were developed by CB&I engineers using commercial equipment, where available, and designing and building custom-made devices, where necessary.
5-7 Monticello Reactor Vessel, Page IV-6 Suitable weather protection devices were provided to shelter the vessel weldments during ground assembly, welding, and postweld heat treatment. The postweld heat treatment furnaces were also used for environmental housings for the welding and radiographic work. Figure IV-7 shows typical postweld heat trea~~ent of the longitudinal ring welds. Figures IV-8 and IV-9 similarly show postweld heat treatment of the bottom head and stub skirt assembly .a and the top head and flange assembly. Temperatures from thermocouples were permanently recorded on multiple point potentiometer instrument. Adequate thermocouples were used to obtain representative readings from all parts of the section being heated. The various parameters for heat treating, such as heating and cooling rate, variation of temperature during ~~e holding period, etc., were in accordance with Section III of the ASME Boiler and Pressure Vessel Code and other requirements of the General Electric specifications. 4.3.1 Site Subassembly The basic assembly yard fabrication process was performed as follows on the head and shell components: (a) shell halves joined . i n t o rings, bottom head to skirt extension, t~p head to flange on level work tables: (b) preheat to 300 0 F to 400 0 F and weld sections
5-8 Monticello Reactor Vessel, Page IV-7 together; i.e., four shell rings, one bottom head with skirt and one top head with flange: (c) magnetic particle check weld periodically during deposition of metal as preliminary inspection step and replace any unsound material found therein; (d) hot ultrasonic test welds before postweld heat treatment: (e) post weld heat treat at 11SOoF; (f) cool and radiograph welds; (g) ultrasonic welds again: (h) manual overlay welds: (i) postweld heat treat: (j) cool and ultrasonic overlay: (k) dye check overlay. 4.3.2 Assembly and Machining The bottom head and stub support'skirt assembly was set in place, leveled, and welded on a l7-foot diameter tubular support skirt furnished in the drywell base of the containment structure. The vessel centerline was established as a vertical line of sight using a precise jig transit instrument located below the bottom head and sighting on a target in the geometric center of the center control rod drive penetration. The leveling and plumbing procedure was repeated after placement of each of the four shell rings. The centerline for the bottom head and skirt assembly was located as shown in Figure IV-10. . The No. 1 shell ring, assembled and ~.,-j. '~ed in the asser..bly yard, was placed as an integral ring in position atop the bottom head. The girth seam was fit, preheated and hand welded. The No. 2 ring was then placed, fit and welded. The preheat was maintained on
5-9 Monticello Reactor Vessel, Page IV-8
~~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 ternporarj 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 s~~e as those performed in the shop. Radiography was performed utilizing a 75 to 100 Curie Gamma source with appropriate shielding. 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-l!4-inch diameter hold~own bolt holes, as shown in Figure IV-ll. With the cover in this same position, the grooves for the two 1!2-inch diameter stainless "0" 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 .the ring. A temporary cover was installed above this work area so ~~at balance of the vessel could be erected without interfering with
5-10 Monticello Reactor Vessel, Page IV-9
.~
the bottom head work. The holes and sleeves for the 121 6-inch di~~eter control rod drive thimbles and the 40 2-inch diameter holes for the in-core flux sensors were machined utilizing precision-oored guide templates, optically aligned in a temperature"controlled housing to guide a verticle boring bar and cutter head, as shown in Figur~ IV-13. These methods not only assured that the holes were on accurate centers but that they were plumb. The vessel closure flange was drilled and tapped in the assembly yard after i t was welded to the No.4 shell ring. The gasket sealing face on the vessel flange was machined in the assembly yard using the sarne equipment that was used for ~achining the top head flange. Drilling of the control rod drive sleeve holes, welding the sleeves and boring them to the final precision dimension was performed in parallel with the work on the vessel as described above. 4.3.3 Cleanina and Hydrostatic Test Upon completion of the machining work on the control rod drive sleeves, the reactor head was attached to the vessel in preparation for cleaning. The cleaning of the interior surfaces of the vessel was done using high pressure (approximately 8000 PSI) deionized water containing SaO-ppm by weig~~ of TSP. Special care was taken to thoroughly water-blast rinse all areas and crevices to
5-11 Monticello Reactor Vessel, Page IV-10 insure complete removal of the TSP solution. The rinsing continued until the effluent conductivity was 5 micro-mho/em. Upon completion of the initial cleaning, the vessel was filled with heated deionized water and tested per the requirements of t..'1e 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 "0" ring seals at the design pressure. Upon completion of the hydrostatic test at design pressure, .i the test caps were removed from the vessel and replaced wit..'1 temporary covers. The vessel was once again nigh pressure blasted th deionized water. After drying the interior surfaces of the vessel, the vessel was sealed to prevent entry of dirt or other foreign materials. 404 REACTOR VESSEL QUALITY CONTROL 4.4.1 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.
5-12 Monticello Reactor Vessel, Page IV-ll 4.4.2 Project Quality Control Oraanization Authority lines for project management and project quality control were separated by having both managers report directly to the Regional Operations Manager, who, in turn, reported to the Vice President and Manager of Operations. Company standards and policies for quality control -- or more aptly, quality assurance were" set by the Quality Control Administrator, who also reported to ~~e Vice President and Manager of Operations. The latter was on the same level as the Vice President and Manager of Welding and Inspection. The Q.C. Coordinator for manufacturing was concerned with only nuclear reactors. The Q.C. Coordinator for engineering, purchasing, and construction was concerned with ~nly this vessel from the date of contract until vessel completion. 4.4.3 Compliance with Specifications By using check-off type records, spot Checking operations as the work progressed, and by auditing all inspections, the plant and site Q.C. Coordinators were able to assure that:
- 1. Approved procedures were used;
- 2. The approved procedures were being followed:
- 3. Required inspections were properly performed:
- 4. Inspections were witnessed by the customer's Q.C. representative; and
- 5. The material or part met the required level of quality before it was further processed.
5-13 Monticello Reactor VesseL, 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 underwent. Each operation or inspection was given a unique reference number so that it could be referenced to ~~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 obtained. 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 Q~C~ 4Itepresentative after each item or piece was reviewed and accepted. 404.4 Documents and Records In addition to the usual records required for presssure vessels built to Section III of the ASME 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, J 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 4iJsed for similar operations regardless of whether performed in the shop or at the site. Traveler Cards, Thermal History, and Spread
5-14 Monticello Reactor Vessel, Page IV-13 Sheets were initiated in the shop and were carried through to ~~e completion of the job.
5-15
- CHICAGO BRIDGE & IRON COMPANY
'60 SANSOME STREET. SAN FRANCISCO. CAI..IFORNIA 94104 April 1, 196~ A,.ea Cede: 4'5981-7530 In Quintuplicate General Electric Company !-1onticello P:::-oject Atomic **Power Equipment Department P.o. 205-55582-I Nuclear Energy Division Reactor Vessel 175 CUrtner Avenue Contract 9-5624 San Jose, California 95125 Seq. No *. SFC-259 Attention: Mr. B. K. Lloyd, Buyer Mail Code 522 Re: Vessel Fabrication and Assembly Report Gentlemen:
Following our discussions in your office on March 25, . .* 1969, we have once again reviewed Section IV of ~~e Report prepared for the AEC and issued in November 1966 under the title "Honticello Nuclear Generating Plant - Design, Fabrication and Erection of the Reactor Vessel." Accordingly, we have marked the approp:::-iate technical changes to indicate the revisions made during fabrication and erection of the vessel. Our attached sheets, marked Attachment A, dated April 1, 1969, describe the technical changes made in Section IV. This Attachment could be modified and issued as an Erratum to the original Report. As for the submittal of different photographs, we ass~~e that you can review the photographs that have been furnished you in accordance with our terms of the contract and choose those pic~ures that best depict the actual work done at the jobsite. Wi th this transmittal, we assume the Vessel Fabrication and Assembly Report is complete as far as Chicago Bridge & Iron Company is concerned. Very truly yours, CHICAGO BRIDGE & IRON COMP~j[ RCB:aer . Robert C. Baker Enclosure Contracting Engineer 9-5624
5-16 CHICAGO BRIDGE & IRON COMPANY MONTICELLO REACTOR VESSEL April 1, 1969 ATT ACHMENT A SECTION PAGE COML'1ENT 4.2 IV-3 First Paragraph: Revise to reflect that all shell fittings were shop .installed. 4.3 IV-3 First Paragraph: Site subassembly began about three months a=ter work started on the containment. Erection the reactor bottom head followed ~~e completion 0= of the leak rate test of the containment vessel. 4.3.1 IV-4 First Paragraph: Revise Item (f) to read "cool and ultrasonic welds II. and Item (.g) to read IIradiograph welds. II 4.3.2 IV-5 First Paragraph, Ninth Line: Delete reference to the shroud support skirt. 4.3.2 IV-5 Second paragraph: Revise to indicate that the radiography work was done with a 75 to 100 CUrie Gamma source. tv-5 Last Paragraph: Revise to indicate that the vessel closure flange was drilled and tapped in the asse~ly yard after being welded to the No. 4 shell ri~g. 4.3.2 IV-6 First Paragraph: Revise to read that the gasket sealing face o~ the vessel flange was machined in the assembly yard using the same equipment as was used for maChining the top head. 4.3.2 IV-6 Second Paragraph: Delete this paragraph. 4.3.3 IV-6 First Paragraph: Revise the Section to indicate that, after completion of work and placement of the reactor head on the vessel, the interior surfaces of the vessel were~cleaned usi~g high-pres sure_ -
5-17
- ATTACHHENT A (cant Id)
CIUCACO DnlJ)CE & InON COMJ.>ANY Page: 2 April 1, 1S169 SECTION ~ COMMENT deionized water containing 500 parts per million by weight of TSP. Special c~re was take~ to thoroughly waterblast rinse all areas and crevices to insure complete ra~oval of t~e TSP solution. The rinsing continued until the conductivity of the effluent was measured at 5 micro-mho/em. Following the Code hydrotest, the vessel was once again high-pressure blasted with. deionized water. 4.4.1 IV-6 First Paragraph: Add the comment that, in addition to the control rod penetrations, the instrument nozzles in t.:"le third and fourth ring were partial penetration weld connections. These partial penetrat.ion welds used details per Figure N~462.4(d) in Section III of the 1965 ASME Code. 4.4.2 IV-7 Second Paragraph: Delete the note in the parenthesis. All nozzles were installed in the shop. 4.5.8 IV-20 First Paragraph: In the second line, delete "CB&I Quality Control. Coordinators will maintain a Daily Progress Recond." 4.5.8 _ IV-21 First Paragraph: Delete this paragraph. 4.5.8 IV-21 Fourth Paragraph: Delete the first two sentences.
MONTICELLO 6-1 EXHIBIT 6 INDEPENDENT STRESS ANALYSIS REPORT
- REV 4 12i85
6-2 e:: 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 study has been performed by Teledyne Materials Research Division of the Teledyne Company, Waltham, Massachusetts. Teledyne's summary letter concerning their review is included herewith as Exhibit 6 of this report. e* e
6-3
- A TH!:DYNf rO."IP.ANY TUEDYNE MATtRlALS RESEARCH September 15, 1969 Project E-1113 General Electric Company Nuclear Energy Division 175 Curtner Avenue San Jose, California 95125
Subject:
GE PO #205-F0144 Audit of Monticello Vessel Design Analysis Attention: Mr. D. K. Reising Gentlemen: Teledyne Materials Research has completed the audit of the stress analysis report of the Monticello - NSP Reactor Vessel. On the basis of our review of the final report, we are of the opinion that:
- 1) The analytical methods employed by General Electric Co. and Chicago Bridge and Iron Co. are consistent with the state-of-the-art as generally practiced in the indus try.
- 2) The ASME code interpretations employed with respect to the analytical results are proper.
Very truly yours, TELEDYNE MATERIALS RESEARCR v.e~~ William E. Cooper Vice President WEC/mef
- ENGINEeRS .ANO METAllURGISrS , FORMERLY LESHLlS AND ASSOCIATES. I~C. & *'1EW E~CL,\"U MATeRIALS LABORATORY, INC
e* MONTICELLO 7-1 EXHIBIT 7
- e REACTOR VESSEL DESIGN SPECIFICATION (REPAIRS)
REV 5 12/86
7-2 REVISION SiATUS SHEET GEt4ERAL@ELECTRIC 22A5541 CONT ON $.tEET 2 SH Nt', 1
~;IJCLEAR ENERGY DIVISION DOCUMENTTITlE ______R~EA~C~T~O~R~V~E~5~SE~L~___________________________________________
Cl SPECIFICATION 0 DkA~G 0 OT~ER - - - - - - - - - TnE D~SIG~I SP:CIF!CATI(j~j (REP;"IR) FMF N/A LEGUIO OR DESCAlmOI Of GROUPS
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7-3 GENERAL 0 ELE~TRIC NUCLEAR ENERGY DIVISION 22A554l SM. HO. 2 REV. Z
- 1. SCOPE 1.1 This specification gives the functional and engineering requirements for feed-water nozzle and safe end repair. The repair con$ists of removing cladding from the nozzle blend radius and bore, machining the safe end to accept a feedwater sparger interference fit thennal sleeve with a piston ring seal, arid removal of any remaining lin2ar liquid penetrant indications.
1.2 This specification replaces the original Reactor Prp~sure Vessel Design Specification for the Reactor Pressure Vessel Feedwater Ilotzles and Safe Ends.
- 2. APPLICABLE DOCUMENTS 2.1 G~neral Electric Documents. The following documents form a part of this specification to the extent specified herein.
2.1.1 Supporting Documents 11201696
- c. Cleaning and Cieanliness Control for Assembly of Reactor Components llAZ045 do Reactor Vessel Modification 769E367
- e. Vessel Feedwater Nozzle Blend Radii Crack Removal Tooling '22A4705
- f. Thermal Sleeve End 11201693 2.1.2 Supple~nta\y Documents. Hr.ne
7-4 GENERAL Q ELECTRIC 22A554l - SH. NO. 3 NUCLEAR ENERGY DIVISION REV. 2 2.2 Codes and Standards. The following codes and standards form a part of this spec:ifi cation to the c!xtent specifi ed herei ~
- a. American Society of Mechanical Engineers (AS:~E) Boiler and Pressure Vessel Code
~(l) Section Xl, Inservice Inspection of Nuclear Reactor Cooling Systems, 19;4 Edition with Addenda to and including the Summer 1975 Addenda. ~(2) Section III, Nuclear Power Plant Co~onents, 1974 Edition with Addenda to and including the Sf.liT':ITIer 1976 Addenda.
(3) Code Case 1804.
- 3. DESCRIPTION 3.1 The repairwi11 mini-mize damage to the fe:l!dwater no"!zle due to thennal cycling.
This repair will be perfo~d in accordance with Section XI of the ASHE Boiler and Pressure Vessel Code.
- 4. REQUIREMENTS
.;..~ Cw..! ... 7::~ .;;!~~. ~~~'!".~:-: -~-:-~---- ... "r' ~,.,.~~!'~",.o ~j.,.;, Tnp I""Pouirement.s or Art1c~WA-4000 and IWB-4000 of Section X! of the ASME Code. 4.2 Functi ona 1 4.2.1 The machined safe end shall be compatible with interfacing thermal sleeve shown on Drawings 11201693 a~~ 11201696. 4.2.2 Clad removal a~d the safe end mac!lining shall be compatible with the generic feedwater nozzle inside machinin~ speci fied on Dra\'iing 767E~04 and with Drawing 769E361, and shall be performed in ar.r.ordance with SpeCification 22A4705. 4.2.3 All work shall be performed in accordance with Specification 21A2045. 4.Z.4 Repair of Linear 1ndicat10ns. If any linear indications are detected after machining ;s completed, they shall be removed in the following manner. 4.2.4.1 Remove !ll unacceptable indications by grinding. 4.2.4.~ After the unacceptable indications have been removed. the sides of the cavity shall be ground to merge s'~othly with adjacent surfaces. In the hoop direction. t~e sides shall be merged with ~ minimum blend slope of 4:1 (width to depth). In the axial direction, the sides sha'il be merged with a minimun blenci slope of 2:1 (width to depth). The ~avity shall be round bottomed with a minimum radius of two times the depth of the material removed from the final machined surfaces.
7-5
- b Er~E RA LV ELEe TRIC NUCL~R ENER~Y :lIVISION 22A5541 SM, NO. 4 REV. 2 4.3 Design 4.3.1 Thema1 Sl~eve Reactions. The Incone1 thermal slee ... e 3hown on Drawing 11201693 will be ins~o11ed with a coid nominal interference of 0.010 inch across the diameter. The effects of the thermal sleeve on the safe end and nozzle shall be con-sidered in the desigrl ariii1ysis. The geometry is shown in Figure 2 and on Drawing 769E367.
4.3.2 Design pressure \~ 1250'psig. Normal o~erating pressure is 1111 psig. 4.3.3 Design temp~rature is 575°F. Normal operating temperature is 546°F.
~ 4.3.4 Normal operating cond:tion pipe reaction loads are !ihown in Figure 1. There are no upset, emerge~cy. or fault pipe loads specified for this Fe 3.0 kips 19.26 FL 5.7 kips .F-:-. 3.2 Idps 1-'
Me 156.0 in.-kips M... 336.0 fn.-kips Hz 348.0 in. -ki ps 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 Al'~ance. All exposed exterior ferritic steel surfaces of pres-sure containing parts snall have a corrosion 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 tn 40 years~ 4.3.7 Desiqn Life. The design life of this repair shall be not less than 24 months. If design life is extended beyond 24 monU,s, then additional analysis,according to this specification, is required.
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- G~ .4 Efi A L @ ELEe TRIC HUC~EAR ENERGY DIVISION I
I REV. 22A5541 Z . SH. No.6 7-7 4.4 Opert'.t~~ma1 Environment 4.4.1 Ne~tron fluence is ne~~igibje at th~ feedwater nozzle. 4.4.2 It shall be assumed that thE' interior of the nozzlE'! and safe end ~'lterior of the nozzle and safe end are exposed to saturated stE'lJI1 and demfneral ized water ynder operating conditions. 4.4.3 Insulation. Exterior surfaces of the noz~le and safe end are insulatad. The average heat transfer rate ~nder operating conditions is 80 Btu/hr-ft~. 4.4.4 Heat Transfer Coefficients. The heat transfer coefficients defined ce10w are from emperlcal data for thlS deslgn and are to be used in the analyses r~qu;red in Sectien 5. :ieat transfer coefficients for other locations shall be calcullted by conventional methods. 4.4.4.1 The heat transfer coefficient for the nozzle inside surface (areas A thru o in Figure 3) for all leakage flow rates is:
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4.4.4.2 The heat transfer coefficient for the inside of the safe end and thermal sleeve that is exposed to the feeaoVate." flow is: Z @ Feedwater temoeratul~ Q ~8
;t..{mr, 26 0 Btu h
- Z@1000 F x O ~h-r-~f~t~Z_~b~F 4.4.4.3 Nomenclature K pl/3 r
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- Prandtl Number v
- Kinematic viscosity Q
- Feedwater flow per nozzle (gpm) o.R
- Feedwater flow per nozzle at 100: rated power (gpm) as defined in Paragraph 4.5.1.3
- GEU ERAl@ELECTRiC NUCLEAR ENFRGY DIVISION !'lEV 222AS541-~H
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7-9 GEN ERAL~* ELECTRIC ~SS41 SM. NO.8 NtJCLEAR ENe. ~GY DIVISION I REV. 2 4.4.4.4 the -ht;~t t;"ansfer coeff::ient for the inside of the vessel si.el1 outside of areas A through D ~hall be 1000 Btu/hr-ft 2 -OF for all conditions. 4.4.5 Annulus Fluid Temoerature where T D annulus fluid temperature TFW = feedwater fluid temperature TA = Region A fluid temperature = 546°F Cl D Coefffcient from Table 1 C Coeffi ci ~nt from Fi gu re 4 2 D Tab1e 1 Ccefficient C1 I
----------------------~-------i ~~eFlow . Figure 3 A B C 0 100% rated feedwater flow 0.44 0.59 0.72 0.88 20~ rated feeewater f1 ow 0.66 0.88' 0.96 0.96 0% rated feeawater flow 1.0 1.0 1.0 1.0 Interpolate line!rly between defined points A. Bt C. and 0 and between flow rates given
- 7-10 GENERAL 0 N~ JCLEAR E'~ERGY ELECTRIC DIVISION IREV. ;2ASS41 SH.NO.9 1.0 0.9 0.8 N
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7-11
,"~
GtHE ~ A L ~!i$ ELEe TRIC 22A5541 SH. NO. 10 NUCLEAR DIVISION EN~RGY HEV. 2 A 4.5 Cyclic Conditions. There ar.! three sources of nomal ,perat-;on themal cyc1es *
. system CyCllng, t;m;taole fla.: cycl ing, a"lr. rapid r"ixir.g cycling. ihere are no upset, emergency, or f~lult thermal cycles specified for tni~ design.
4.5.1 System C.. ~. This type cycling results from changes in the .flow and temperature of tne-Tieedwater and/or of the reactor w~ter. 4.5.1.1 Seventy-five cyeles of the following transient represent the equivalent of 24 months of this type of C,Ycling. 4.5.1.2 The temoerature transient consists of: 4.5. 1.2.1 Io;ti~l Condition. The nozzle, safe end, thermal sleeve and all contained water is isothermal at 100°F and is at 0.0 psig. 4.5.1.2.2 The flozzle. safe end. and themal sleeve are heated by the contained water. The wat!~r is heated from 100°F to 54~oF at a rate of 100°F/hr. The pressure is inc~eased to 1111 psig. 4.5.1.2.3 The hot feedwate~' 1s displaced by 100°F feecWater with a velocity of 5 ft/sec. This condition exists until steady state is achieved. 4.5.1.2.5 The fe~dwater temperature then is increased to 376°F at 250°F/hr. Simultaneous.ly the feedwater flow velocity is increased from 5 ft/sec to 20 ft/sec. The end points, 376°F and 20 ft/sec are reached simultaneously. This condition exists until steady state is achieved. 4.5.1.3 Feedwater flow rate shall be obtained from feedwater velocity by using an area of 64.5 square inches. The velocity that corresponds to 100~ rated feed-water flow is 20 ft/sec at a temperature of 375°F
- GENERAL 0 ELECTRIC 22A5541 SH. No.ll 7-12 e\
NUCL!A'. E!"ERGY DIVISION REV. 2 4.5.1.4 Thd temp~rature trans1ent is shown in tabular form below: TEMPERAn.;RE TRANSIENT IFl ui d ITemp. Rate Fluid Start Temp. Fluid End iemp. State of fluid Fluid Velocity Vessel Pressure Notes I 10e F/hr C ,?50 F/hr 100 100 260 546 100 376 Water Water Water a 5 ft/sec 5* it/sec 1111 psi 9 1111 psig 1111psig 1Followed Followed by Step To by Step To
- Velocity changes linearly 5 ft/sec to 20 ft/sec 4.5.2 Unstable Flow eyel ing. During reactor startup under low power condi tions temperatures 1n the top half 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 F1gurp. 5. for 1: of the operating time, i.e ** 88 hours 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
~"Q'~C:I
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. . . . . . . . . ."... ~
- "'". . .' . . . . .,._ .J""' ....
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,,~
.. _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 boundar,y conditions for the top half of the nozzle. l'he 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 form E a (TToo - TBottom) Z 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"ns1on TTop* mean temperature of top half of safe end TBottom- mean temperature of bottom half of safe end
7-13 GENERAL() ELECTRli: 22A5541 iH. HQ. 12 NUCLEAR ENERGY DIV' llON RE'/, __, ___===----~
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C I-so 120 1111 300 360 TlJC (S£C) FIGURE 5
7-14
- GENE RAL(i) EL t CTRIC 22A5S41 SH. NO. 13 NUC'.LEAR E:NERGYDIVISION REV. 2
~ .5.3 Rapid Cycling 4.5.~.1 Rapid temperature cycling (en the order of 0.1 Hz to 1.0 Hz) occurs as a result of cold feedwater beir.g injected into a hot reactor. The most dominant cause of this cycling in the nozzle bore and on the blend radius is turbulent mixing ~, leakage flow with region A f1~id. Rapid cycling is caused iri the Absence tf lnkage flow by turbulent region A fl uid causing the thermal bounciary 1ayer around the cold thermal sleeve to be broken up and swe~~ ~~~inst the nozzle. Incomplete1y mixed sparger discharge flow and region A fluid that is carried back to the nozzle also causes some rapid cycling. . 4.5.3.2 The metal ~urface temperature ranges are given by the following em~1rical equation: where: t."'"p_p
- metal :aurface peak to peak temperature range A
- amplitude coefficient for a given frequency of cycling, from Table 2
- coefficient from Table 3
"'4
............ _ ............ ~. * ~:r-{~
C'.J TFW and TA are defined 1n Table 4. 4.5.3.3 The ampl itudes and cycles given in Table 2 and the f1ow/temp~rature/time data from Table 4 are to be used in the fatigue evaluation. (The design life is given in Paragraph 4.3.7.) Table 2 Amplitude/Frequency Data for Rapid Cycling I
!ndex Pmp1itude Frequenc;.v I A I.:ycles/hr 1 1.00 15 2 0.95 30 3 0.9(; 30 4 0.85 75 5 0.77 120 6 0.66 150 7 0.56 180 8 0.46 225
.. 9 0.36 375 10 0.26 375 11 0.15 1125
_L
7-15 Gfr~ {R~~-L G ELE~iR I r 22A5541 SH.NO. 14 NUCLEAR ENERGY DIVISION REV. 2 Table 3 Coefficient C3 100% Rated 20% Rated
- I 0% Rated Pt. Feecjo.late:- Fi I)W Fecdwater Fl\i~ Feeciwater Flew l1J 0.20 0.12 0.12 0.20 0.12 0.12 I
0.30 0.18 0.18 0.10 0.06 0.06 Interpol~te linearly between defined pOints A. 8. C. and 0 and beo'ieen given flow rates. Table 4 Flow. Temperature. and Time ~ata for Rapid Cycling "FW TA Feedwater Feedwater Region ,~ Hours j Index J I Flow
~ Rat~!! .
I Temperature of I Tempera tUI~ of i iime III Per Year 1 1 100 367 546 61.33 5373 2 84 358 546 14.00 1 1226 3 57 330 546 7.47 654 4 37 300 546 3.73 327 5 10 195 546 3.73 327 6 20 101 546 0.47 41 7 10., 101 546 0093 81 8 131 546 0.47 41 9 2 70 546 0.09 7.9 10 84 278 546 0.19 16 11 100 315 546 0.93 81 12 100 27a 546 0.00 0.0 11 0 200 . 200 0.067 5.88 14 0 300 300 0.15 13.38 15 0 400 400 0.24 20.88 16 0 ~O 340 0.999 8705 17 1 350 360 0.003 0.25 18 2 190 350 0.020 1.78 19 2 125 340 0.016 1.38 20 2 70 330 0.003 0.25
. 21 2 190 400 0.018 1.60 22 3 200 340 0.004 0.38 23 0 70 70 5.14 450
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Ul FIGURE 6
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....I 0\
7-17 G~H-i~ALO ELECTR Ie 22A554l SH. NO. 16 NUCLEAR Ef/EqaV DIVISION REV. 2 4.5.3.4 The alterna:ing stress produced by the rapid cycling shall be calculatelJ using: where: E (Youngs ~~dulus) and a (instantaneous co~fficient of thermal expansion) are evaluated at a temperature of Y = Pcisson's Ratio ATp_p ' TA, C3, C4 , and TF~; are defined in Paragraph 4.5.3.2 4.5.3.5 The fatigue usage factor d:.!e tQ rapid thermal cycling is given by:
,,)
II U * ~ UiJ" ~ Design Life
, 1-1 j-l where U
- usage factor due to rapid cycling ulJ..
- usage factor due to ith amplitude and frequency for the jth flow.
temperature. and time 4.5.4 Leakage flow rates are to be calculated for all conditions. The following assumptions are to be used:
- a. Neglect the pressure of any seal r1ngs, springs. and ring grooves. i.e **
assume the thermal sleeve looks like Figure 3.
- b. There is zero leJkage flow when there is zero clearance between the thermal sleeve and nozzle.
- c. The pressure drop across the thermal sleele is 10.9 psi at 100~ rated feedwater flc/'-I.
7-18 GENERAL { ) ELeCTRIC [ 22A5541 SH.NO. 17 NUCLEAR ENERGY DIVISION REV. 2 4.5.4 {Continued}
- d. Yielding of the therma1 sleeve and s"fe end (and thus relaxing the initial interference fit) at this jeeka9~ gap shall be considered.
- e. Changes in the le~kage gap due to differential expansions between the nozzle and thermal sleeve must be considered. See Figures 2 and l for dimensions and materials for determining 1ea~age gap.
- f. The leakage gap (i.e., radial gap) increases at the rate of 0.0017 inch per year due to corrosion.
- g. The leakage flow velocity averaged over the annulus area at the discrete point of interest shall be used in determining C? and ~~ from Figures 4 and 6 except for zoroes C and D. Use the maximum average leakage velocity in ,one C to deter-mine C2 and C4 and use these values for all of zone C. Assume that the leakage velocity varies from the zone C value to zero at point D.
- 5. ANAlYSIS 5.1. Primary ~tr~s::~s. The f~n;l ;r.:)::h~:-:c~ ~nd 9~!!::d nczzl~ ~.,d s~f!! e!1~ t.*"f:r.!!:~se~
- a.,et. aCl ... ~ * .1 WI .. til .....ni'-.1 S\.IIC ... ., ... .,_i.~~ .._.. ~~'.;~ :.::~ :...!_ : __ :.: . . .: ::: * ....
~ :!c~:*
HB-lOCO.
- 5.2 Secondary and Peak Stresses. The nozz1* and safe end shall separately be sh~wn
. to satisfy the seconaary and peak stress requin:ments of .A.s~lE Code Section III. Article NS-3000. The fatigue curve shown on Figure 7 shall be used. The operating pressure and temperature identified in Paragraphs 4.l.2 and 4.3.3 shall be used.
- 6. DOCUHENTATION 6.1 The required ana~ysis shall be documented in a manner suitabl~ for submission to enforcement and reguldtory agencies.
6.2 The required analysis shall be certified.
* ~G-E r~ 'E-R-j\ L~ ELE CiA I C NUCLEAR Er~ERGY DIVISION REV.
22A5S"1 2 Final SH. NO. 10 1"1111111"1""1""1'1-106
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10* 101 102 103 10' 105 ' 106 10 7 108 109 1010 1011 10 12 CYCLES E
- 3Oxl0 6 psi FIGURE 7
- MONTICELLO 8-1 EXHIBIT 8 REACTOR VESSEL SYSTEM CYCLING (STRESS REPORT)
REV 5 1.2/86
EIS IDENT: RV (SYSTEM CYCLING) REVISION STATUS SHEET GENERAL fJ ELECTRIC
~:'JCLEAR ENERGY BUSINESS GROUP DOCUMENTTITLE _______ RE_A_c_T_o_R__VE
__s_s_E_L__<_sY_s_T_E_M__CY __C_L_IN_G_)____________________~________________ TYPE~S~T~RE~S~S~RE~P~O~R~T______________ LEGEND OR DESCRIPTION OF GROUPS FMF~M~ON~T~I~C~E~L~L~O~_______________ MPL ITEM NO. (S) PRODtiCT SUMM.,!.RY (SECTION 7)
- DENOTES CHANGE IMPORTANT TO SAFETY. THIS ITE~A IS OR CONTAINS A SAFETY RELATED ITEIA YES 0 NO ~ EQUIP. CLASS. CODE REVISIONS j/bh o DMH-\ 1"1'2.
J/rr 1 NH14523 CHK BY:£4Ht~ L AMARAL PS 11C; PRINTS TO LOCATION S&~ JOSE 2 1 R ..
?
- B-3 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH No.2
. BUSINESS GROUP "EV 1 CERTIFICATION OF STRESS REPORT This certification for the Monticello Reactor Vessel (System Cycling) feedwater nozzle and safe end repair Stress Report and accompanying documents comprises the Stress Analysis required by Paragraph NCA-3SS0 of the ASHE Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components.
1977 Edition with Addenda through Summer 1978. I certify that to the best of my knowledge and belief the Stress Analysis Report is correct and complete and in accordance with Design Specification 22A6996, Revision 0, and in compliance with the requirements of Article NB-3000 of the ASME Boiler and Pressure Vessel Code. Section III, Nuclear Power Plant Components, 1977 Edition with Addenda through Summer 1978. I hereby certify that this report was prepared by me or under my direct supervision and that I am a duly Registered Professional Engineer under the laws of the State of Minnesota
- Document Revision Type of Document Title Number Number Stress Report Reactor Vessel 22A7227 0 Rapid Cycling Design Spec Reactor Vessel 22A6996 0 System Cycling Date: _q-,--_~~;;..;;.:"~__/....:..'1~f__z.._
- NEBG-a07A (6/10)
8-4 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH No.3 BUSINESS GROUP REV 1 TABLE OF CONTENTS
- 1. ABSTRACT
- 2. SUXMARY AND CONCLUSIONS
- 3. DESIGN REQUIRDIENTS
- 4. ANALYSIS 4.1 Thermal Transient Analysis 4.1.1 Thermal Model 4.1.2 Feedyater Nozzle Heat Transfer Coefficients 4.1.2.1 Cool-Down Transient 4.1.2.2 Heat-Up Transient 4.1.2.3 Normal Operation 4.1.3 Feedyater Nozzle Annulus Fluid Temperatures 4.1.4 Thermal Analysis Results 4.2 Stress Analysis 4.2.1 Selected Loc.ations for Stress E.valuation 4.2.2 Thermal Stress Analysis 4.2.2.1 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 4.2.4.1 Pressure Stress Analysis Results 4.2.5 Total Primary Plus Secondary Stress Range 4.2.5.1 Thermal Stress Ranges 4.2.5.2 Nozzle End and Thermal Sleeve Load Stress Ranges 4.2.5.3 Pressure Stress Ranges Total P + Q Range 4.2.5.4 4.2.6 Interference Fit Stresses NEBo.a07 A (6/10)
8-5 NUCLEAR ENERGY GENERAle ELECTRIC 22A74S4 SH No.4 BUSINESS GROUP Plev 1 TABLE OF CONTENTS (Continued) 4.3 Fatigue Analysis 4.3.1 Stress Concentration Factors 4.3.2 Alternating Stress Range 4.3.3 Usage Calculation 4.3.4 High Cycle Fatigue 4.3.5 Accumulated Fatigue Usage 4.3.6 Total Fatigue Usage S. RESULTS 6* REFERENCES APPENDIX 10 LISTING OF 'NONO' APPENDIX 20 INTERGRANULAR SnESS CORROSION INDEX CALCULATIONS APPENDIX 30 RECALCULATIONS REQUIRED DUE TO MANUFACTURING DEVIATIONS
- NEBG-a07A (6/101
8-6 NUCLEAR ENERGY GENERAL., ELECTRIC 22A74S4 SH NO. S BUSINESS GROUP "ev 1
- 1. ABST.RACT This report documents the stress analysis performed for the feedwater nozzle and safe end assembly. The analysis is concerned with Service Level A, B,*and C events, and design conditions. A fatigue analysis was also performed.
This analysis of the feedwater nozzle and safe end assembly is required because of the complete redesign of the existing safe end and thermal sleeve assembly. As a consequettce of this redesign, the component's geometries will cha~ge from the ones originally analyzed, thus necessitating this report. The nozzle and safe end assembly in this report are analyzed in accordance with the requirements of the ASHE Code (Reference 6.2), and the General Electric design specification (Reference 6.1). NE&G-&07A (6/110)
- 8-7 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 6 BUSINESS OPERATIONS REV 1 I
- 2.
SUMMARY
AND CCNCLUSIONS 2.1- It is sh01f1l by aDalysis iD this report that the feedwater Doz.z.le aDd safe eDd assembly fully meet the stress :. ~.9\sity limits for all desigD" Service Le~el A. B. aDd C cODditioDs. Some of the sigDificaDt results of this aDa lysis are as follows: NOTE: These results iDclude the results of AppeDdix 30. Maximum desigD primary stress iDteDsity: (Table 3-2 aDd Table 30.3.1-1)
= 14.05 ksi ; Pm Allowable = 18.1 ksi PX+B = 24.38 ksi PX+B Allowabl e = 27.9 ksi Maximum Level 'c' primary stress iDteDsity: (Table 3-2 aDd Table 30.3.1-1)
P m
= 15.46 kli ; P m
Allowable = 2701 ksi PX+B = 33.89 ksi ; P X+B Allowable = 41.7 ksi Maximum raDge of primary plus secoDdary stress iDteDsity. P + Q: (Paragraph 4.~.S.4 aDd Paragraph 30.3.3.4.4) P + Q = 62.9 ksi MAximum raDge of primary plus secoDdary stress iDteDsity excludiDg thermal beDdiDg: (Paragraph 4.2.5.4 aDd Paragraph 30.3.3.4.4) P + Q = 45.94 ksi P + Q Allowable = 55.8 ksi Maximum total fatigue usage due to low aDd high cycle fatigue plus existiDg accumulated fatigue: (Paragraph 4.3.6 aDd Paragraph 30.3.4.2) U max
= 0.439
- "EO 107A (REV. 10/'"
8-8 GE N ERA L " 22A7454 NUCLEAR ENERGY BUSINESS GROUP ELEe TRIC PlEV 1 SH No.7
- 3. DESIGN REQUIREHENTS The safe end and thermal sleeve geometry is provided in References 6.8.
6.10. and 6.13. The nozzle geometry is provided in References 6.1 and 6.7. The operating thermal 'and mechanical loads are provided by Reference 6.1. This section illustrates accceptance for the design and Service Level C conditions. Primary mombrane and primary membrane and bending (sizing) calculations are performed. In Sections G through I, moments due to thermal sleeve axial loads are assumed negligible. NEBGoe07A (6,80)
8-9 GENERAL. ELECTRIC 22A74S4 S NUCLEAR ENERGY SH NO. BUSINESS GROUP REV 1 W I/ACAUON STEEL (SA-SO-S - CL.I) - Od&lnal Safe End 1\ \ \ 5c.umON sn:EL (SA-333 - GD.6) - Piping
&\ § ICARBON STEEL (SA-50S - CL.2) - Nonle V 7 L1CARBON STEEL (SA-3S0 - U'2) - Safe End I ISTAINLESS STEEL (SA-3SI - CF3) - Thermal Sleeve SCH. gO PIPINfr o ==
103 POla)T,s FoR. STA.£SS CRL.C:IJ.r..A7"',o~S L
- FIGURE 3.1 NOZZLE. SAFE END. AND THERMAL SLEEVE GBOME1'B.Y
8-10 NUCLEAR ENERGY GENE RAL* ELECl RIC 22A74S4 SH NO. 9 e" BUSINESS GROUP PlEV 1 TABLE 3-1 SECl'ION PROPERTIES FOR NOZZLE (Corrosion Included) Section Area Hodulus Thickness Section ( In.) ( In 2 )" ( In 3") Material A 0.531 19.03 Sl.98 SA-S08 (CL.2) B 0.531 19.03 51.98 SA-508 (CL.1) C 0 531 0 19.03 51.98 SA-50 8 (CL.l) D 0.531 19.03 Sl.98 SA-350 (U2) 0.S31 19.03 51.98 sA-3S0 (LF2) E
- F 0.4917 15.89 39.09 SA-3S0 (U2)
G 0.37S 10.46 22.32 SA-3S0 (U2) H 0.375" 10.46 22.32 SA-3S0 (U2) I 0.494 13.78 29.07 SA-3S1 (CF~)
] 0.494 13.78 29.07 SA-351 (Cl"3)
- Section A properties used "here (conservative)
=
e Area n/4 (D 2 _ D 2) o i Do = (Outside Diameter - Corrosion) rr/64 (D04 - Di4) Section I Modulus = C = D 12 Di = (Inside Diameter - Corrosion) o Corrosion Allowances (Reference 6.1) Exterior Exposed Carbon Steel 1/32 inch Interior Exposed Carbon Steel 1/16 inch Interior Exposed Stainless Steel 0.003 inch Material Allowables (Reference 6.2) Sm (at SSOOF) SA-S08 CL.l 18.1 kosi SA-S08 CL.2 26.7 ksi Carbon Steels 18.6 ksi SA-3S0 LF2 SA-333 GD.6 18.1 kosi Stainless Steels SA-3S1 CF3 16.0 ksi NEBGoa07A (6,80) e
22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 10 BUSINESS GROUP "EV 1 Loading Nozzle Safe End Loads (Reference 6.1) Forces in kips Moments in in-kips F F F II )I )I R(in) Condition ~ ..::I. ..! ~ ..::I. ..! Design 1.54 3.1S 2.28 387 .6 172.9 324.6 131.6 Nozzle Dead 'ft. - 0.11 - 0.63 0.15 11.6 - 14.1 - 11.1
- !: 158.9 :!: 313.4 131.6
'A' Seismic :!: 0.29 :!: 2.51 :!: 2.23 + 9.3 Loads 'Dlema1 0.02 0.16 - 0.21 12.0 - 12.1 45.0 Nozz~e Dead lit. - 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 in in-kips F F F M II M R(in) Condition J. ..::I. ..! ~ ..::I. -1 Design 2.5 0.6 S.7 1.4 2.0 0 103.0 Dead 'ft. 0 - 0.3 - 0.5 - 1.2 o -. 0 Seismic :!:. 2.S :!: 0.3 :!:. 1.S :!: 1.2 :!: 2.0 0 103.0 Thema1 0 0 - 1.2 0 0 0 Hydraulic 0 0 - 2.S 0 0 0
- NEBGoa07A (6/80)
8-12 NUCLEAR ENERGY, GENERAL. ELECTRIC 22A7454 SH NO. 11 BUSINESS GROUP REV 1 Emergency condition (Service Level 'C') defined in Reference 6.1 as follows: Normal operation plus two times Seismic Loads To calculate the largest nozzle loads, use the following: p = (F 2 + F 2)1/2
% Y M = (M 2 + M 2 + M 2)1/2
% Y z Nozzle 'A' Loading (Service Level ' C')
p = 5.70 kip M = 721.5 in-kip F z = 4.61 kip Nozzle 'B' Loading (Service Level ' C') p = 6.44 kip M = 789.3 in-kip F z = 0.56 kip Therefore, the following loads are used for the design and Service Level 'c' conditions: (Note: No faulted condition exists) Nozzle Loads Forces in kips Moments in in-kips Condition l M Design 4.05 534.4 2.28 Service Level 'c' 6.44 789.3 4.61 NEBGo-e07 A (6/80)
8-13 llA74S4 GENERAL .ELECTRIC 12 SH NO. NUCLEAR ENERGY BUSINESS GROUP REV 1 Thermal Sleeve Loads Forces in kips Moments in in-kips F Condition l 11 -' Des ign l.S7 3.1l4 5.7 Service Level 'c' 5.08 S .39 6.0 Pressure Loads (Reference 6.1) Design Pressure Service Level 'c' Pressure
- 1,375 psi Loading Sign Convention
- Sign Convention applies to both safe end and thermal sleeve loadings.
x
- NEBGoe07A (6/80)
8-14 22A7454 NUCLEAR ENERGY __ GENERALe ELECTRIC SH NO. 13 BUSINESS GROUP REV 1 Section A Design Pressure Stress: P D. = 1.250 (10,87S)
= 12.800
=
__ 1 ae psi 2t 2 (0.531) ae a~ = 2 = 6.400 psi a r
= -1.250 psi Stress Due To Nozzle Loads:
P = 4.05 kip M = 534.4 in-kip F = 2.28 kip z m = 534.4 + 4.05 (12.83) + 2.28 (0.56) = 587.64 in-kip
= H = 587.64 = 11.31 1:5i Z 51.98 F
= ~
A
= 2,28 19.03
= 0.12 ksi NEBGoa07A (6/10)
- 8-15 22A74S4 GENERAL" ELECTRIC 14 SM NO.
NUCLEAR ENERGY BUSINESS GROUP .. EV 1 Stress Due to Thermal Sleeve Loads: P = 2.57 kip H = 3.124 in-kip F z
= 5.7 kip m = 3.124 + 2.57 (15.76) + 5.7 (2.36) = 57.08 in-kip
= ! = 57.08 = 1.1 ksi Z 51.98 F
= ~ = S .7 = 0.30 ksi A 19.03 Total Stress
( a~ = 6,400 + 11.310 + 120 + 1,100 + 300 = 19,230 psi
= 12,800 psi U
e a = -1,250 psi r
- NEBG-a07A (6/80)
8-16 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 15 BUSINESS GROUP REV 1 Section A Service Level 'e' Pressure Stress:
= P Di 2t
.. 1,375 (10.875) 2 (0.531) .. 14,081 psi a e at ..
a 2 e .. 7,040 psi a .. -1,375 psi r Stress Due To Nozzle Loads: P M F z
=
6.44 kip 789.3 4.61 kip in-kip
.)
III = 789.3 + 6.44 <12.83) + 4.61 (0.56) .. 874.51 in-kip _ ! .. 874.51 = 16.83 ksi Z 51.98 F
= --! = 4.61 = 0.243 ksi A 19.03 NEBGoa07A (6/80)
8-17 GENERAL. ELECTRIC 22A74S4 NUCLEAR ENERGY SH NO. 16 BUSI NESS GROUP .. EV 1 Stress Due to Thermal Sleeve Loads: P = 5.08 kip M = 5.39 in-kip F z
= 6.0 kip M = 5.39 + 5.08 (15.76) + 6.0 (2.36) = 99.61 in-kip
= H = 99.61 = 1.92 ksi Z 51.98 F
= J = 6,0 = 0.316 ksi A 19.03 Total Stress a~ = 7,040 + 16,830 + 243 + 1,920 + 316 = 26.349 psi a
e = 14,081 psi a r
= -1,375 psi
- NEBGoa07 A (6/80)
8-18 NUCLEAR ENERGY GENERAL. ELECTRIC 2.2.A7454 SH NO. 17 BUSINESS GROUP REV 1 Section B Design Pressure Stress: P D. 1,250 (10.875)
--.! - a 12.800 psi 2.t 2. (0.531) ae a~ = 2. = 6.400 psi ar - -1.2.50 psi Stress Due To Nozzle Loads:
P = 4.05 kip M = 534.4 in-kip F = 2.2.8 kip z M = 534.4 + 4.05 (10.22.) + 2..28 (0.56) a 577.07 in-kip
= H _ 577.07 = 11.102 ksi Z 51.98 F
= --!
A
= 2.28 19.03
= 0.12 ksi NEBGoa07 A (6/80)
8-19 22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 18 BUSINESS GROUP "EV 1 Stress Due to Thermal Sleeve Loads: P = 2.57 kips H ... 3.124 in-tips F z
= 5.7 kip M ... 3.124 + 2.57 (18.37) + 5.7 (2.36) ... 63.79 in-tip
... 11 = 63.79 ... 1.23 ks i Z 51.98 F
... --! = S.7 = 0.30 ts i A 19.03 Total Stress a~ = 6.400 + 11,102 + 120 + 1.230 + 300 ... 19.152 psi a
e ... 12.800 psi a r
= -1.2S0 psi
- NEBGoa07A (6/80)
8-20 NUCLEAR ENERGY BUSINESS GROUP GENERAL. ELECTRIC
----------------------~--------~
22A7454 REV 1 SH NO. 19 e . Sec-tion B Service Level 'e' Pressure Stress:
= P Di 1,375 (10.875)
G e 2t = 2 (0.531) = 14,080.2 psi Gt = 2e = G 7,040 psi G = -1,375 psi r Stress Due To Nozzle Loads: P = 6.44 kip F z
= 4.61 kip e
M = 789.3 + 6.44 (10.22) + 4.61 (0.56) = 857.7 in-kip
= ! = 857,7 = 16.501 ksi Z 51.98 F
= -A A
= 4,61 19.03
= 0.243 ksi e.
NEBG-a07A(6/IO)
8-21 12A7454 NUCLEAR ENERGY GENERAL., ELECTRIC SM NO. 20 BUSINESS GROUP ~EV 1 Stress Due to Thermal Sleeve Loads: P = 5.08 kips H = 5.39 in-kips F z
= 6.0 kip H = 5.39 + 5.0S (lS.37) + 6.0 (1.36)
- 112.87 in-kip
= M = 112.S7 = 1.171 ksi Z 51.98.
F
= -! = 6,0 = 0.316 ks i all. A 19,03 Total Stress a~ = 7,040 + 16,501 + 243 + 2,172 +,316 = 16,272 psi CS e = 14,080 psi CS r
= -1,375 psi
- NEBG.a07A (6/10)
8-22 NUCLEAR ENERGY GENERALe ELECTRIC llA7454 SH NO. II BUSINESS GROUP REV 1 Section C/D Design Pressure Stress: _ _ P Di = 1.250 (10,875) ~ 12.800 PS1' va - lt '2. (0.531)
- a. = ~
2 = 6.400 psi a r
= -1.250 psi Stress Due To Nozzle Loads:
P = 4.05 kip M = 534.4 in-kip F = l.l8 kip z M = 534.4 + 4.05 (7.47) + l.28 (0.56) = 566.0 in-kip
= 11 = 566,0 = 10.89 ksi Z 51.98 F
= ~ = l,l8 = 0.12 kii A 19.03 NEBG0807A (6/80)
- 8-23 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC 1M NO. 22 BUSINESS GROUP JIIEV 1 Stress Due to Thermal Sleeve Loads:
P = 2.57 kips M = 3.124 in-kips F z
= 5.7 kip M = 3.124 + 2.57 (21.12) + 5.7 (2.36) a 70.86 in-kip H 1.2.tH .
= Z = 51.98 = 1.364 kS1
'F
= ~ = 5.7 = 0.30 ksi all. A 19.03 Total Stress a~ = 6,400 + 10,890 + 120 + 1,364 + 300 = 19,074 psi a
e = 12,800 psi a =-1,250 psi r
- NEBG-a07A (6,80)
8-24 NUCLEAR ENERGY _ GENERAL. ELECTRIC 22A7454 SH NO. 23 BUSINESS GROUP REV 1 Se~tion crD Service Level 'c' Pressure Stress: 1.375 (10.875) 2 (0.531)
= 14.080 ps i a
r
= -1.375 psi Stress Due To Nozzle Loads:
p = 6.44 kip M - 789.3 in-kip Fz = 4.61 kip M - 789.3 + 6.44 (7.47) + 4.61 (0.56) = 840 in-kip H 840
= Z = 51.98 = 16.16 ksi F
= .-! = 4.61 = 0.243 ksi A 19.03 NEBGra07A (6/10)
8-25 22A7454* NUCLEAR ENERGY GEN ERAL. ELECTRIC SH NO. 24 BUSINESS GROUP IIIEV 1 Stress Due to Thermal Sleeve Loads: p .. 5.08 kips H = 5.39 in-kips F .. 6.0 kip M .. 5.39 + 5.08 (21.12) + 6.0 (2.36) - 126.84 in-kip
= M =. 128,84 .. 2.44 ksi Z 51.98 F
.. -! = 6,0 = 0.316 ks i A 19,03 Total Stress
- at a
a
=
e = r 7,040 + 16,160 + 243 + 2,440 + 316 14,080 psi
-1,375 psi
.. 26,199 psi
- NEBG0807A (6/80)
8-26 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 25 BUSINESS GROUP REV 1 Section E Design Pressure Stress: P Di = 1,250 (10.87S) 2t 2 (0.531)
= 12,800 ps i CJ e
CJ~ - 2 = 6.400*psi CJ r
= -1,2S0 psi Stress Due To Nozzle Loads:
P = 4.05 kip M = 534.4 in-kip F z
= 2.28 kip K = 534.4 + 4.05 (4.72) + 2.28 (0.56) = 554.8 in-kip
= M Z
= 554,8 51.98
= 10.68 ksi F
= -!
A
= 2.28 19.03
= 0.12 ksi NEBGoa07A (6/80)
- 8-27 22A7454 GENERAL. ELECTRIC SH NO. 26 NUCLEAR ENERGY BUSINESS GROUP ~EV 1 Stress Due to Thermal Sleeve Loads:
p a 2.57 kip K ~ 3.124 in-kip F Z
= 5.7 kip H a 3.124 + 2.57 (23.87) + 5.7 (2.36) = 77.93 i:-kip a M = 77 .93 = 1.5 ksi Z 51.98 F
a AI * = A Z
= 1~:~3 = 0.30 ksi Total Stress
'. a~ = 6.400 + 10.680 + 120 + 1.500 + 300 = 19.000 psi a r
= -1.250 psi
- NEBG.a07A (6/80)
8-28 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 27 BUSINESS GROUP REV 1 Sec.tion E Service Level 'c' Pressure Stress: P D.
~ = 1,375 (10.875) = 14,080 psi 2t 2 <0.531)
C7~ = C7 2 e = 7.040 psi C7 r
= -1.375 psi Stress Due To Nozzle Loads:
P = 6.44 kip M = 789.3 in-kip Fz = 4.61 kip H = 789.3 + 6.44 (4.72) + 4.61 (0.56) = 822.28 in-kip
= -HZ =
822.28 = 15.82 ksi 51.98 F
= 4.61
= .-!
A 19.03
= 0.243 ksi NEBGoa07A (6110)
8-29 22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 28
- BUSINESS GROUP .. EV 1 Stress Due to Thermal Sleeve Loads:
P 0:: 5.08 kip x 0:: 5.39 in-kip F Z
= 6.0 kip M = 5.39 + 5.08 (23.87) + 6.0 (2.36) '" 140.81 i~kip 0:: ! = 140.81 '" 2.71 ksi Z 51.98 F
Z a AX * = A = 1::~3 = 0.316 ksi Total Stress a6 = 7,040 + 15.820 + 243 + 2,710 + 316 = 26,129 psi G e '" 14,080 psi a r
= -1,375 psi
- NEBGoaD7 A (6/1D)
8-30 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 29 BUSINESS GROUP REV 1 Section F Design Pressure Stress: P D. _ _ 1 _ 1,250 (9.794) .. 12.450 psi 2t 2 (0,4917) a r
= -1.250 psi Stress Due To Nozzle Loads: .
p = 4.05 kip H = 534.4 in-kip F z
=
H - 534.4 + 4.05 (1.62)
- 541 in-kip a BEND *
= ~ = 541,0
- 13.84 ksi Z 39.09 F
= -A A
= 2.28 15.89
= 0.144 ksi NEBGoa07 A (6/80)
- 8-31 22A7454 GEN ERAL., ELECTRIC SH NO. 30 NUCLEAR ENERGY BUSINESS GROUP "EV 1 Stress Due to Thermal Sleeve Loads:
P - 2.57 kip F 5.7 kip z I:
)f = 3.124 + 2.57 (26.97) + 5.7 (1.8) .. 82.7 in-kip
!! 82.7 -2.12ksi
= Z - 39.09 F
= -! = 5.7 = 0.36 ksi A 15.89 Total Stress at = 6.225 + 13,840 + 144 + 2,120 + 360 .. 22,689 psi a
e = 12.450 psi a = -1.250 psi r
- NESG.a07A (6/10)
8-32 22A74S4 NUCLEAR ENERGY GENERALe ELECTRIC SH NO. 31 BUSINESS GROUP REV 1 SeC'-t ion F Service Level 'C' Pressure Stress: P Di ... 1,375 (9.794) ... 13,694 psi 2t 2 (0.4917) cs6 = 2cs Q
= 6,847 psi C1 r
= -1,375 psi Stress Due To Nozzle Loads:
P ... 6.44 kip K = 789.3 in-kip F ... 4.61 kip z H - 789.3 + 6.44 (1.62) = 799.8 in-kip CSBEND *
= II = 799,8 ... 20.46 ksi Z 39.09 F
C1AX .
= -A A
= 4.61 15.89
= 0.291 ksi NEBG-807A (6/80)
8-33 22A7454 GENERAL. ELECTRIC SH NO. 32 NUCLEAR ENERGY BUSINESS GROUP ~EV 1 Stress Due to Thermal Sleeve Loads: P = 5.08 kips H = 5.39 in-kips Fz = 6.0 kip M = 5.39 + 5.08 (26.97) + 6.0 (1.8) = 153.2 in-kip
= MZ
= 153,2 39.09
= 3.92 ksi F
= ~ = 6.0 = 0.378 ksi A 15.89 Total Stress
.., a~ = 6,847 + 20,460 + 291 + 3.920 + 378 = 31,896 psi G e = 13.694 psi G r
= -1,375 psi
- NEBG-807A (6/10)
8-34 22A7454 NUCLEAR ENERGY GENERAL., ELECTRIC SH NO. 33 e* 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 ISmn ). Where: t p = Pipe nominal thickness Smp = Pipe allowable (Sm) S mn = Safe End Allowable (SIII ) For our geometry: tp = 0.5405 in. e';
'S mp = 1S.1 E3i t p (SIIl/ SIIll1) = 0.526 in S = 1S.6 Esi IIll1 SAFE END = 0.5S55 in.
'llIICKNESS Criteria Met NEBGoa07 A (6/801 e
8-35 2.2.A7454 GENERAL. ELECTRIC 34 NUCLEAR ENERGY SH NO. BUSINESS GROUP .. EV 1 Section G Deiign Pressure Stress:
/
P*D. 2.22* (8.505) aQ = --...! 2.t = 2. (0.375) II: 2.,518 psi ae a~ a Z = 1.2.59 psi 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 (23.87)'" 64.47 in-kip 64.47 a BEND * ... ! z = 22.32
... 2.89 ksi F
0... -.!. = S.7 = 0.545 ksi all. A 10.46
- 222 psi pressure assumed, twice normal operation
- NEBGoa07 A (6/10)
8-36 2.2.A7454 NUCLEAR ENER,GY GENERAL. ELECTRIC SH NO. 35 e' BUSINESS GROUP REV 1 Stress Due To Nozzle Loads: p = 4.05 kip x = 534.4 in-kip F Z
= 2..28 kip The exact &mount the safe end loads influence the thermal sleeve is unknoYn.
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 137,93 Thermal Sleeve I
= 362.33
= 0.381 Nozzle NOTE: Corrosion not included in calculation (more conservative)
M = 534.4 + 4.05 (4.72) + 2.28 (1.8)
= 0.381 .!z = (0.381)
= 557.62 in-kip 557,62 22.32 = 9.52 ksi
- F 2,28 au. = 0.381 ~
A
= <0.381) 10.46 = 0.083 ksi Total Stress
= 1,259 + 2,890 + 545 + 9,520 + 83 = 14,297 psi a
e = 2,518 psi ar * -22.2 psi NEBG0807A (6,80)
8-37 22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 36 BUSINESS GROUP REV 1 Section .G
~_.~., J.evel C' Pressure Stress:*
peD. 2t c:: 333* (8.505) 2 (0.375)
= 3.776 psi CJ~ =
CJ 2 e = 1,888 psi CJ r
= -333 psi Stress Due to Thermal Sleeve Loads:
P = 5.08 kip H = 5.39 in-kip
- Fz = 6 .* 0 kip H = 5.39 + 5.08 (23.87) = 126.65 in-kip
= M = 126.65 = 5.675 ksi CJ BEND
- Z 22.32 F
= J A
=
6.0 10.46
= 0.574 ksi
- 333 psi pressure assumed (conservative)
- NEBGoa07A (6/80)
8-38 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 37 BUSINESS GROUP REV 1 Stress Due To Nozzle Loads: p = 6.44 kip H = 789.3 in-kip F = 4.61 kip The exact amount the safe end loads influence the thermal aleeve is unknoYn. 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. IThermal Sleeve 137.93
= 362.33
= 0.381 I Nozzle NOTE: Corrosion not included in calculation (more conservative)
H = 789.3 + 6.44 (4.72) + 4.61 (1.8)
= 0.381 11z = (0.381)
= 828.0 in-kip 828,0 22,32 = 14.134 ksi
- F
= 4.61 a AX * = 0.381 --!
A (0.381) 10,46
= 0.168 ksi Total Stress
= 1,888 + 5,675 + 574 + 14,134 + 168 = 22,439 psi a
e = 3,776 psi ar = -333 psi NESG-a07A (6,10)
8-39 2.2.A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 38 BUSINESS GROUP .. EV 1 Section H Design Pressure Stress: 222- (8.50S) a e ..
- 2. <0.375)
= 2,518 psi CJ e
CJ~ .. 2 - 1,2.59 psi a r
= -2.2.2. psi Stress Due to Thermal Sleeve Loads:
P .,. 2.57 kip M .,. 3.124 in-kip ( Fz a 5.7 kip M a 3.124 + 2.57 (20.12.) = 54.84 in-kip CJ BEND *
.,. ~ = 54.84 a 2.46 ksi Z 22..32 F
.,. --! = 5.7 .,. 0.545 ks i CJ}J.. A 10.46
- 222 psi pressure assumed (conservative)
- NEBGoa07 A (6/80)
8-40 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 39 BUSINESS GROUP REV 1 Stress Due To No%%le Loads: p .. 4.05 kip M = 534.4 in-kip F .. 2.28 kip z The exact amount the safe end loads influence the thermal sleeve is unknoYn. 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. IThermal Sleeve 137,93
= 362.33
= 0.381 I 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
.. 0.381 z = (0.381)
B 22.32
.. 9.78 ksi F
.. 0.381 .-! .. <0.381) 2.28 = 0.083 ks i A 10.46 Total Stress a~ = 1,259 + 2,460 + 545 + 9,780 + 83 = 14,127 psi
= 2,518 psi a .. -222 psi r
NEBG-807A (6/aO)
- 8-41 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 40 BUSINESS GROUP "EV 1 Section B Service Level 'c' Pressure Stress:
333* (8.505) .. 3,776 psi 2 (0.375) a r
= -333 psi Stress Due to Thermal Sleeve Loads:
P = 5008 kip M = 5039 in-kip F z = 6.0 kip M = 5.39 + 5.08 (20.12) .. 107.6 in-kip C7BENDo
.. H Z
= 107.6 22 32 0
= 4.821 ksi F 6.0
= ~
A
=
10.46
= 0.574 ksi
- 333 psi pressure assumed (conservative)
- NEBGoa07A (6/10)
8-42 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 41 BUSINESS GROUP REV 1 Stress Due To Nozzle Loads: p = 6.44 kip H = 789.3 in-kip F Z
= 4.61 kip The exact amount the safe end loads influence the thermal sleeve is uninoYn.
RoYever, from previous analysis it has been determined that a conservative approach is to ratio the safe end loads to correspond yith the moments of inertia. IThermal Sleeve 137.93
= 362.33
= 0.381 I Nozzle NOTE: Corrosion not included in calculation (more conservative)
H = 789.3 + 6.44 (8.47) + 4.61 (l.S) (fBEND. = 0.381 H Z
= <0.381)
= 852.15 in-kip 852.15 22.32 = 14.55 ks i
- F
= 4.61 (fAX. = 0.381 -!
A (0.381) 10.46 = 0.168 ksi Total Stress (f~ = 1,888 + 4,821 + 574 + 14,550 + 168 = 22,001 psi
= 3,776 psi (fr = -333 psi NEBQ.a07A (6/80)
8-43 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SM NO. 42 BUSINESS GROUP REV 1 Section I Design Pressure Stress: 222- (8,386) ae = 2 (0,494)
= 1,884 psi ae a~ = 2 = 942 psi a
r
= -222 psi Stress Due to Thermal Sleeve Loads:
P = 2.57 kip M = 3.124 in-kip Fz = 5.7 kip H = 3.124 + 2.S7 (20.12) = 54.84 in-kip
= -M = 1.887 Z =
54.84 ksi a BEND
- 29.07 F 5,7 all. = --!.
A
= 13.78 = 0.414 ksi
- 222 psi pressure assumed (conservative)
- NEBG-807A (6/10)
8-44 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 43 BUSINESS GROUP REV 1 *
----------------------~~~----~
Stress Due To Nozzle Loads: p = 4.0S kip H = 534.4 in-kip F = 2.28 kip z The exact amount the safe end loads influence the thermal sleeve is unknown. 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 I = 362.33
= 0.381 Nozzle NOTE: . Corrosion not included in calculation (more conservative)
M = 534.4 + 4.05 (8.47) + '2.28 (1.8) = 572.81 in-kip a BEND * = 0.381 M z = (0.381) 572.81 29.07
= 7.~1 ksi F
2.28
= 0.381 ~
A
= (0.381) 13.78 = 0.063 ksi 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 GENERAL. ELECTRIC 44 SH NO. NUCLEAR ENERGY BUSINESS GROUP IUV 1 Section I Service ~vel 'e' Pressure Stress: p*n_i _ .. 333* (8,386) .. 2,826 psi C1 e .. 2t 2 (0,494) C1 .. -333 psi r Stress Due to Thermal Sleeve Loads: P .. 5.08 kip M .. 5.39 in-kip F .. 6.0 kip z
)( = 5.39 + 5.08 (20.12) .. 107.6 in-kip
= ! .. 107.6 .. 3.702 ksi Z 29.07 F 6,0
-A C1Al.. = A = 13.78
= 0.436 ksi
- 333 psi pressure assumed (conservative)
- NEBG-a07A (6/10)
8-46 NUCLEAR ENERGY GEN ERAL. ELECTRIC 22A74S4 SH NO. 4S 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 &mount the safe end loads influence the thermal sleeve is unknown. Bowever, 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 I = 362.33
= 0.381 Nozzle NOTE: Corrosion not included in calculation (more conservative)
H = 789.3 t 6.44 (8.47) + 4.61 (1.8) a BEND * = 0.381 z =
}! (0.381) 852.15 29.07 852.15 in-kip
.. 11.17 ks i F
4.61 a AX * .. 0.381 -! A
.. (0.381) 13.78
= 0.128 kosi Total Stress
= 1,413 + 3,702 + 436 + 11,170 + 128 = 16,849 psi a
r
= -333 psi NEBCoa07 A (6/80)
8-47 22A74S4 GENERAL. ELECTRIC NUCLEAR ENERGY SH NO. 46 BUSINESS GROUP PlEV 1 Section] Design Pressure Stress: peD. __ 1 .. 222- (8.386) ae = 2t 2 (0.494)
- 1,884 psi a
r
= -222 psi Stress Due to Thermal Sleeve Loads:
=
I. p 2.57 kip M .. 3.124 in-kip Fz .. S.7 kip M = 3.124 + 2.S7 (19.33)" S2.81 in-kip
= 11 = S2.81 .. 1.82 kai Z 29.07 F
aJJ... = ...! A = 5.7 13.78
= 0.414 ksi
- 222 psi pressure assumed (conservative)
- NEBG.a07A (6,10)
8-48 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 47 BUSINESS GROUP REV 1
----------------------~~------~e Stress Due To Nozzle Loads:
p = 4.05 kip H = 534.4 in-kip Fz "" 2.28 kip The exact &mount the safe end loads influence the thermal sleeve is unknown. 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. IThermal Sleeve 137.93 I . = 362.33
= 0.381 Nozzle NOTE: Corrosion not included in calculation (more conservative)
M = 534.4 + 4.05 (9.26) + 2.;8 (1.8) ... 576.01 in-kip eO' CJ BEND *
... 0.381 !z = (0.381) 576.01 29.07
= 7.55 ksi F
0.381 -.! A
... <0.381) 2.28 13.78
... 0.063 lsi Total Stress CJ~ = 942 + 1,820 + 414 + 7,550 + 63 ... 10,789 psi CJ e = 1,884 psi CJ ... -222 psi r
e. NEBGoa07A (6/80)
8-49 22A7454 GENERAL. ELECTRIC 48 NUCLEAR ENERGY IH NO. BUSINESS GROUl' JIIEV 1 Section] Service Level 'e' Pressure Stress: P-D
= __ i = 333- (8.386)
= 2,826 psi as 2t 2 (0.494) a r
= -333 psi Stress Due to Thermal Sleeve Loads:
P = 5.08 kip M = 5.39 in-kip F z
= 6.0 kip
)l = 5.39 + 5.08 (19.33) - 103.59 in-kip a BEND * = -IIZ = 103,59 = 3.564 kai 29.07 F
6,0
= ..A A
= 13,78
= 0.436 ksi
- 333 ps i pres sure as sumed (conservative) *
- NEBGoa07A (6,aO)
8-50 NUCLEAR ENERGY GENERALe ELECTRIC ZZA74S4 SH NO. 49 BUSINESS GROUP REV 1 Stress Due To No%%le Loads: p = 6.44 kip H = 789.3 in-kip F
= 4.61 kip
. The exact &mount the safe end loads influence the thermal sleeve is unknoyu.
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 I = 362.33
= 0.381 No%zle NOTE: Corrosion not included in calculation (more conservative)
M = 789.3 + 6.44 (9.26) + 4.61 (1.8) in-kip
= 8S7.~4
.~
857.Z4 CSBFND * = 0.381 11z = (0.381) 29.07
= 11.24 lsi F
= -L§.1 CSAX * = 0.381 ...!
A (0.381) 13.78 = 0.128 lsi Total Stress
= 1,413 + 3,564 + 436 + 11,240 + 128 = 16,781 psi
= Z,826 psi CS = -333 psi r
NEBG-a07A (6/80)
8-51 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 50 BUSINESS GROUP "EV 1 TABLE 3-2 XAXIHUM PRI!lAR.Y SlRESS INTENSITY Pm - Primary Membrane All Stresses in ksi p - Primary Bending B Pm + Pb ** P P m
- Pm :t Pb Condition Section .J! Alloy , Allo ... Xaterial A 14,05 26.7 20.4S 40,05 SA-50S (CL.2)
B 14.05 lS.l 20.41 27,15 SA-SO 8 (CL.1) C 14.05 lS.1 20.33 27.15 SA-S08 (CL.l) D 14.05 1S.6 20.33 27.90 SA-350 (U2) Design E 14,05 lS.6 20.25 27.90 SA-350 (LF2) Event F 13.70 lS.6 23.94 27.90 SA-350 (U2) G 2.74 18.6 14.52 27.90 SA-350'-rLF2) -- R 2.74 18.6 14.35 27.90 SA-350 (U2) 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-50S (CL.2) B 15.46 27.10 27.65 40.65 SA-50S (CL.l) c- 15.46 27.10 27.5S 40.65 SA-50S (CL.1) Service D 15.46 27.S5 27.5S 41.77 SA-3S0 (U'2) Level E 15.46 27.85 27.51 41.77 SA-3S0 (LF2)
- C* F IS .07 27.SS 33.2S 41.77 SA-3S0 (U'2)
Event G 4.11 27. S5 22.77 41.77 SA-3S0 (U2) R 4.11 27.S5 22.33 41.77 SA-3S0 (U'2) I 3.16 19.2 17.1S 28. SO SA-3S1 (CF3) J 3.16 19.2 17.11 2S.S0 SA-3S1 (CF3)
- P is Sm for Design and the larger of 1.2 Sm or Sy for Service Level C.
mAlloYable
** P + P is 1.5 S for Design and the larger of 1.S S or 1.5 S for m BAlloyable m m y Service Level C.
- 4. ANALYSIS This section provides all the detailed thermal and stress analysis required to shoy an acceptable design for the operating trLnsients imposed on the nozzle and safe end assembly.
NEBG-a07 A (6/10)
8-52 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 51 BUSINESS GROUP REV 1 4.1 Thermal Transient Analysis. The only feedwater nozzle thermal transient for the vossel operating conditions (Sorvice Levels A and B) is defined in the dosign specification (Reference 6.1). This transient is also illustrated in Figure 4.1-1 for convenience. In order to simplify the thermal analysis, the feodwater transient was idealized as two separate transients (a heatup and cooldowu). 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 rosults in slightly nonconservative strosses. For moro detailod information on the two idealized tranSients, aoe Reference 6.1. 4.i.1 Thermal Model. Th.o uisymmetric finite element model of the feedwater nozzle is shown in Figure 4~1.1-1. The model is made up of 2-D axisymmetric isoparametric tomperature elements (STIF 55, Reference 6.3). A portion of the vessol 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 thermal properties used are as follows. Thermal properties are those of approximately 360°F.
=
Carbon Shel K 0.03972 BTU/min inoF p .. 0.283 lb/ inS 0.1226 BTU/lbmoF Stainless Steol K = 0.01327 BTU/min !n°F p = 0.290 lb/ inS 0.1191 BTU/ lbmoF NEBGoa07A (6/80)
8-'i1 NUCLEAR ENERGY GENERAL., ELECTRIC 22A74S4 SH NO. 52 BUSINESS GROUP REV 1
~\~~~______J_D_O_~_o__ __
F_~_~
,.,f FIGURE 4.1-1 FEEDWATER 'l'1I:ERMAL TRANSIEN:l (FEEDWA'IER TE.ltIPERATt1R.E VS. TIME) * * - LOAO srl!P.$
HEAT- ~p FIGURE 4.1-2 IDEALIZED THERMAL TRANSIENTS (FFEDWATER TEMPERATURE VS. TIME)
8-54 llA74S4 NUCLEAR ENERGY GENERAL., ELECTRIC SH NO. 53 BUSINESS GROUP REV 1.
~
~~:~
C ,~ r.:-
~ ~
,~
=.
-: ~
= -
~
_e
~
~ i:
JJ
'":t"' uw.
)Hi
- A
':2 ~
0
~ 1/ \ .~
W
<t /LIlli 0 x
///7 I w
//7/7 I N
N 0
/ / I / 1\11 Z
CI:: I I I I! "2 W I i tt') I f I I 1 1 I' 0 W, 'C. n 1 I ~ w I w
\ i I I 1;1 UJ\!l :illii IJ)0
-.1 1 I ) I II tt. JJ U!
0
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i I I I I Iii ex:
~
I I 1/// u 1-1 e-I I II II I ! In m z 0 I ' x
. I I I~
T I
~
1//7/1/ ry I I I I I I I I I I -. I
\
-:r e~....
8-55 NUCLEAR ENERGY, GENERALe ELECTRIC 22A7454 SM NO. 54 BUSINESS GROUP fIIEV 1-4.1.2 Feedyater Nozzle Beat Transfer Coefficients. The heat transfer co~ficients yere 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~._ltified in Figure 4.1.1-1. The calculated values for each of these surf.ces follow. Table 4.1.2-1 contains the Yater properties used at the various ttmperatures analysed. 4.1.2.1 Cool-Down Transient Region 1 ID a 9.67 in. D 0.80575 ft. No F10y Condition (Natural Convection) For natural convection the film heat transfer equation is as fo110ys (Reference 6.4): h = 0.14 I (GR Pr)1/3 f L Ar (~(pr~ 1/3
/3 a 0.14 K using 'Water properties at 350 oF. and assuming a film temperature diff"erentia1 (AT) of 10 oF. obtain
... 218.44 BTU/Br Ft 2 OF
'. NEBGoI07A (6/10)
8-56 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 55 BUSINESS GROUP REV 1 TABLE 4.1.2-1 WATER PROPERTIES T (OF) Yater Prooertv ~ W. 250 350 SOO SSO Ibm 55.6 49.0 45.9 p 62.0 60.1 58.8 3 Ft mu 1.01 1.OS 1.19 1.31 Cp lbm OF 0.998 1.00 11 Ibm 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 BID 0.364 0.394 0.396 0.391 0.349 0.325 K
Hr Ft OF P 4.52 1.88 1.45 1.02 0.87 0.93 r L 2 x 10-4 4 x 10-4 4.8 x 10-4 -4 1 x 10-3 1.1 % 10-3 p 6.9 % 10 IF 11
....1l!!.. 1.649 0.738 0.569 0.378 0.256 0.230 Ft Hr (Values tak= f%tlll Befermce 6.4) nere:
p = Density Cp = Specific Beat 11 = Viscosity K = Conductivity R e = Reynolds No. = (DVp/l1) V = Fluid Velocity Pr = Prandtl No. Mm NEBGoa07A (6/80) 1 gal = 0.1337 Ft 3
.0
8-57 NUCLEAR ENERGY GENE RAL* ELE CT RIC 22A7454 .SH NO. S6 BUSINESS GROUP "EV 1 .----------------------~~------~ Forced Convection For turbulent floy. the film heat transfer coefficient equation is as folloys: (Reference 6.4): 0 "3 x: DO. 8 pO. 4 hf = 0 .* D Ae r And for 2" floy (lO~ f10y = 3,720 Gal/min) (Reference 6.1) v = .Q = 0.25(3,720)(0.1337)(4)(60) = 14,631 Ft/Hr A n (0.80575)2 At T = 550°F R e
= DVp J1
= 2.35 x 10 6 R 0.8 e
= 1.25 x 10 5 P = 0.93 PrO. 4 .= 0.9708
'r
... h f
= 1125.13 BTU/Dr Ft 2
of At T = 500°F R = DVp
= 2.26 x 10 6 R 0.8 -= 1.211 x 10 5 e p. e PrO. 4 =
P r = 0.87 0.9465
... h f =
2 1142.23 BTU/Hr Ft of At T = 350°F 6 It 0.8 4 R e = DVp Jl
= 1.734 x 10 e = 9.80 x 10 P = 1.015 PrO. 4 = 1.006 r
2 h f = 1100.4 BTU/Hr Ft of NEBGoa07A (6/10)
8-58 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 57 BUSINESS GROUP REV 1 At T ... 200°F R = DVp
= 9.6 x 10 5 R. 0.8
= 6.107 x 10 4 e ~ 0 p = 1.88 ; PrO. 4 = 1..2853 r
2 h f = 882.84 BTU/Hr Ft of At T = 100°F DVp 5 R. 0.8 4 R. e = .~
= 4.433 x 10 0
= 3.291 x 10 p = 4.52 PrO. 4 = 1.828 r
2 h = 625.27 BTU/Hr Ft OF f N£BGoa07A (6/80)
8-59 22A7454 NUCLEAR ENERGY GEN ERAL. ELECTRIC SH NO. 58 BUSINESS GROUP Plev 1 Region 2 ID = 8.38 in. a 0.69833 ft. No Floy Condition (Natural Convection) For natural convection, the film heat transfer equation is identical to that of Rogion 1. Again using water properties at 350°F and a AT of 10°F, obtain h = 218.44 BTU/Dr Ft 2 OF f Forced Convection For turbulent f10y, the film heat transfer equation is as fol10ys: (Reference 6.4): 0.023 ID 0.8 p 0.4
= R.
e r And for 2~ flow (10~ flow = 3,720 Gal/min) (Reference 6 ~1) v = Q = 0.25(3,720)(0.1337)(4)(60) = 19,478.3 Ft/Er A 7t (0.69833)2 At T = 550°F 6 It e
= 2.7098 x 10 It 0.8 e
a 1.4007 X lOS P 0.4 r
= 0.9708 2
h f
= 1455.6 BTU/Dr Ft OF At T = 500°F It - 2.6076 X 10 6 It 0.8 = 1.358 X 10 5 e e p 0.4 r
= 0.94648 2
h f
= 1477.75 BTU/Hr Ft OF N£BGoa07A 16/80)
8-60 NUCLEAR ENERGY - GENERAL. ELECTRIC 22A74S4 SH NO. S9 BUSINESS GROUP REV 1 At T - 350°F R 0.8 e
= 1.0989 x 105 p 0.4 "" 1.006 r
2 h f
= 1423.6 BTU/Hr Ft OF At T = 200°F 6 = 6.8477 x 104 Ro = 1.1077 x 10 R 0.8 o
p 0.4 = 1.2853 r 2 h = 1142.12 BTU/Hr Ft of f At T = 100°F R 0.8 = 3.6903 x 10 4 - e p 0.4 = 1.828 r 2 h f
= 808.92 BTU/Hr Ft of NEBGoa07A (6/801
8-61 22.A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 60 BUSINESS GROUP "'EV 1 Region 3 ID ... 6 .16 in. ... 0 .51333 ft. No Floy Condition (Natural Convection) For natural convec"tion, the film heat transfer equation is identical to that of Region"l. Again using Yater properties at 350°F and a ~T of 10°F, obtain 2 h ... 2.18.44 BTU/Hr Ft of f Forced Convection For turbulent floy, the film heat transfer equation is as folloys: (Reference 6.4): And for 2~ floy (10~ floy = 3.720 Gal/min) (Reference 6.1) 0.2.5(3,720)(0.1337)(4)(60>
=
~
v ... ... 36,047.6 Ft/Hr A n (0.51333)2 At T = 550°F R = 3.686 % 10 6 R 0,8 ... 1.792 % 10 5 e o p 0.4 = 0.9708 r
... h f
... 2.533.05 BTU/Hr Ft2. OF At T = 500°F 6 R 0.8 ... 1.737S 105 R ... 3.5474 % 10 e e
p 0.4 ... 0.9465 r 2 h ... 2571.5 BTU/Hr Ft OF f NEBGoa07A (6/80)
8-62 22A7454 NUCLEAR ENERGY GENERAL" ELECTRIC SH NO. 61 BUSINESS GROUP REV 1 At T .. 350 0 F 6 R 0.8 = 1.4057 x 10 5 R c 2.7218 x 10 e 8 P 0.4 = 1.006 r At T = 200 0 F R = 1.5069 x 10 6 R 0.8 = 8.7594 x 10 4 e e p 0.4 = 1.285 r At T = 100 0 F p 0.4 r
= 1.828 R 0.8 e
= 4.7206 x 104 NEBCPa07 A (6/801
8-63 22A74S4 NUCLEAR ENERGY GENERALe ELECTRIC aH NO. 62 BUSINESS GROUP "EV 1 Region 4 As given in the design specification (Reference 6.1). the exterior surfaces are to be insulated yith material having a conduction rar cf 0.2 2 BTU/Dr Ft OF. The ambient air outside the insulation is to be at least 1000F during normal operation. Therefore. for all feedyater floy and no floy conditions. use BTU Region 5
.As given in Appendix 20 of-the design specification (Reference 6.1). the heat transfer coefficient against the vessel yall is constant for all feedyater floy conditions.
BTU
= 500 For no feedyater flow, natural convection is the medium of heat transfer.
The film heat transfer equation for natural convection is identical to that of Region 1. Using water properties of 3S0 oF and a AT of 10 o F, obtain BTU
= 218.44 Region 6 Region 6 is broken up into five separate sections. This is illustrated in Figure 4.1.2-1. In addition to this, the film heat transfer coefficient on the lower surface must be adjusted to compensate for the secondary thermal sleeve which is present there. Since the temperature of the water in the annulus is given in Appendix 20 of .the design specification (Reference 6.1),
an equivalent heat transfer coefficient for the lower surface yill be used which includes the secondary sleeve. This is illustrated in Figure 4.1.2-2 *
- NEBG-807A (6/80)
8-64 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 63 BUSINESS GROUP REV 1 d
~
'2
-uJ
()
~
\Q S
1-1 t.:l
=:
~
~
.I 0 til S1-1 t
~
til
~
1-1
~
~
- c E-<
I N
.;:r
-e:
~
t.:l 1-1
~
"E.G a07A
8-65 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 64 BUSINESS GROUP REV ]. I _
-1\&$
......-----/
~-----------------
/ ---/
'~,----~--_-~____ , -~e.~~-_-
S£c.o,,) DR A.Y
-r1lE.R.m II L. S I..E.£VE FIGURE 4.1.2-2 EQUIVALENT BEAT TRANSFER COEFFICIENT
. ~
8-66 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 65 BUSINESS GROUP REV 1 Upper Surfac e For all feedwater flows, the film heat transfer coefficient for the upper surface is given in Appendix 20 of the design specification (Reference 6.1). The coefficients given are as follows (h in BTU/Ft 5%2 OF); f Section h at Left End h at Right End f f 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 found. The equivalent heat transfer analysis for the primary sleeve is made up of the following :
- a. h of outer surface secondary sleeve (hI) f
- b. h of inner surface secondary sleeve (h2 )
f
- c. h of outer surface primary sleeve (h )
f 3 The conduction through the secondary sleeve will be neglected along with any conduction through the water. 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.
NEBc;e07 A (6/80)
- 8-67 llA74S4 NUCLEAR ENERGY GENERAL. ELECTRIC IH NO. 66
~USINESS GROUP "EV 1 Natural convection will be assumed for the annulus between the two sleeves. Therefore, from Reference 6.4. for natural convection, h co 0.14 (~) (Gr Pr)1/3 f
(0 ~ 1/3 A~/3 co 0.14 X \ ~2 (pr) a~ using properties at T m 3S0 oF. obtain BTU
- - - -..fSsu~a.ng the AT from the water to the thermal sleeve surface is the SUle
-i~r both primary and secondary sleeves, obtain q -= h eqT A (AT )
T where: A = surface area _1_ = h eqT ATT = TAnnulus - TFeedwater And q = h eql
- A (AT )
1 where: A = surface area _1_ = h e~ ATI = lAT And where: A = surface area NEBG0807 A (6/110)
8-68 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 67 BUSINESS GROUP REV 1 Using the fact that the heat flow is constant through the heat transfer patp, obtain q = constant Therefore,
=
Recalling that for natural convection,
/3
= = 101.4 !J.r obtain
(;3 101.4 Arl /3 (AT) 1
=
+ _ _-=2=-__
101.4 !J...;/3 This reduces to yield the following, 4/3
=::...a..;;:;...=.:=--_
101.4 !J.T h3
+ 2!J.T = (T Annulus
- T .)
Feedwater 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 OF Hr Ft*2 of 500 0 F !J.T = 18.74°F !J.T = 19.2s o F 3S0 o F !J.T = 73.35°F AT = 82.PF 200°F !J.T = 124.2°F !J.T = 141.25°F 100°F AT = lS6.9OF AT = 179.8°F Recalling the equation for the equivalent heat transfer coefficient, obtain
...L h
eq NEBGoa07 A (6/80)
8-69 NUCLEAR ENERGY GENE RA L fiE LEe TRIC 22Ai454 SH NO. 68
- BUSINESS GROUP fII!V 1 The following is a summary of these equivalent heat transfer coefficient calJ:ula tions.
All h's in BTU Rr Ft 2 of h eg Feedwater Temperature h3
- 750 = 1500 500°F 114.6 124.58 350 0 F 165.41 192.12 200°F 189.15 2.24.5 100 0 F 2.00.37 240.3 Due to a calculation error in the trial and error process explained earlier, the foliowing equivalent heat transfer coefficients were used instead of the correctly calculated values shown above.
h eg Feedwater h3 = 750 h3 = 1500 Temperature 500°F 111.57 124.58 350°F 158.03 185.65 200°F 178.84 215.03 100°F 188.5 228.95 To assess this five percent difference in the coefficients, the Biot number for each pair of coefficients was compared. The comparison showed that the change in the Biot number was very small. This coupled with the fact that the portion of the thermal sleeve affected by this five percent difference is relatively far away from any highly stressed regions. Thus yielding the conclusion that the effect of this five percent difference in coefficients on the highly stressed areas, is negligible. The Blot numbers are contained in DRF (Reference 6.5) *
- NEBG.a07A (6,80)
8-70 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 69 BUSINESS GROUP REV 1 4.1.2.2 Heat-Up Transient Region 1 Forced Convection For turbulent floY, the film heat transfer equation is as folloys: (Reference 6.4): For 2S percent floy, from previous section (Paragraph 4.1.2.1), v = 14,631 Ft/Hr and for 100 percent floy v = 58,524.1 Ft/Hr At T = 100°F (FloY = 25 percent) The heat transfer coefficient is identical to that calculated for the Cool-DoYn transient (Paragraph 4.1.2.1) BTU
= 625.27 At T = 180°F (FloY = 25 percent)
Will assume the heat transfer coefficient to be identical to that calculated at 200°F in the Cool-doYD section. BTU NESG-e07 Po (6/10)
8-71 22A7454 70 GENERAL. ELECTRIC SH NO. NUCLEAR ENERGY "EV 1 BUSINESS GROUP At T "" 260 0 F (Flow a 25 percent) 6 4 Re - 1.2187 x 10 R 0.8 "" 7.3912 x 10 e p 0.4 "" 1.1605 r
... h f
= 969.55 BTU Rr Ft 2 of At T = 376 0 F (Flow = 100 percent)
R .= 6.936 x 10 6 R O* S = 2.9709 x 10 5 . e e P 0.4 = 1.006 r h = 3335.8 BTU. f Rr Ft 2 of
- ~.EBG0807A (6/80)
8-72 NUCLEAR ENERGY GENERAL. ELECTRIC llA7454 SH NO. 71 BUSINESS GROUP REV 1 Region 2 ID = 8.38 in. = 0.69833 Ft. Forced Convection
. For turbulent flow. the film heat transfer equation is as fo11oys:
(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 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 h f
= 1142.12 NEBGoa07A (6/10)
8-73 22A7454 GENERAL. ELECTRIC 72 SH NO. NUCLEAR ENERGY BUSINESS GROUP .. EV 1 At T a 260 0 F (Floy - 25 percent) Re s 1.406 % 10 6 R 0.8 = 8.2875 % 10 4 e P 0.4 .. 1.1605 r At T = 376°F (FloY ~ 100 percent) Re = 8.0 % 10 6 Re0.8 = 3.33 % 10 5 p 0.4 = 1.006 r BTU 4315.7 2 Hr Ft of
- NEBGoa07A (6/10)
8-74 22A7454 NUCLEAR ENERGY _ GENERALe ELECTRIC SH NO. 73 BUSINESS GROUP REV 1 Region 3 ID = 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 = 100 0 F (Flow = 2S percent) The heat transfer coefficient is identical to that calculated for the Cool-Down transient (Paragraph 4.1.2.1) BTU
= 1407.67 At T = 180 0 F (Flow = 2S percent)
Will assume the heat transfer coefficient to be identical to that calculated at 200 0 F in the Cool-down section (Paragraph 4.1.2.1>. BTU
= 1987.5 NEBGoa07A (6/110)
8-75 22A7454 NUCLEAR ENERGV GENERAL. ELECTRIC SH NO. 74 BUSINESS GROUP "EV 1 At_T = 260°F (Flow = 25 percent) 6 10 5 Re = 1.913 % 10 R 0.8 e K 1.06 % p 0.4 = 1.1605 r At T = 376°F (Flow = 100 perc.nt) R = 1.0887 % 10 6 ; R 0.8 = 4.261 % -10 5 e e p 0.4 = 1.006 r hf = 7509.98 BTU 2 llr Ft of
- NEBG0807A (6/80)
8-76 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 7S BUSINESS GROUP REV 1 Region 4 As given in the design specification (Reference 6.1), the exterior surfaces are to be insnlated with material having a conduction rate of BTU 0 .* 2 BTU
= 0.2 Region 5 As given in the design specification (Reference 6.1), the heat transfer coefficient against the vessel wall is constant for all feedwater flow conditions.
BTU
= 500 Region 6 The analysis of Region 6 here follows the same procedure of the analysis of Region 6 during the cool down transient (Paragraph 4.1.2.1). Refer to Paragraph 4.1.2.~ for greater details.
Upper Surface For all feedwater flows. the film heat transfer coefficient for the upper surface is given in Appendix 20 of the design specification (Reference 6.1). They are listed in Paragraph 4.1.2.1. Lower Surfaces The procedure used is identical to that of Paragraph 4.1.2.1. For more details. refer to Paragraph 4.1.2.1. The following equation must be solved by trial and error to obtain the AT across the thermal sleeve. 101.4 A~/3
=::..a..:.-=.:=-_ + 2 AT = (T - T )
h3 Annulus Feedwater . NEBGoa07A (6/10)
- 8-77 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SM NO. 76
. BUSINESS GROUP "ev 1 The following is a summary of the solutions.
BTU BTU Feedwater h3 = 7S0 h3 .. 1S00 Temperature Ft 2 Hr OF Hr Ft 2 OF 100 0 F b.T .. 1S7°F b.T = 180°F 180°F b.T .. 131°F b.T . 149°F 260°F b.T . 104°F IlT 117.7°F 376°F b.T -= S8°F b.T 64.soF Recalling the equation for tho equivalent heat transfer coefficient, obtain The following is a summary of these equivalent heat transfer coefficient calculations. All h's in Rr Ft 2 of h Peedwater e9 Temperature ~ = 1500 1000 P 200.4 240.4 191.7 228.0 180.9 213.14 ISS.S 179.04 4.1.2.3 Normal Operation. The heat transfer coefficients used for the normal operation run are the s&me as those given during the heat up transient (Paragraph 4.1.2.2) when T .. 376°F *
- NEBG0807 A (6,80)
8-78 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 77 BUSINESS GROUP REV 1 4.1.3 Feedyater Nozzle Annulus Fluid Temperature, Appendix 20 of the design
.pe~ification (Reference 6.1). defines the annulus fluid temperatures for all conditions. For completeness it is repeated be1oy.
lfhere: T - Annulus fluid temperature TFW = Feedyater fluid temperature TA = Region A fluid temperature = S46°F C = Coefficient defined by table be10y Points Feedyater Floy A ~ £. d + e i. 2~ 0.70 *0.92 0.96 . 0.96 0.96 10~ 0.68 0.83 0.86 0.86 0.90 Use the above values of C for all cases yith feedyater floy. Use C = 1.0 yhen no feedyater floy. Use C = 1.0 doynstream of f for all feedyater floys. Interpolate linearly betyeen defined points; see Figure 4.1.2-1 for point locations. NEBGoa07A (6/80)
- 8-79 NUCLEAR ENERGY GENERAL. ELECTRI-C 12.A7454 SM NO. 78 BUSINESS GROUP - .. EV 1 4.1.4 Thermal Analysis Results. By applying the boundary conditions (heat traFsfer coefficients and flow temperatures), a transient heat transfer analysis was performed using the finite element program ANSYS (Reference 6.3) for the thermal transients _~~.;~- :: 'bed earlier. During these transients, the isolated face of the disc in the finite element model was maintained at a constant vessel temperature of 546°F to simulate the vessel as a heat source.
The temperature solutions obtained for the various times of the transients yere saved on tape for later use in the thermal stress analysis. Some of the isotherm plots obtained for various times of the transients are shown in Figures 4.1.4-1 through 4.1.4-7 *
- NEBG-a07A (6/10)
8-80 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 79 BUSINESS GROUP REV 1 i r
\
I i i I J C'-. j
~
_-~M---**;* r'-&-_.-..-.-;-----_-..------------.--------~ FIGURE 4.1.4-1 COOLDOWN TRANSIENT NEa~07A (6"0) e*
8-81 GENERAL. ELECTRIC ZZA74S4 80 NUCLEAR ENERGY SH NO. BUSINESS GROUP REV 1
----\ ---, \
i I
\I
- MONTICELLO fEEDWRTER NOIZLE SiRESS ~LYSIS - lEMPERAiURE PLOT (TlME~lS.0 S( :'
J
~;,I I
I FIGURE 4.1.4-Z COOLDOWN TRANSIENT HEBG-a07A (6180)
8-82 NUCLEAR ENERGY GENERAL .'ELECTRIC 22A7<4S4 SH NO, 81 BUSINESS GROUP REV 1 I' I I i I Ii I MO~TlCELLO FEEtWlTER t-.OlZLE ST1\ESS F\~LY51S - TE~EAAIURE PLOI (iIHE:.60.0 SECI I I II J..
~ I I
FIGURE 4.1.4-3 COOLDOIN TRANSIENT NESa.a07A (6/80)
8-83 GENERAL. ELECTRIC 22A7.S4 IH NO. 82
*J NUCLEAR ENERGY BUSINESS GROUP REV ~
'. MONTICELLO FEEDWATER NOZZLE STRESS RNALYSIS HERT-UP TEMPERATURE PLOT
-t; .. :2.0 SEC..
o o '. HElSa..07A (6/10)
8-84
----- GENE RAL* ELEe TRIC 11A70454 NUCLEAR ENERGY SH NO. 83 BUSINESS GROUP REV 1 MONTICELLO FEEDriRTER NOZZLE STRESS RN *. L 1515 HERT -UP TEMPERRTURE PLOT
,e. 3.0 SEC..
N C FIGtJ'KE 4.1.4-5 NEB~07A (6/80)
8-85 NUCLEAR ENERGY GEN ERAl. ELECTRIC 22A74S4 SH NO. 84 IUSINESS GROUP REV 1
- MONTICELLO fEEDWATER NOZZLE STRESS RNAl1SlS HEA1-UP TEMPERATURE PL01 i:"" / 2.0 SEC.
/I")
If) FIGURE 4.1.4-6 NEBc..07A (6/aO)
8-86 NUCLEAR ENERGY GEN ERAL. ELECTRIC ZZAHS-4 SH NO. 8S BUSINESS GROUP REV J If"
- r.
i ..... 1 I PlGUU -4.1.-4-7 NEBGoa07 A (6"0)
8-87 22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 86 BUSINESS GROUP 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 the fatigue usage factor for the system. 4.2.1 Selected Locations for Stress Evaluation, The sections showu in Figure 4.2.1-1 are the locations selected for stress evaluation. Finite element stresses are integrated across each section to dettrmine 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 shoYn in Figure 4.2.1-2. These calculations yere performed using an engineering aid computer program 'NONO'. A listing of this program is included in Appendix 10 *
- '. j
- NEBG-a07A (6/10)
8-88 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7 .. S4 SH NO. 8*/ BUSINESS GROUP REV 1 Sect ion Element ~bers (Out/In) A 272/275 B 260/263 C 184/187 D 180/183 E 70, 71, 73, 81. 89 I II I F 29/32 G 96. 88. 80, 72 B 220/223 I 224/227 1 2 .. 0/2 .. 3
/
0 0 I I
.1 I 0@
F'IGtJU 4.2.1-1 LOCATIONS FOB. EVALUATING SUESSES NEB~07A ",.0)
8-89 NUCLEAR ENERGY GEN ERAL. ELECTRIC 22.AHS4 SH NO. 88 BUSINESS GROUP REV 1 PEAl< STRES S BENDI~G STRESS I 1 ACn:Al. STRESS
---DIS'l1UBl"TION
- - - EQUIV ALEh"T
. LINEAR STRESS SECTION - .:. - - EQrIV Al.E~"T nUCKNESS MEXBRA!\E STRESS LINEARIZATION OF STRESS DISTRIBUTION ACROSS A SECTIO~
THICKNESS FIGUe '4.2.1-2 LlNEAlIZATIOO OF SnESS DISnIBOTIC!i ACROSS A SEctION THICDmSS
- NEBGoa07 A 16,.0)
--------___ ____---J.
. 8-Qn NUCLEAR ENERGY GEN ERAL. ELECTRIC 22A74S4 SH NO. 89 BUSINESS GROUP REV 1
~--.:::-.
4.2.2 Thermal Stress Analysis, The thermal stresses were obtained using the fi~ite element computer program ANSYS. The same finite element model used for the thermal transient analysis (Figure 4.1.1-1) was also used for the thermal stress analysis. However, instead of temperature elements being used, 2-D axisymmetric isoparametric stress elements (STIF-42, Reference 6.3) were used. The modeling of the portion of the vessel wall as a disc is a conservative assumption for the thermal stress solutions since the inplane radial stiffness of a disc is higher than that of a shell. The applied boundary condition at the isolated surfaces of the vessel, safe end. and thermal sleeve is the generalized plane strain condition. The elastic properties used were considered constant for all temperatures. These properties are: 6 Carbon Steel E = 27.2 x 10 . psi 6 a = 7.33 x 10- in./in.
= 0.3 3
p = 0.283 Ibf/in. Stainless Steel E = 26.85 x 10 6 psi 6 a = 9.87 x 10- in.lin.
= 0.3 p . 0.29 Ib fl in.
3 6 psi Vessel Steel E = 28.8 x 10 6 a = 7.33 x 10- in.1 in.
= 0.3 3
p = 0.283 Ibfl in. NEBGoa07A (6/80)
8-91 22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 90 BUSINESS GROUP "EV 1 4.2.2.1 Selection of Times for Stress Evaluation, The times in the trLnsients considered important for subsequent thermal stress evaluation are determined by a review of temperature differences between selected nodes of the thermal model. This process is p'erformed for the cool doY'll t~nsient. Figure 4.2 .* 2.1-1 illustrates the selected nodal pOints and Table 4.2.2.1-1 contains the temperature information. Actual temperatures are in DRF B13-909 (Reference 6.5). 4.2.2.2 Thermal Stress Analysis Results. The thermal transient stresses were evaluated at the indicated times of the transients. The thermal membrane plus bending stresses for each of .the previously noted transient times are tabulated in Tables 4.2-1 through 4.2-9. The thermal membrane plus bending plus peak stresses for each of the previously noted transient times are tabulated in Tables 4.2-10 through 4.2-18 *
- NEBG-I07A (6/80)
8-92 NUCLEAR Ef-JERGV BUSINESS GROUP I 22A7454 REV 1 SI-I NO. 91 Location Nodj'l p::tir.t~ 1 354. 353. 352 1 324, -. 3't~
~.,.,
~-.
3 13G. i3S. 134, H3. 131
.. 131. 130. 129. 123.127, 126
... .,... 3eo. 319. 378 S ."~' ~tSl.
6 40. 39. 3a. 37. 35 7 80. 19. 18. 71. 16
- 0)
FIGUP.E 4.2.2.1-1 LOCATIONS US:!!) FOn EVALUATING TEUPITRAmn..E DIF1-r'.!?"iCE NESG-a07A (6/80) * :
8-93 22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SHNO. 92 BUSINESS GROUP "EV 1 TABLE 4.2.2.1-:1 T'EMPERATORE DIFFERENCES FOB. COOL DOWN ntANSIENT Load Time Step (Minutes) T7 .- T6 T7 - T3 AT, AT3 AT4 6 0.10 59 99 205 129 30 7 0.1167 69 117 220 128 33 8 0.1333 78 130 232 125 34 9* 0.150 84 141 250 122 36 10 0.1667 90 149 265 120 36 11 0.2083 102 163 274 115 36 12 0.250 111 169 281 111 36 13 0.333 128 172 269 36 14 0.50 147 156 235 32 15 0.750 158 122 204 24 16 1.00 156 94 17 2.00 127 18 3.00 112 (All AT's in OF) DRF B13-909 contains actual nodal temperature data (Reference 6.S) Times Selected Load Step Time (Minutes) 6 0.10 12 0.25 16 1.00 23 Steady Sta te
- NEBG407A (6/10)
8-94 22A7454 NUCLEAR ENERGY . GENERAL. ELECTRIC SH NO. 93 BUSINESS GROUP REV 1 4.2.3 Xechanical Load Stress Analysis, The mechanical load stresses were obtained by using the procedure followed in Paragraph 3 to obtain the design stresses. The only variation is in obtaining the applied loading (however, this will be investigated in the following Paragraph), For completeness, the calculations are presented in this report in Paragraph 4.2,3.2. 4.2.3.1 Applied Mechanical Loading. To obtain the largest load ranges, the loadings given in Section 3 are used. Note that in load range calculations, dead weight loads are not included. These loads simply cancel out. Nozzle Load ins P = (F 2 + F 2)1/2 x y H = (M 2 + H 2 + H 2)1/2 x y z
, A' Nozzle Loading ( Service Level 'B' )
P = (0.02 2 + 0016 2 )1/2 = 0.161 2 (12.0 2 + 12 .1 2 + 45.0 )1/2 Static II = = 48.12 F z = 0.21 P = (0.29 2 + 2.512 )1/2 = 2.53 Dynamic M = (9.3 2 + 158.9 2 + 313.4 2 )1/2 = 351.5 Fz = 2.23 NEBGoI07 A 16/10)
- 8-95 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 94 BUSINESS GROUP Plev 1
'B' No%%le Loading (Service Level 'B')
p ~ (0.82 2 + 4.34 2 )1/2 ~ 4.42 Sta tic F .. 1.37* Dynamic F z
= 0.26
- NEBGoa07A (6/80)
8-96 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 95 BUSINESS GROUP REV 1 Thermal Sleeve Loadins (Service Leve 1 ' B') P = (F 2 + F 2)1/2 x y
)l = (M 2 + H 2 + H 2)1/2 x y z P D 0 Static II = 0 F
z
= 3.7 P = (2.5 2 + 0.3 2 )1/2 = 2.52 Dynamic M = (1.2 2 + 2.0 2 )2 1/2 = 2.333 F
z
= 1.5 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 D. 4.42 kip Static M = 275.4 in kip 1.37 kip z = F P = ;t 3.136 kip Dynamic )l = +/- 390.9 in kip F z = +/- 0.26 kip NEBGoa07A (6/80)
- 8-97 l2A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 96 BUSINESS GROUP ~ev 1 Thermal Sleeve Loads (Service Level 'B')
p -= :!: 2.52 kip Dynamic H -= :!: 2.333 in kip F z
= :!: 1.5 kip 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 summary of the calculations is given in Table 4.2.3.2-1 *
- NEBGoa07A (6/10)
8-98 NUCLEAR ENERGY --- GENERAL. ELECTRIC 22A7454 SH NO. 97 BUSINESS GROUP REV 1
----------------------~--------~.
Section A No%% Ie Loads P = 4.42 kip p = +/- 3.136 kip M = 215.4 in kip )l = +/- 390.9 in kip F z
= 1.37 kip F z
= +/- 0.26 kip 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 431.3 332.9
= 6.41 ksi 51.98
= 8.3 ksi 51.98
- a. AX.
F
= -% =
A Thermal Sleeve Loads 1.37 19.03
= 0.072 ksi p = +/- 2.52 kip 0.26 19.03
=0.014 ksi If z
= 3.7 kip K = +/- 2.333 in kip F
= +/- 1.5 kip K = 3.7 (2.36) = 8.73 in kip M = 2.333 + 2.52 (15.76) + 1.5 (2.36)
= 45.59 in kip 8.73 M = 45,59 51.98 = 0.168 ksi a~rw = Z 51.98 = 0.877 ksi 3.7 19.03
= 0.195 ksi a,AI. =
F J. A
= 1.5 19.03
= 0.079 ksi Total Stress a,BEND = 6.578 psi = +/- 9,177 psi a,AX. - 267 psi .. +/- 93 psi NEBGoa07A (6/80)
8-99 22A7454 NUCLEAR ENERGY GENERAL. ELECTRI.C SH NO. 98 BUSINESS GROUP JIIEV 1 Section B Nozzle l.,,"ds > * *: - p .. 4.42 kip p ... ~ 3.136 kip M .. 275.4 in kip M .. ~390.9 in* kip F .. 1.37 kip F s ~ 0.26 kip z z M s 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 .. 423.1 in kip
= .M = 321,34 = 6.182 ks i 423,1
= 8,14 lsi a"BEND Z 51.98 51.98 F
--1 .. 1.37 .. 0.072 ksi a., ..
. AX.
F
-=
A z 0,26 19.03
.. 0.014 lsi 19.03 Thermal Sleeve Loads p .. +/- 2.52 kip F
z
.. 3.7 kip M .. +/- 2.333 in kip F .. +/- 1.5 kip z
M .. 3.7 (2.36) = 8.73 in kip M .. 2.333 + 2.52 (18.3~) + 1.5 (2.36)
.. 52.17 in kip a~END=
-MZ .. 8,73 51.98 = 0.168 ksi 52,17 51.98
.. 1.004 ksi a., .. F
-' = 3.7 .. 0.195 ksi a., .. F
-'-= 1,5 19.03
.. 0.079 ksi AX. A 19.03 AX. A Total Stress a., .. 6,350 psi .. ~ 9.144 psi BEND a.,AX.
.. 267 psi s ~ 93 psi NEBG0807A (6/aO)
8-100 22A7454 NUCLEAR ENERGY GENERAL" ELECrRIC SH NO. 99 BUSINESS GROUP REV 1 Section C/D Nozzle Loads P c 4.42 kip P c +/- 3.136 kip M = 275.4 in kip M .. +/- 390.9 in kip F = 1.37 kip F .,. :!:. 0.26 kip z z H = 275.4 + 4.42 (7.47) + 1.37 (0.56)" H .. 390.9 + 3.136 (7.47) + 0.26 (0.56)
= 309.19 in kip = 414.48 in kip 309,19 = 5.952 ksi }l = 414.48
= 7.974 ksi 51.98 Z 51.98 F F 1.37 0.26
-z= = 0.072 ks i -1.= = 0.014 ksi 19.03 19.03 Thermal Sleeve Loads p .,. :!:. 2.52 kip Fz - 3.7 kip H = :!:. 2.333 in kip F .,. :!:. 1.5 kip z
M = 3.7 (2.36) = 8.73 in kip H .,. 2.333 + 2.52 (21.12) + 1.5 (2.36)
= 59.10 in kip 8.73 ! 59.10 51.98
= 0.168 ksi a~mD
=
Z = 51.98
= 1.137 ks i
- a. ., -
AX. Fz A
=
3,7 19.03
= 0.195 ksi at AI.
F
-1.=
A
-L.L 19.03
= 0.079 ksi a.
Total Stress 6,118 psi '"' :!:. 9.111 psi
=
a. BEND
= 267 psi = :!:. 93 psi AX.
NEB~07A (6110)
8-101 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 100 BUSINESS GROUP JIIEV 1 Section E No%%le Loads p .. 4.42 kip p a ~ 3.136 kip H = 275.4 in kip )( a ~ 390.9 in kip F = 1.37 kip F z
= ~ 0.26 kip z
)( .. 275.4 + 4.42 (4.72) + 1.37 (0.56) )( = 390.9 + 3.136 (4.72) + 0.26 (0.56)
= 297.03 in kip = 405
- 85 in kip 405.85 297.03 51.98 = 5.715 ks i a~ = !Z = 51.98
= 7.808 ksi BEND F
F
-=
% 1.37
= 0.072 ksi a~ =
...!. = 0.26 = 0.014 ksi 19.03 AX. A 19.03 Thermal Sleeve Loads
~ p = ~ 2.52 kip F
= 3.7 kip )( - ~ 2.333 in kip F = :t 1.5 kip H = 3.7 (2.36) = 8.73 in kip )( = 2.333 + 2.52 (23.87) + 1.S (2.36)
= 66.03 in kip 8.73 !l 66.03 a = !Z = 51.98 = 0.168 ksi a~END = z = 51.98 = 1.27 ksi tBEND F F a.AX. = -%=
A
.2.:.L 19.03
= 0.195 ksi a~
AX.
= -
A
= 1.5 19.03
= 0.079 ksi Total Stress a
tBEND
. 5.883 psi ... :t9.078 psi at
... 267 psi = :t 93 psi AX.
NEBG.807A (6/80)
8-102 22A7454 NUCLEAR ENERGY GENERAL., ELECTRIC SH NO. 101 BUSINESS GROUP REV 1
----------------------~~------~tt
'Section F Nozzle Loads P = 4.42 kip p = +/-. 3.136 kip x = 275.4 in kip H = +/-. 390.9 in kip F = 1.37 kip F z
= +/- 0.26 kip z
x = 275.4 + 4.42 (1.62) H = 390.9 +'3.136 (1.62)
= 282.56 in kip = 395.98 in kip 282.56 395.98 10.13 k5 i
= 7.23 ks i 39.09
=
39.09 F F 1.37 z 0.26
-=
z
= 0.087 ksi = - = = 0.017 ksi tt 15.89 A 15.89 Thermal Sleeve Loads p = +/-. 2.52 kip F
z
= 3.7 kip H = +/- 2.333 in kip F
z
= +/- 1.5 kip x = 3.7 (1.8) = 6.66 in kip H = 2.333 + 2.52 (26.97) + 1.5 (1.8)
= 73.0 in kip 6,66 = 0.171 ksi }! 73 1.868 ksi a = = =
39,09 eBEND Z 39.09 F F z 3,7
= = -z= --L.L 0.095 ksi
-=
15.89 0,233 ksi a eAX. A 15.89
=
Total Stress at - 7,401 psi = +/-. 11,998 psi BEND at = 320 psi = +/-. 112 psi tt AX. NEBGoa07A (6/aO)
8-103 22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 102 BUSINESS GROUP AEV 1 Section G Nozzle Loads p = 4.42 ~ip p = +/- 3.136 kip
)[ = 275.4 in kip )[ = +/- 390.9 in kip F = 1.37 kip F z
.. +/- 0.26 kip z
)[ = 275.4 + 4.42 (4.72) + 1.37 (1.8) )[ .. 390.9 + 3.136 (4.72) + 0.26 (1.8)
= 298.73 in kip = 406.17 in kip M
a~ = (0.381) = ~ = (0.381) 2~~:~; = 5.1 ksi ~~ = 0.381 Z = (0.381) 406.17 22.32 = 6.934 ksi BEND a, F
= 0.381 ~ = (0.381) ...L.ll = 0.05 ksi a, F
= 0.381....!. = (0.381) 0.26 = 0.01 ksi AX. A 10.46 AX. A 10.46 Thermal Sleeve Loads p = +/- 2.52 kip F = 3.7 kip H = +/- 2.333 in kip z
F z
= +/- 1.5 kip
)[ - 2.333 + 2.52 (23.87)
= 62 .49 in kip
= 11Z =
62.49 a~ 22.32 = 2.8 ksi BEND F
-=
z 3.7 10.46
... 0.354 ksi a,AX.
=
F
-z=
A 1.5 10.46
= 0.144 ksi Total Stress a,BEND - 5.100 psi = +/- 9.734 psi a, = 404 psi = +/- 154 psi AX
- NEBG-807A 16/80)
8-104 22A7454 NUCLEAR ENERGY GEN ERAL. ELECTRIC SH NOo 103 BUSINESS GROUP REV 1 Section H Nozzle Loads P = 4.42 kip P = ~ 3.136 kip M = 275.4 in kip M = ~ 390.9 in kip F z
= 1.37 kip F z
= ~ 0.26 kip X = 275.4 + 4.42 (8.47) + 1.37 (1.8) H = 390.9 + 3.136 (8.47) + 0.26 (1.8)
= 315.3 in kip = 417.93 in kip a~BEND '. 1 = Z
= (038) M
= (03 . 81 )315.3_
22.32 - 538
- 3.LS1
",0 a()BEND = 03. 81 ZM
= ( 0.381 )417.93 22.32 = 7.134 i:.si F
....Lll = 0.05 = 0.381 ~ = (0.381) ~ = 0.01 i:.si (0.381) ksi 10.46 A 10.46 Thermal Sleeve Loads P = ~ 2.52 kip F = 3.7 kip M = ~ 2.333 in kip z
F = ~ 1.5 kip z M = 2.333 + 2.52 (20.12)
= 53.04 in kip 53.04 a()BEND = H Z
= 22.32 = 2.377 ksi
- a. AX.
=
F
-z=
A 3.7 10.46
= 0.354 ksi a OAX. =
F
-z=
A 1.5 10.46
= 1.44 ksi Total Stress .~ ..
a~ co 5.383 psi = ~ 9.511 psi BFND
- a. AX.
.. 404 psi = ~ 154 psi NEBGoa07 A (6/80)
8-105 22A7454 104 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. BUSINESS GROUP PlEV 1 Section I Nozzle Loads P '" 4.42 kip p ... ~ 3.136 kip M = 275.4 in kip M "" ~390.9 in kip F '" 1.37 kip p .. ~ 0.26 kip z z M = 275.4 + 4.42 (S.47) + 1.37 (1.8) M ... 390.9 + 3.136 (8.47) + 0.26 (l.S)
= 315.3 in kip = 417.93 in kip a,. = (0.381) ~ ~= (0.381) ;~~o; = 4.133 hi all "" 0.381 ~ = (~.381) .4;~:~; = 5.478 hi WID ~
a, = 0.381 ...! F .
= (0.381) ...L.ll = 0.038 hi a, F
= 0.381...! = (0.381) 0.26 = 0.008 ksi AX. A 13.78 AX. A 13.78 Thermal Sleeve Loads P = ~ 2.52 kip F
z
= 3.7 kip M "" ~ 2.333 in kip
?
F z
= ~ 1.5 kip M ... 2.333 + 2.52 (20.12)
"" 53.04 in kip K 53.04 aOBF1.Tt>
=
Z
=
29.07 = 1.83 ksi a,AX. F F
-z= -LL ... 0.269 ks i "" -!= -L.L ... 0.109 ks i 13.78 A 13.78 Total Stress at ... 4.133 psi = .+/-7,308psi BFND at = 307 psi = ~ 117 psi AX.
NEBGoa07A (6/80)
8-106 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 105 BUSINESS GROUP REV 1 Section r Nozzle Loads P = .4.42 kip P = ~ 3.136 kip H = 275.4 in kip M = ~390.9 in kip F z
= 1.37 kip F z = ~ 0.26 kip M = 275.4 + 4.42 (9.26) + 1.37 (1.8) M = 390.9 + 3.136 (9.26) + 0.26 (l.S)
= 318.8 in kip = 420.4 in kip (l't mID z
=(0. 381) = M = ('03 . 81 ) 318.8 - 4 1 7 .....
29 .07 -
- 9 ....$1 (I' b BEND
=. 1 Z =.
0 38 M (0 381) 420.4 = 5 511..-' 29 .07 . ~1 F F (1'. .. 0.381 .J. = (0.381) ...Lll = 0.038 = O. 3 81 -Z = (O.3 81) 0.26 = 0.008 ksi ksi IJ.. A 13.78 A 13.78 Thermal Sleeve Loads p .. ~ 2.52 kip F z
.. 3.7 kip M = +/- 2.333 in kip F
z = +/- 1.5 kip M = 2.333 + 2.52 (19.33)
= 51.05 in kip H 51.05 (l'b = Z
=
29.07 = 1.756 ksi BEND F (1', F (1'. = -z = 3.7
= 0.269 ksi = .J.= ~ = 0.109 ksi AX. A 13.78 AX. A 13.78 Total Stress (1'. - 4.179 psi = ~ 7,266 psi BF1ID (1'. = 307 psi = ~ 117 psi AX.
NEBG0807 A (6/80)
8-107 22A7454 106 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. INESS GROUP REV 1 TABLE 4.2.3.2-1 lLUIlfIDf MECHANICAL LOAD STRESS INTENSIlY (Service Leve 1 ' B')
..I)"L __ '.c Stress*
Section Static 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 5.504 :!:. 9,888 B S.787 :!:. 9.665 I 4,440 :!:. ~ .425 1 4,486 +/- 7383 (All stress in psi)
- Dynamic stresses are due to seismic only *
... EBCPa07A (6/80)
8-108 NUCLEAR ENERGY GENERAL. ELECTRIC ZZA74S4 SH NO, 107 BUSINESS GROUP REV 1 4.2.4 Pressure Stress Analysis. The pressure stress'es were obtained using the finite element computer program ANSYS. The I.ame finite element model used for the thermal stress analysis (described in Paragraph 4.2.2 and Figure 4.1.1-1) was also used for the pressure stress analysis. The modeling of the portion. of the ves'sel "a11 as a disc is a nonconservative assumption. It is recognized that the pressure .tresses in the vessel "all regions obtained from using this model are not .trictly valid. Ho"ever, stresses in the safe end-thermal sleeve regions are valid since the effect of this modeling in these regions is insignificant. The applied boundary condition at the isolated surfaces of the safe end and thermal .leeve "as the equivalent meridional stress caused by the pressure. The vessel boundary condition was the average of the hoop and meridional stresses. 4.2.4.1 Pressure Stress Analysis Results. The pressure stresses were evaluated for the specified pressures of 1,111 psig nozzle pressure and 1,000 psig vessel pressure. The pressure membrane plus bending stresses are tabulated in Table 4.2-9. The pressure membrane'plus bending plus peak stresses are tabulated in Table 4.2-18. Figures 4.2.4-1 and 4.2.4-2 sho.." an isostress plot and a deflection plot of the pressure case. NEB~07 A (6/80)
8-109 NUCLEAR ENERGY GENERALe ELECTRIC 22A7454 SH NO. 108
..,SINESS GROUP REV i I I "ll.--+--+--\+----'I U---+---+--+-~
I I I I I I+---I---+--+--I-Ij
, tt--+--+--4-4-I:
I tt---t--+--l-...u+
~\ .
I n----t--+----+-i..AJ
- / f'-~ =f'-.
I~'\.:TH-t-+-I--LL"~ I~ '\ITi-t-+--J-L-.L'~~ ,
,~~ r--.~ I'-.~t':~~
' ...... -~'-
- -~ _ _ /'t--!'---"...
' r--~ ~~~~
__ _ _ _ _ ~~_~~_~,=E~~~~~~
~
~~~~_!:16'~~_-
EEE3~~===z~~3E3:~~~~~ FIGURE 4.2.4-1 MONTICELLO FEEDWATER NOZZLE STRESS ANALYSIS - DISPLACEMENT PLOT (PRESSURE RDN) NEB~07A (6180)
8-11 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 109 BUSINESS GROUP FIGURE <4.2.4-2 MONTICELLO FEEDWATER NOZZLE SnESS ANALYSIS - SnESS PLOT * (PRESSURE RUN)* N£BG-a07A (6/10)
8-111 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 110
- BUSINESS GROUP ~EV 1---
4.2.5 Total Primarv Plus Secondary Stress Ranges, This .ectio~of the report calculates the P+Q stress intensity ranges at the previously selected locations in order to validate the subsequent fatigue iL'::"_:. sis. These stress ranges are calculated as the sum of three independent ranges; one for thermal stresses, a second for the maximum safe end and thermal sleeve load stresses, and a third for the pressure stresses. 4.2.5.1 Thermal Stress Ranges, The maximum thermal membrane plus bending stress ranges can be found by inspection using Tables 4.2-1 through 4.2-8. Table 4.2.5-1 yill be used to identify the transient times used in the thermal aJ;la lys is. TABLE 4.2.5-1 TRANSIENTS USED IN THERMAL STRESS EVALUATION
- 1. Cool Do'W11 (t = 6.0 sec)
- 2. Cool DOWll (t = 15.0 sec)
- 3. Cool Do'W11 (t = 60.0 sec)
- 4. Cool DOWll (t = steady state)
- 5. Heat Up (t = 2.0 sec)
- 6. Heat Up (t = 3.0 sec)
- 7. Heat Up (t = 12.0 sec)
- 8. Normal Operation
- 9. Stress Free The maximum thermal membrane plus bending stress ranges are found to be as folloys:
Maximum Stress Location Range (ps 1) Cases A 4,709 1-7 B 4,907 1-7 C 11.917 1-6 D 12.188 1-6 E 18,275 2-9 F 28,777 1-7 G 34,852 2-9 H 38,999 1-9 I 43,017 1-9
] 38,310 2-9 NEBG-807A (6/10)
8-112 22A7454 NUCLEAR ENERGY GEN ERAL. ELECTRIC SH NO. 111 BUSINESS GROUP REV 1 4.2.5.2 Nozzle End and Thermal Sleeve Load Stress Ranges. The maximum nozzle end and thermal sleeve load stress ranges are found using Table 4.2.3.2-1. The maximum stress ranges are as follows: Maximum P+Q Location (hi) A 18.54 B 18.48 C 18.41 D 18.41 E 18.35 F 24.22 G 19.78 H 19.33 I 14.85 J 14.77 4.2.5.3 Pressure Stress Ranges. The pressure membrane plus bending stress ranges are found using Table 4.2-9. The P+Q pressure stress intensities are found to be as' follows: sn Location (psi) A 5,963 B 9,437 C 8,941 D 8,873 E 6,524 F 8,678 G 7,654 H 4,567 I 4,670 J 6,229 NEBG-a07 A (6/80)
8-113 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 112 BUSINESS GROUP "EV 1 4.2.5.4 Total P+Q Range. The total primary plus s~condary stress range for the selected locations is as follows: P+Q Location (i:s i) A 29.22 B 32.83 C 39.27 D 39.48 E 43.15 F 61.68 G 62.29 B 62.90 I 62.54 J 59.31 The limit for the primary plus secondary stress intensity is 3 S. At S50 o F, m carbon steel (SA-3S0 LF2) has a S of 18.6 ksi, and stainless steel (SA-351 m CP3) has a S of 16.0 ksi. In calculating 3 S , it is seen that Sections F m m through J are in excess of their limits, therefore, the simplified elutic-plastic method "i11 be used. For the simplified elastic-plastic fatigue analysis to be valid, the primary plus secondary minus thermal bending must be less than 3 S. So for Sections F through J, P+Q excluding thermal bending m will be calculated. The maximum thermal membrane stress ranges are found to be as follows: [Calculations included in DRF (Reference 6.5)] Thermal Membrane Location Range (ps i) Cases F 3,705 2-7 G 17,979* 3-9 B 16,964 1-9 I 13,669* 8-9 J 3,117 8-9 Therefore, the total P+Q stress intensity ranges excluding thermal bending are found to be: P+Q 3 Sm Location .full (ksO F 36.61 55.8 G 45.42 55.8 B 40.87 55.8 I 33.19 48.0 J 24.12 48.0 NEBG0807A (6/801
8-114 NUCLEAR ENERGY - GEN ERAL" ELECTRIC 22A7454 SH NO. 113 BUSINESS GROUP REV 1 4.2.6 Interference Fit Stresses, Since both male and female members are stainless steel. both member's respective moduli are ~qual From Shigley - Page 78 (Reference 6.14) p _ ES (c 2 _ b 2 )(b 2 _ a 2 ) b 2b 2 (c 2 _ a2) Now the dimensions (from Reference 6.13, 6.9. and 6.8) a = 2.71 in max & - 0.0065 in b = 3.19 in or o*013 in d ia
- c '" 3.69 in Also - yield highest stresses at room temperature 6
E - 28.3 x 10 psi (3.69 2 ~ 3.19 2 )(3.19 2 - 2.71 2 ) 2
= 4.401. psi 2(3.19)2 (3.692 - 2.71 )
Boop stresses for the inner and outer members are as follows: 2 2 b + a ae inner = -p b2 2 - -27,228 psi
- a 2 2 c + b a
e outer
= P c 2 _ b2
= 30.438 psi The total hoop stresses at this location includes a pressure effect.
ae - -2t PI>
= 111(6.16) = 684 psi 2(0.5)
By inspection, the total P+Q stresses for this location are less than the
.ll~wed 3 S limit (3 S - 48 ksi). This section will not be examined again.
m m NEBGsa07A (1/10)
8-115 22A7454 114 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. BUSINESS GROUP REV 1 TABLE 4.2-1 'IlIER."IAL MEMBRANE PLUS BENDING S'l1l.ESSES (CASE 1) Cool Do1l't1 (t = 6.0 seconds) All stresses in psi Inside Outside Section CJ~ CJ e CJ r CJ., CJ e CJ r A 2,679 2,390 -2,442 -2,165 B 2,892 2,972 -2,620 -2,659 C 2,723 2,662 -2,413 -3,880 D 2,748 2,596 -2,437 -3,953 E 9,820 169 -9,251 -9,885 F 28,589 21,490 -26,331 -20,696 G 16,087 29,153 -15,027 -7,305 B 27,845 38,999 -26,414 -5.073 I 28,682 15,876 -23,569 -43,017
- 1 26,741 31,274 -22.590 -34,654 Sections Illustrated in Figure 4.2.1-1 *
- NEBGoa07A (6/80)
8-116 22A7454 115 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. BUSINESS GROUP REV 1 TABLE 4.2-2 THERMAL MEMBRANE PLUS BENDING STRESSES (CASE 2) Cool Down (t = 15.0 seconds) All stresses in psi Inside Outside Soction 0'6 O'e 0' r 0'6 O'e 0' r A 2,532 2,223 -2,345 -1,933 B 2,381 2,615 -2,180 -2,150 C 1,226 1,370 -1,034 -3,756 D 1,243 1,230 -1,050 -3,894 E 16,142 -3,012 -15,293 -18,275 F 28,685 19,599 -26,724 -17,769 G 3,827 34,852 -4.,068 7,377 B 28,805 33,739 -27,954 -10,155 . I 29,154 20,294 -25,852 -39,015 1 29,003 34,673 -26,491 -38,310 Sections Illustrated in Figure 4.2.1-1. NEBG-807A (6/80)
8-117 NUCLEAR ENERGY GENERALe ELECTRIC 22A7454 SH NO. 116 BUSINESS GROUP REV 1 TABLE 4.2-3 'I'IIERMAL HEMBRANE PLUS BPNDING SntESSES (CASE 3) Cool Down (t = 60. seconds) All stresses in psi Inside Outside Section t1 r A 328 863 -291 -8 B 554 1,109 -517 103 C -3.992 -1,074 -3,866 202 D -4,168 -1,298 4,036 93 E 9,787 -8,278 -9,304 -16,283 F 16.829 8,280 -15,898 -7,339 G -4,453 26,80B 3.696 9,150 II 28,236 29.932 -27,688 -12,507 I 28,528 21.439 -26,195 -36,052
- 1 30,698 35,913 -29.186 -37.833 Sections Illustrated in Figure 4.2.1-1 *
- NEBGoa07A (6/10)
22A7454 NUCLEAR ENERGY GENERAL" ELECTRIC BUSINESS GROUP REV 1 TABLE 4.2-4 THERMAL MEMBRANE PLUS BElmING S'rn.ESSES (CASE 4) Cool Down (t = steady state) All stresses in psi Inside Outside Section
- a. a e ar a. a e a r
A -1.083 7S 1.050 655 B -1.084 766 1.054 1.308 C -8.213 -3.595 7.884 1.686 D. -8.452 -3,914. 8,114 1,579 E 6,431 -8,180 -6,247 -8,762 F 12,842 1,706 -12,215 -6,227 G 186 17,261 -545 99 B 27.722 30.407 -27.204 -11,813 I 28.097 21.945 -25.794 -35,367
;r 30.926 36.270 -29.414 -37.645 Sections Illustrated in Figure 4.2.1-1.
NEBGoa07A (6/80)
8-119 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 118 BUSINESS GROUP REV 1 TABLE 4.2-5 'I'BERMAL MEMBRANE PLUS BENDING STRESSES (CASE 5) Beat Up (t = 2.0 seconds) All streSle s in psi Inside Outside Section 116 l1e 11 r I1b l1e 11 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 -8,486 -5,005 -7,446 F 5,694 -4.237 -5,899 -8S5 G -5,412 10,737 4,408 3,684 B 22,221 25,867 -22,841 -7,180 I 22,531 14,531 -21,391 -30,069
]- 24.286 . 28,860 -24,354 -30,295 Sections Illustrated in Figure 4.2.1-1 *
- NEBG-807 A (6/10)
NEBG0807 A (6/110)
8-121 22A7454 NUCLEAR ENERGY GENE RA L
- ELE CTR I C SH NO. 120
. .~ __ _B_US_I_N_E_~_G RO_U_P________________________________________~A_E_V__1________________~ TABLE 4.2-7 l1IER)tAL HDmRANE PLUS BENDING S'rnESSES (CASE 7) Heat Up (t = 12.0 seconds) All * .;_)Sses in psi Inside Outside Section
"~ "e "r "~ "e "r A -2.030 -738 -. 1.927 1.439 B -2.015 -244 1.906 2.155 C -8,781 -4,249 8.375 3.056 D -9.024 -4.522 8,608 2,993 E 1,166 -7.155 -1,263 -2,599 F 167 -7,287 -415 1.709
- '~ G -2,698 4.763 2,319 -1,662 B 19.881 26,925 -20,309 -2.510 I 20.516 9.822 -18.572. -28.958 1 -19,335 24,115 -19.124 -24.777 Sections Illustrated in Figure 4.2.1-1.
~
NEBGoa07A (6/80)
8-122 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 121 BUSINESS GROUP REV 1 TABLE ... 2-8 'l"HERMAL HEHBRANE PLUS BENDING STRESSES (CASE 8) (Normal Operation) All stresses in psi Inside Outside Section a at> a O'~ O'e r O'e r A -3 -357 3 -302 B -610 209 593 549 C -3,615 -1,675 3,470 674 D -3,711 -1,812 3,563 619 E 3.456 -3,492 -3,349 -4,116 F 4,754 285 -4,521 -2,520 G -432 7,902 218 289 B 12,072 25,112 -13,090 5,539 I 12,769 -1,322 -10,602 -26,016 1 7,786 10,666 -7,411 -16,899 Sections Illustrated in Figure 4.2.1-1. NEBG-807 A (6/80)
8-123 22A7454 NUCLEAR ENERGY GENE RA L
- ELE eTR I e SH NO. 122 BUSINESS GROUP REV 1
----------------------~--------~
TABLE 4.2-9 'l'BERMAL MEMBRANE PLUS BENDING S'mESSES Pressure Ca.se All stresses -in psi No:ule Pressure = 1,111 psi Vessel Pressure = 1,000 psi Inside Outside Section a6 ae ar a6 ae ar A 1.127 4.963 -1,000 3,496 5,194 B 589 8,437 -1,000 4,276 8,618 C 1.037 7,941 -1,000 3,837 7,931 D 1,086 7,873 -1,000 3,790 7,846 E S.524 5,476 -1,000 -1,961 2,522 F 5.432 ',567 -1,111 4,393 6,492 G -1,430 3.615 -1,111 6,654 5,531 -1,000 H 3,189 -1,378 -1,111 2,294 -1,557 -1.000 I 3.289 -1,381 -1,111 2.194 -1,612 -1.000
- 1 4.962 -1,267 -1,111 658 -2,457 -1.000 Sections Illustrated in Figure 4.2.1-1.
- NEBGoa07 A (6/10)
NUCLEAR ENERGY GENE RAL fl,E LEe TRIC 22A7454 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 Xo is found from Figure A.7-1. since the scale stops at r/t = 3.6. that value ~ is assumed. xo = 1.24. Using Paragraph A.7.2.4. for r <h
~. _i =1 _ (~) 1 + 2.4 Vr/h o
Solving for X'. obtain X' = 1 + 0.24
~
NEBG-807A (6/aO)
8-125 22A7454 NUCLEAR ENERGY GEN ERAL. ELECTRIC SH NO. 124
- BUSINESS GROUP REV 1 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 Lnd outer surface will be calculated, however, only the largest will be used.
Assume r : o. hence X o
= 4.0.
Then using Paragraph A.7.2.4 (Reference 6). (X' - 1) 1-(X - 1) ... 1 - 90 o Solving for X'. using ~ inner = 72.3° and ~ ou t er .... 75.74° X'. ... 1.59 Ull1er X' ... 1.48 outer
*
- Xt '" 1.'9 i.
- NEBG-a07A (6/10)
NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. BUSINESS GROUP REV 1 T ".0" II 3.88 ". --L~-_ _-----l
+
S£c..'ION ~ G£om"::TA..,.., FlGURE 4.3-1 3.~J"'" -,
},..oo 1.8/"
/I 0.S"8S" T
~I -:: 7S-.7°
- 8. '38" fa:&. = 72..3 0 SECT/oIU F GEOmE.:r~ y FIGURE ... 3-2 .'
NEBGoa07A (6/10)
8-127 NUCLEAR ENERGY GENERAL. ELECTRIC 2.2.A74S4 SH NO. 126 BUSINESS GROUP REV 1 -.----------------------~--~----~ Sections E and G The gr..;:, -~~'Y of Sections E and G are illustrated in Figure 4.3-3. Also in Figure ~.3-3, is the idealized geometry used to obtain the factor. To obtain the stress concentration factor, Reference 6.6 will be used. Using Figure A.7-1 of Reference 6.6, obtain For Section E; r/t ~ 0.30 and for Section G; r/t m 0.50 Section J The geometry of Section J is illustrated in Figure 4.3-4. ANSYS. which was used to obtain the stress levels of the nozzle and safe end accounts for global discontinuities only. Therefore, only the local discontinuity stress concentration factor must be found. Conservatively. use Paragraph A.7.2.4 (Reference 6.6). Assume r : 0, and hence Ko = 4.0
- Then using Paragraph A.7.2.4 (Reference 6.6), obtain (K' 1) = 1 _ L (K - 1) 90 o
Solving for K', using ~ = 76.44° K' =1 + 3 (1 - 76 96 4 )
- NEBG0807A (6/80)
8-128 NUCLEAR ENERGY GENEJtAL. ELECTRIC 22A7454 SH NO. 127 BUSINESS GROUP REV 1 0.2.5" I, ~. 8.3&" IDE-ln... } ~E.D IlS ASs<.>m..e. c.o ")SER.. V.4 T/t) ~L.:r O* 2t For S£C.-rION Eo .. Fctt... SE:C.'T/o~ (:, r /c :a o. '30 r"/~ :a o. $"0 FIGURE 4.3-3 SECTION E AND G GEOMETRY FIGURE 4.3-4 SECTION J GEOMETRY NEBGoa07A (6,aO)
8:-129 NUCLEAR ENERGY GENERAL-. ELECTRIC 22A74S4 SH NO. 128 BUSINESS GROUP REV 1 4t----------------------~--------~ Sections B, C, Dr Hr and I All of these sections are at locations of welds. A conservative stress concentration factor will be put on these locations. For pipe jV1~v~ by butt welds, the thermal It is 1.8 (Reference 6.2). This will be assumed here. X t ""' 1.8 4.3.2 Alternating Stress Range. To calculate the alternating stress range, the following equation is used. where: It = stress concentration factor xe = simplified elastic-plastic factor SN - P+Q stress intensity 4t F1 ""' peak stress identified by 'NONO' program The peak stress identified by the 'NONO' program is f011l1d using Tables 4.3-3 through 4.3-11. These tables contain the total surface stresses at each section. A conservative method of obtaining these peak stresses is to find the largest stress intensity range for the surf~ce stresses and subtracting the P+Q stress intensity found earlier in Paragraph 4.2.S. This was followed and the results are given in Table 4.3-1. Also in Table 4.3-1 is the stress intensity range for the mechanical load case which docs not include seismic loads. These are needed to calculate the P+Q range excluding seismic, hence the alternating stress range excluding seismic is given. 4t NEBG-a07A (6/aO)
8-l30 22A74S4 NUCLEAR ENERGY GENERAL., ELECTRIC SH NO. 129 BUSINESS GROUP REV 1 4.3.2 (Continued) The total alternating stress range is calculated for both seismic loading included and seismic loading not included. The simplified elastic-plastic factor is defined by the following (Reference 6.2). 1.0 for SN < 3Sm 1 + ,.1; : ~l (~.. - ~ for 3S.. < SN < ..3S..
..here:
for carbon steel n = 0.2 III =3 for stainless steel n ... 0.3 III ... 1.7
~ ~ P+Q stress intensity range The results are tabulated in Table 4.3-2.
NEBG-a07A (6/80)
8-131 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 130 BUSINESS GROUP REV 1 TABLE 4.3-1 CALCULATION OF PEAK S'IllESS All .. *.Tesses in xsi The~l 'Skin' Stresses Included lIechalli cal Range Location (Seismic) Kechani c al Range (No Seismic)
'1l1emal Range Cases*
Pressure lange Huimum Surface Stress Range
~ ..
P+Q (F ) 1 Peak Stress A 18.54 6.85 5.93 1-7 *6.22 30.69 29.22 1.47 B 18.48 6.62 6.78 1-6 9.64 34.90 32.83 2.07 C 18.41 6.39 14.69 1-6 8.96 42.06 39.27 2.79 D 18.41 6.39 14.98 1-6 8.89 43.28 39.48 2.80 E 18.35 6.15 26.20 2-9 9.53 54.08 43.15 10.93 F 24.22 7.73 40.55 1-6 8.83 73.6 61.68 11.92 G 19.78 S .51 38.81 2-9 10.15 68.74 62.29 6.45 H 19.33 5.79 44.73 1-9 4.76 68.82 62.90 ~.92 I 14.85 4.44 50.94' 1-9 4.91 70.7 62.54 8.16 J' 14.77 4.49 55.52 1-9 7.79 78.08 59.31 18.77
- Cases showu in Table 4.2.5-1
** From Section 4.2.5.4
- NEBGoa07 A (6/80)
8-132 NUCLEAR ENERGY GENE RAL. , ELE CT_R I C 22A7454 SH NO. 131 BUSINESS GROUP AEV 1 TABLE 4.3-2 CALCULATION OF ALTERNATING S'IRESS All stresses in ksi
* ~+* ~:iJ I
- X
- S
- S
- Location ~1 2 X
t Stress e l e 2 alt 1 alt 2 A 29.22 17.53 1.21 1.47 1.0 1.0 18.42 11.34 B 32.83 20.97 1.8 2.07 1.0 1.0 30.6 19.91 C 39.27 27.25 1.8 2.79 1.0 1.0 36.8 26 .0 D 39.48 27.46 1.8 2.80 1.0 1.0 37.0 26 .2 E 43.15 30.95 2.15 10.93 1.0 1.0 51.9 38.8 F 61.68- 45.19 1.59 11.93 1.21 1.0 66.6 41.9 G 62.29 48.02 1.8 6.45 1.233 1.0 73.1 46 .S B 62.90 49.36 1.8 5'.92 1.255 1.0 74.8 47.4 I 62.54 52.13 1.8 8.16 2.01 1.287 . 121.4 65.7 1 59.31 49.03 1.45. 18.77 1.786 1.072 93.6 48.2
- Subscript 1 for seismic; Subscript 2 for no seismic loads
+ ~ found using Section 4.2.5 and Table 4.2.3.2-1 NEBGo807A (6/80)
8-133
- _N_UC_L_EA_R_E_N_E_R_GY BUSINESS GROUP _ _ _ _ G_E_N_E_R_A_L_0_,
_.' _E_L_E_C_T_R_IC ______L..2_2_A7_4.:.54 REV 1 _ _ _S_H_N_Oo_1_3_2--...J TABLE 4.3-3 'I'BERMAL 'SKIN' S'mESSES (P + Q + F1 ) (CASE 1) Cool Down (t = 6.0 seconds) All stresses in psi Inside Outside Section C1. C1 e C1 r C1., C1 e C1 r A 3,623 3,419 -1,873 -1,494 B 4,183 4,353 -1,797 -1,792 C 4,340 4,422 -1,292 -2,749 D 4.368 4,357 -1,317 -2.823 E 16,358 3,038 -7,253 -8,440 F 38.712 29,769 -21,225 -15,334 G 23,724 37,262 -10.544 -1.992 11 33,822 44,723 -27,107 -4,255 I 50,935 38.423 -4,123 -24,635 1 50,037 55,518 -5.889 -17,421 Sections Illustrated in Figure 4.2.1-1. '. NEBGoa07 A (6/~O)
8-134 22A7454 133 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. BUSINESS GROUP REV "I TABLE 4.3-4 'I'BER..'!AL ' SKIN' S'IRESSES (P + Q + F1)( CASE 2) Cool Down (t = 15.0 seconds) All stresses in psi Inside Outside Section 0' <19 <1
<16 0'9 r 0'6 r A 3,113 2,877 -1,992 -1,470 B 3,206 3,518 -1,602 -1.531 C 2,230 2,558 -128 -2,882 D 2.250 2.418 -140 -3.019 E 26.195 5 -12.272 -16.508 F 36,613 25,636 -23,420 -13,914 G 7,629 38,809 -75 9,598 B 29,686 35,301 -33,577 -13,094 I 39,914 31,376 -17,613 -31,359
- 1 40,400 46,428 :"19,565 -31,097 Sections Illustrated in Figure 4.2.1-1.
_ . NEBGoa07A (6/80)
8-135-NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO* 134
- BUSINESS GROUP REV 1 TABLE 4.3-5 THERMAL 'SKIN' STRESSES (P + Q + F1 ) (CASE 3)
Cool DoYll (t = 60. seconds) All stresses in psi Inside Outside Section CJ~ CJ e CJ r CJ~ CJ e CJ-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 -6,3il G -2,865 28,192 1,709 9.610 B 26,786 29,469 -35,539 -17,295 I 33,583 26,665 -23,875 -34,231 1 35,206 40.230 -29,417 -37,795 Sections Illustrated in Figure 4.2.1-1 *
- N£BGoa07A (6/aOl
---1.'
8-136 NUCLEAR ENERGY GENE RA L . , ELEe TRIC 22A7454 SH NO. 135 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _..!.."'_E_V_l~_ _--:-_ _ _ _ _BU_S_I_N_E_SS_G_R_O_UP TABLE 4.3-6 THERMAL ' SKIN' S'mESSES (P + Q + F ) (CASE 4) 1 Cool Down (t = Iteady state) All stresses in psi Inside Outside Section CJ. CJ e CJ r CJt> CJ e CJ r A -1.123 42 1.160 649 B -1.158 722 1.112 1.321 C -8,705 -3.860 8.344 1,738 D -8.961 -4.203 8.593 1,631 E 11.513 -8,565 -5.060 -9,135 F 14.726 2.289 ':"12.418 -6,132 G 1.099 18.047 1.083 -166 B 26,237 29,924 -3'5,042 -16,606 I 33.098 27,126 -23.496 -33.584 1 35.424 40,531 -29.711 -37.67-4 Sections Illustrated in Figure 4.2.1-1. NEBG.a07A (6/80)
8-137 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 136 BUSINESS GROUP REV 1 TABLE 4.3-7 'IBERMAL 'SKIN' STRESSES (p. + Q + F ) (CASE 5) 1 Heat Up (t = 2.0 seconds) All str -ll in psi Inside Outside Section ert ere er r er., ere er 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 8.835 -9,759 -4,209 -8,158 F 2,762 -7.988 -8,384 -3,018 G -8,872 7.052 3,722 816 H 14,115 20,018 -34,578 -15,131 I 20,488 11,739 -23,383 -32,975 1 17,903 22,761 . -30,218 -35.591 Sections Illustrated in Figure 4.2.1-1. NEBGoa07A (6/10)
8-138 NUCLEAR ENERGY GENE RA L . , ELEe TRIC 22A7454 SH NO. 137 _B_U_SI_N_E_SS__ G_R_O_U_P________________________________________~R_EV__~1____----------~-J4It TABLE 4.3-8 'I'RElUlAL ' SKIN' STRESSES (P + Q + F 1 ) (CASE 6) Heat Up (t = 3.0 seconds) All stres ses in psi Inside Outside Section a41 a e a r ab ae a r A -2.278 -1,029 1,591 1,048 B -2,592 -763 1,599 1,813 C -10.347 -5,531 8,800 2,485 D -10,611 -5,861 9,056 2,395 E 7,369 -9,907 -3,462 -7,280 F -1,128 -10,780 -5,368 -752 G -10,334 4,345 5,903 1,739 4Iti B 11,751 18,369 -33,126 -13.254 I 18.842 8,825 -21.948 -31.365
'1 14.040 19.115 -29,368 -33,807 Sections Illustrated in Figure 4.2.1-1.
4It NEBG.807 A (6/80)
- NEBG0807A (6/80) fBi ELECTRIC 8-140 22A7454 NUCLEAR ENERGY GENERAL SH NO. 139 BUSINESS GROUP REV 1 TABLE 4.3-10 ~~ 'SKIN' STRESSES (P + Q + F ) (CASE 8) 1 (Normal Opera Hon) All stresses in psi Ins ide Outside Section (1., (19 (1 (10 (19 (1r r
A -11 -364 2 -303 B -648 187 625 556 C -3,833 -1,791 3.673 698 D -3,936 -1.936 3,773 645 E 5,932 ~3,610 -2,775 -4,261 F 5,453 483 -4,590 -2.492 G 15 8,293 1,181 214 11 5,573 20.891 -22,570 -758 I 20,091 4.523 -4,088 -21.360
] 8,568 12,150 -7.694 -16,972 Sections Illustrated in Figure 4.2.1-1.
NEBG.a07A (6/801
8-141 NUCLEAR ENERGY GENERAL" ELECTRIC 22A7454 SH NO. 140 BUSINESS GROUP TABLE 4.3-11 HEMBRANE PLUS BENDING PLUS SKIN S'IRESSES (P + Q + F ) 1 (Pressure Case) All stresses in psi Ins ide Outside Section Cf~ Cf e Cf r Cfb Cf e Cf r A 1,052 4.978 -1,000 3,632 5.220 B 436 8.436 -1.000 4.394 8.637 C 931 7,954 -1.000 3.923 7.946 D 981 7.886 -1.000 3.872 7,859 E 8,103 5.982* -1.000 -1.426 2.662 F 6,096 7,719 -1.111 4.690 6,545 G -969 3.700 -1,111 9, 036 5,931 -1,000 H 3,202 -1,367 -1,111 2.284 -1,551 -1.000 I 3,290 -1.376 -1,111 2.152 -1,620 -1,000 J 5,311 -1,231 -1,111 543 -2.479 -:1,000 Sections Illustrated in Figure 4.2.1-1.
- NEBGoa07A (6/80)
8-142 454 NUCLEAR ENERGY GEN ERAL" ELECTRIC SH NO. 1 BUSINESS OPERATIONS REV 1 4.3.3 Usage Calculations. Using the total alternating stress levels found in Table 4.3-2, the fatigue usage factor can be solved. From the design specification (Reference 6.1), there are 1,500 thermal cycles and it is assumed there are 10 seismic cycles. Using Figures I-9~2 and I-9.1 from Reference 6.2. the fatigue usage factors are found. Table 4.3.3-1 contains these calculations. Notice, the stress ranges must be adjusted to the elastic modulus of the fatigue curves. The total usage factors are as follows. Location Fatigue Usage A 0.0016 B 0.0257 C 0.0609 D 0.0609 E 0.233 F 0.2789 G 0.4065 H 0.4074 I 0.2006
] 0.0556 NEO 107.11 (REV. 10(111
fBi ELEe TRIC 8-143 22A74S4 142 NUCLEAR ENERGY GENE RAL SH NO.
- BUSINESS OPERATIONS REV 1 TABLE 4.3.3-1 CALCULATION OF FATIGUE USAGE Location (hi)
S aIt1
- Elastic Modulus Factor S,\
(ksi) Allow. Usage Cycles Factor S a1t (ksi) 2
- Elastic Modulus Factor
- S~
(:t.si) Allow. Usage Cycles Factor lOS 6 A 18.42 1.042 19.2 0.0001 11.34 1.042 11.9 10 0.00149 B 30.6 1.103 33.8 13,000 0.0008 19.91 1.103 22.0 60,000 0.0249 C 36.8 1.103 40.6 8,000 0.0013 26.0 1.103 28.7 25,000 0.0596 D 37.0 1.103 40 .9 8,000 0.0013 26.2 1.103 28.9 2S ,000 0.0596 E 51.9 1.103 57.3 2,750 0.0037 38.8 . 1.103 42.8 6,500 0.2293 F 66.6 1.103 73.5 1.250 0.008> 41.9 1.103 46.3 5.500 0.2709 G 73.1 1.103 8>.7 1,100 0.0091 46.5 1.103 51.3 3,750 0.3974 B 74.8 1.103 82.6 1.000 0.0100 47.4 1.103 52.3 3,7SO 0.3974 I 121.4 0.968 119.' 700 0.0143 65.7 0.968 64.8 8,000 0.1863
- r 93.6 0.968 92.3 1,700 0.0059 48.2 0.968 47.6 30,000 0.0491
- Subscript 1 for Seismic; Subscript 2 for no seismic loads.
** For carbon steel fatigue curve, E = 30 x 10 6
psi; for stainless steel fatigue curve, E = 26 % 10 psi 6 NED 107A (REV. 10/1t)
8-144-22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 143 BUSINESS GROUP REV 1 4.3.4 Digh Cycle Fatigue, The cumulative high cycle fatigue usages due to rapid cycling "as determined to be as follows (Reference 6.12). U = 0.000063 4.3.5 Accumulated Fatigue Usage. The exUsting nozzle and remaining part of the safe end have accumulated a fatigue usage of 0.22 (Reference 6.1). 4.3.6 Total Fatigue Usage. The total fa Ugue usage is as fo11oys. System Digh Existing Total Location Fatigue Cycle Fatigue Usage A 0.0016 0.0001 0.22 0.2217 B 0.0257 0.0001 0.22 0.2458 C 0.0609 0.0001 0.22 0.2810 D 0.0609 0.0001 0.0610 E 0.233 0.0001 0.2331 F 0.2789 0.0001 0.2790-G 0.4085 0.0001 0.4086 B 0.4099 . 0.0001 0.4100 I 0.2006 0.0001 0.2007
;r 0.0556 0.0001 0.0557 NEBG0807A (6/80)
8-145 NUCLEAR ENERGY GENERALe ELECTRIC ZZA74S4 SH NO. 144
- BUSINESS GROUP REV 1 5.. RESULTS Detailed stress analysis of the feedwater nozzle and safe end assembly shows that the nozzle and assembly fully meet the ASHE Code (Reference 6.Z> stress intensity limits. Further, the nozzle and assembly are shown to~sat"isfy the requirements for cyc.lic operation, with the maximum cUlllulative fatigue usage determined to be:
6* REFERENCES 6.1 'Reactor Vessel System Cycling', Design Specification 22A6996, Rev. O. 6 *.2 ASME Boiler and Pressure Vessel Code, Section III, Subsections NA and NB, 1977 Edition including Addenda through Summer 1978. 6.3 'ANSYS. Engineering Analysis System, User's Manual,' Swanson Analysis Systems, Inc. 6.4 Xreith, F ** 'PrinCiples of Heat Transfer'. Third Edition. IEP Publisher, 1973. 6.S DRF# B13-909. Monticello Feedwater Nozzle Stress Analysis Design Record File. 6.6 'Tentative Structural Design Basis for Reactor Pressure Vessels and
.D.irectly Associated ~omponents', 1958.
6.7 GE Drawing No. 769ES31. Rev. O. As-Built Feedwater Nozzle. 6.8 GE Drawing No. 10SD6009, Rev. 2. Thermal Sleeve 6.9 GE Drawing No. 137C7841. Rev. 1, Reducing Tee 6.10 GE Drawing No. PDS-3108, Layout. 6.11 GE Drawing No. 137C7843. Rev. 2, Safe End. 6.12 Reactor Vessel Rapid Cycling. GE Document No. 22A7227, Rev. O. 6.13 GE Drawing No. 11202892, Rev. 3. Safe End Assembly. 6.14 I.E. Shigley. Mechanical Engineering Design, McGraw-Hill. Second Edi tion. 1972
- NEBGoa07 A (6/80)
8-146 NUCLEAR ENERGY - GENERAL" ELECTRIC* 22AHS4 SH NO. *145 BUSINESS GROUP REV 1 APPEm>IX lOLl STI~ OF ' NONO ' 10 REAL ~EMS1P
- 0 REAL HEUS~~
30 REAL NUAVGS 40 REAL NUDlS 50 REAL MOMARM 60 REAL MOMENT 70 DIMENSION NELM(2S).CORDX(:7).STRS(27).DISAVG(:7).5T~VG(~:).TST~(:~ 80& HEUSTR(2?).NUAVGS(2?).NU~IS(:7).MOMARM(27).MOMENT'~7).~OU~T'99~; 90 PRINT:"" 100 PRINT:"" 1 10 PRINT:"" 120 PRINT:"DO YOU NEED INSTRUCTIONS' (1 =YES. O::NO';" 130 REArl: NOYESO 140 IF(NCYESO.EQ.O) GOTO 10 150 PRI~T:"" 160 PRINT:"" 17Q PRINT:"" 180 PRINT:"" 190 PRINT:"THIS PROGRAM LINEARIZES THE STRESSES THROUGH THE" 200 PRINT: '"I 210 PRINT:"THIC.~NESS OF A SECTION." 220 PRINT:"" 230 PRINT:"INPUT THE LOCATION OF THE POINTS UHERE STRESSES ARE" 2~0 PRINT:"" 250 PRINT:"ACTING, I.E. THE FIRST SURFACE. THE CENT~OlDS AND" 260 PRINT:"" 270 PRINT:"THE SECOND SURFACE." 280 PRINT:"" 290 PRINT:"INPUT STRESSES AT THE FIRST SURFAFE. CENTROIDS AND n 300 PRINT:"" 31(i PRINT:"THE SECOND SURFACE." 320 PRINT:"" 330 PRINT:" IF N IS THE NUMBER OF ELEMENTS THEN TOTAL STRESS AN[I" 3~0 PRINT:"" 350 PRINT:"LOCATlON ENTERIES REQUIRe:rl ARE N+2. tI.E. 2 SURFACE~"
.360 PRINT:""
370 PRINT:"AND N CENTROIDS.)" 38'j PRINT:"" 390 PRINT:"THE PROGRAM THEN COMPUTES THE MEMBRANE STRESS BY" 400 PRINT: .... 410 PRINT:"THE EQUIVALENT AREA METHOD." 420 PRINT: .... 430 PRIHT:"THE BENDING STRESSES ARE COMPUTED BY LINEARIZING THE" 4 ~O PRINT:"" 450 PRINT:""O~ENT ACROSS THE SECTION THICKNESS." 460 PRINT:"" 47(' PRINT:"PEAK STRESSES ARE THE TOTAL STRESS MINUS THE" 480 F'RItlT:"" 49.) f'F:l,.,:,: "I\E~HANE AN!: FEND!NG _STR~SSES AT THE SURFACES." 500 PRINT:*" 501 PRINT:""
8-147 NUCLEAR ENERGY GENE RA L
- ELEe TRIC' 22A7454 BUSINESS GROUP REV 1 APPENDIX 10 (Continued)
- 0
- PRINT:"*
%3 f'~:: ";.~ : ",.
510 PRINT:"UHE~ FINlSHEt iNFU1 HUK~tR OF E~En~~:S =0 T~ ~liT' 5:0 PRINT:""
'530 PRINT:""
540 lOP RI NT: ,,,. 55,:; F:RINT: "" 560 PRINT:"UANT A LIS7ING OF THE INPUT STRS & COORDS., (1=YES.O=~OI* 570 READ: NOYES 580 DO 300 L=1, 999 590 PRINT:" " 600 PUNT:" J' 61C NSTRS=O 6~O f"R!Ni:" IHF'Ui NO. OF ELEM ACROSS THeK " 630 READ: NEL 640 IF(NEL,LE.O)GOTO 301 eSO LOCATE=NEL+2 660 KOUNHL )=L 670 IFCKDUNT(L).ED.l) GOTO 11 680 PRINT:"" 690 F'RINT:"" 700 PRINT:"USE COCRDS. FROM THE PREVIOUS RUN?? (YES=1,NO=O)"
- "'10 READ: NOYES1 720 IF(NOYES~.ED.O) GOTO 11 no GOTD 12
~"40 11 PRINT: " INPUT COORD. LOCATIONS Xl,X2,o ** ETC. ~
. 750 READ: (CORDX(I) ,1=1 ,LOCATE) 760C 770 12 PRINT:" INPUT CORRES STRESSES" 780 NSTRS=LOCATE 790 READ: (STRS(J).J=1,NSTRS) 800 TOTSTR=O.
810 TOTIlIS=O. 820 DO 100 K=1,CNSTRS-l) 830 STAVG(K)=(STRS(K)+STRS(K+l>>/2.0 840 DISAVG(K)=(CORDX(K+l)-CORDX(K>> 8S0 TSTR(K)= STAUG(K)*DISAUG(K) 860 TOTSTR=TOTSTR+TSTR(K) 870 TDTDIS=TOTDIS+DISAUGlK) 880 100 CONTINUE 8% MEIISTR=TOTSTR!TOTDIS 90C TOnOIl=.O. NE8G407A (6"0)
8-148 NUCLEAR ENERGY ---~-G ENE RAL* ELEe TRIC llA7454 SH NO. 147 BUSINESS GROUP REV 1 APPENDIX 10 (Continued) 910 DO 101 I=l.NSTRS 910 101 ME~STR(I)=STRS(Il-"EMST~ 93C- DO 102 1=1, (HSTRS"-I i 940 NUAVGS(I)=(NEUSTR(I)+NEUSTR(I+17)/2.0 NUDIS(I)= CORDX(I+l)-CORDX(I) 960 "Q~AR"(I)=(CORDX(I)-CORDX(l>>+(CORDX(I+ll-CORDX~I)I!:. 9;"(: HOMENT(I)=NUAVGS(ll*NUDIS(I)~"OMARN(I) 960 TOTMOM= TOTMOH+HONENT(I) 990 102 CONTINUE
, COO SBEND=(6~TOTHOHI/(TOTDIS~TOTDIS) 1010 SBENDI=A~S(SBENDl 1020 SBENI:2=SBEN!llr (-1 .)
, O~1 SBEN!I=ABS (SBDm 1
, 03-~ IF(STRS(l I.LT.STRSINSTRSl) GO TO 40C
": C40 PEA~Sl=STR5(II-HEMSTR-SBENDl
~ 050 PEAKS2=STRS(NS1~SI-HEHSTR-SBEND:
1 06':: GO TO 401 " 10iO 400 PEARS 1=STRS (1 ) -MEMSTR-SBEND2 1060 PEAKS2=STRS(NSTRS1-MEMSTR-SBEHDI 1090 401 IF(NOYES.EG.OI GOTO 20 1100 PRINT:" "
~ i 10 PRINT:"lNPUT STRESSES ARE:"
1120 PRINT:" "
" 1 130 ~RITE(06,202)(STRS(I).I=1,NSTRS) 1140 PRINT:""
11 SO PRINT:"" 1160 PRINT:"INPUT COORD. ARE:" 1170 PRINT:"" 1 t 80 URITEt06.203)(CORDX(I),I=1.LOCATE)
'190 20 URITE(06,:00) HEHSTR,SBEHD.PEAKS1,PEAKS2 1200 300 CONTINUE
"' ~1 0 301 URITE(06,204)
'220 200 FORMAT(//.4X."HEHBRANE STRESS = ",Fl0.l.4X.
, 23C & *BENDING STRESSES =(+ OR-) ",Fl0.1.lI.4X.
, 240~ ,"PEAKSI =".FIO.I,4X."PEAKS2 =".Fl0.1.;;)
~ 250 202 FOR~AT(4X.4F1C.l./)
126":> 203 FORMAT(4X,4Fl0.3./)
~ :70 204 FORMAT(//," HAVE A NICE DAY!! ".'" I}
, 26(' STOf"
, 2~"0 ENt:
NEBGoe01A (6,.0)
8-149 NUCLEAR ENERGY GEN ERAL., ELECTRIC llA74S4 SH NO. 148
- BUSINESS GROUP Rev 1 APPENDIX 20 INTERGRANULAR S'nESS CORROSION INDEX CALCULATION The areas requiring an IGSCC stress rule index calculation are Sections C/D and B/l. These sections are at new weld locations.
10.1 Sections C and D Q + F + RESID S.l. = S + O.002E y Section C is SA-S08 (Class 1) carbon steel and Section D is SA-3S0 (LF2) carbon steel. 20.1.1 Mechanical Load Stress. The ~echanical load stresses are the result of dead weight, thermal, and hydrauliC loads during normal operation. The loads are obtained from Reference 6.1, and are as follows. No%zle Loads Dead Weight (Nonle 'A' Loads used>. F x - -0.11 kip ID x
= +11.6 in-kip F
Y
= -0.63 kip m y = -14.1 in-kip Fz = +0.15 kip m z = -11.1 in-kip Thermal (Noule 'B' Loads used)
F x
= +().82 kip m x
= +267.2 in-kip F
y
= -4.34 kip m y = +66.7 in-kip F
z
... +1.37 kip m
= +1.4 in-kip
- NEBGoa07A (6110)
8-1 NUCLE~R ENERGY GENERAL. ELECTRIC 22A74S4 SH NOo 149 BUSINESS GROUP REV 1 20.1.1 (Continued) Thermal Sleeve Loads Dead Weight Plus Hydraulic Fy = -0.3 kip Fz = -3.0 kip m x*
= -1.2 in-kip Thermal F
z
= -1.2 kip The same procedure used in Section 3 will be used here. If more detail is required. refer to Section 3 foroa reference.
P _ (F 2 + F 2)1/2 x y m = (m 2 + m 2 + m 2)1/2 x y z Nozzle Loads Dead 'Weight P = 0.64 kip m = 21.4 in-kip Fz = 0.15 kip Therm~J. p - 4.42 kip
= 275.41 m in-kip F
z
= 1.37 kip NEBG-a01A (6/80)
NUCLEAR ENERGY GEN ERAL. ELECTRIC 22A74S4 SH NO. 150
- BUSINESS GROUP "ev 1 20.1.1 (Continued)
Thermal Sleeve Loads Dead Weight plus Hydraulic P ... 0.3 kip III ... 1.2 in-kip F ... 3.0 kip z Thermal Fz = 1.2 kip Section C and D Nozzle Load Stresses Dead Weight III ... 21.4 + 0.64(7.47) + 0.15 (0.56) ... 26.27 in-kip
=Jaz = 26.27 CJ BEID 51.98 = 0.506 ksi F
z 0.15 CJl l . =-;:- = 19.03 = 0.008 ksi Thermal III = 275.41 + 4.42(7.47) + 1.37(0.56) = 309.2 in-kiP. 309.2 CJ BEID = -IIIz = 51.98
= 5.95 ksi F
= L1L z
19.03 = CJl l .
= - 0.72 ksi A
- NEBG-807A (6/110)
8-152 NUCLEAR ENERGY GENERAL. ELECTRIC 22A74S4 SH NO. 151 BUSINESS GROUP REV 1 20.1.1 (Continued) Thermal Sleeve Stresses Dead Weight plus Hydraulic m = 1.2 + 0.3(21.12) + 3.0(2.36) = 14.62 in-kip 14.62 aBEND = -mz = 51.98 = 0.282 ksi F
--L all. = A = 19.03 =
.J 0.158 ksi Thermal m = 1.2(2.36) = 2.84 in-kip m
a BEND = -z - 0.055 ksi F a AI * = ~z .= 0.063 ksi Total Stresses. Primary stresses are dead weight and hydraulic stresses, while secondary stresses are thermal load stresses. Primary Stress Secondary Stress NEBG-807.A (6/80)
8-153 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 152 ~ _B_U_S_IN_E_SS __G_R_O_U_P__________________________________________~_~_E_V__1__--______________-J 20.1.2 Pressure Stress Prim ~~essure Stress
= Pri = 1,000(10,875)
.. 10.24 ksi t 2(0.531) 5.12 ksi a
r
= -1,000 psi Secondary Pressure Stress. The stresses in Table 4.2-9 are the P+Q pres~ure stresses corrosion not included. The primary stress (corrosion not included) is expected to be as follows.
a e = 8,600 psi a~ = 4.300 psi The actual stress from Table 4.2-9 is given as Section C Section D a e = 7.931 psi ae = 7.846 psi
- a. = 3. 837 psi a~ = 3.790 psi Therefore. the secondary pressure stresses are as follows.
Section C Section D a e = -669 psi O"e = -754 psi
- a. = -463 pai O"~ = -510 ps i
~ NEBG-B07A (6/80)
8-154 NUCLEAR ENERGY GENE RAL* E. LEe TR I C 22A7454 SH NO. 153 BUSINESS GROUP REV 1 20.1.3 Thermal Stresses. The thermal stresses given in Table 4.2-8 for steady state normal ope~ation are: Section C Section D
<1e -= 674 psi C1 e = 619 psi 20.1.4 Peak Stresses. The peak stresses are obtained for both the pressure and thermal eases by comparing Tables 4.2-8 and 4.2-9 to 4.2-17 and 4.2-18.
respectively. The total peak stress is the addition of these two values. Section C Section D C1 e = 203 + 86 = 289 psi C1 e = 210 + 82 = 292 psi
<1. = 24 + lS = 39 psi C1~ = 26 + 13 = 39 psi
- The peak stresses due to mechanical loads are small and, therefore, are not included.
20.1.S Index Calculation Stress S11lIIIIIary:
* +/- refers to locations 180 0 apart Primary Stresses Loading Kechanical Load :!:. 0.954*
Pressure 10.24 5.12 -1.0 TOTAL 10. "'.4 6-.074/4.166 -1.0 NEBGoa07A (6/80)
8-155 22A7454 NUCLEAR ENERGV GEN ERAL., ELECTRIC SH NO. 154
- BUSINESS GROUP ~EV 1*
20.1.5 (Continued) Secondary Stresses Loading C1 S k5i C1 p (ks i) Sections (C) I (D) (C) I (D) Mechanical Load +/-. 6.14* Pressure -0.671=0.76 -0.461-0.51 Thermal 0.671 0.62 3.471 3.57 Section C Section D TOTAL C1 S
=- C1 S = -0.14 k5i C1~ = 9.15/-3.13 k5i C1~ = 9.2/-3.08 k$i Peak Stresses Section C Section D C1 S
... 0.04 psi C1~ = 0.29 psi For, SA-3S0 (FL2) Carbon Steel S ... 27.85 ksi Y
E ... 26.0 % 103 ksi RESID = 37 ksi SA-SOS (Class 1) Carbon Steel S D 27.1 ksf Y 3 E ... 26.0 % 10 ks i. RESID = 37 ksf NEBG-a07 A (6,10)
8-22A7454 NUCLEAR ENERGY GENERAL" ELECTRIC SH NO. ISS BUSINESS GROUP REV 1 20.1.5 (Continued) For Section C Based on Stress Intensities
~ The two possible stress intensities are based on the folloYing stress components:
Comb 1 9.44 (Q+F) lila.%.
= 9.44 lsi 2 0.04 -2.84 Therefore, S.l. = 11.24 = 9.44 + 37 27.1 27.1 + 52
- Oo4I4~ + 0.5871 = 1.0019 .',1I Now, since S.l. > 1.0, allowed, the S.I. is recalculated based on positive principle stresses.
Based on Positive Principle Stre~ses Q + F = 9.44 lsi S.l. = 10.24 = 9.44 + 37 27.1 27.1 + 52
= 0.3779 + 0.S871 = 0.96S < 1.0 allowed By inspection, the S.l. at Section C is higher than that at Section D.
NEBG-a07 A (6/80)
8-157 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 156
- BUSINESS GROUP "EV 1 20.2 Sections R and I, Section H is SA-350 (LF2) carbon steel and Section I is SA-351 (CF3) stainless steel.
20.2.1 Section R 20.2.1.1 Mechanical Load Stress, The mechanical load stresses are the result of dead weight, thermal, and hydraUlic loads during normal operation. The loads are obtained from Reference 6.1. Note, in Paragraph 20.1.1 of this appendix, the mechanical loads were reduced to: Nozzle Loads p = 0.64 kip (Dead Weight.) m = 21.4 in-kip F z = 0.15 kip P = 4.42 kip (Thermal) m = 275.41 in-kip F .. 1.37 kip z Thermal Sleeve Loads p = 0.3 kip (Dead Weight and Hydraulic) 1Il ... 1.2 in-kip Fz = 3.0 kip F z
= 1.2 kip (Thermal)
The same procedure used in Section 3 will be used here. If more detail is required, refer to Section 3 for a reference. NEBG-807A (6/80)
8-158 NUCLEAR ENERGY --- GENE RAL* ELEe TRIC 22A74S4 SH NO. 157 BUSINESS GROUP REV 1 20.2.1.1 (Continued) Thermal Sleeve Load Stresses Dead 'We ight It = 1.2 + 0.3(20.12) = 7.24 in-kip C7 = -Itz = L1L = 0.325 kai BEND 22.32 F z 3 C7 =- = = 0.287ksi AX A 10.46 Nozzle Load Stress Dead We ight It = 21.4 + 0.64(8.47) + 0.15(1.8) ~ 27.09 in-kip C7 BEND = 0.381 ~ = <0.381) i~:~; = 0.463 ksi (0.381) 0.15 = 0.006 ksi 10.46 Thermal m = 275.41 + 4.42(8.47) + 1.37(1.8) = 315.32 in-kip 315,32 C7 BEND
=, 0.381 !!
z = (0.381 ) 22.32 = 5.39 ksi F C7}J. = 0.381 Az = (0.381) 1.37 10.46 = 0.050 ksi Total Stresses, Primary stresses are dead weight and hydraulic stresses, while secondary stresses are thermal load stresse _, __,_ Primary Stresses
= 1,081 ksi Secondary Stresses NEBGoa07A (6/80)
8-159 NUCLEAR ENERGY GENE RAL* ELE CT RIC 22A7454 SH NO. 158
~ _B_U_S_IN_E_S_S_G_R_O_U_P__________________________________________~_~_E_V__1__________________-J 20.2.1.2 Pressure Stresses Primary Pressure Stress PD. 111(8.505) e = 2t = =
a ---! 1.26 ksi 2(0.375) O'e a~ = 2 = 0.63 ksi a r
= -0.111 ksi 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:
- 0.93 ksi
.~ O'~ == 0.465 ksi The actual stress from Table 4.2-9 is given as: O'e = -1.38 ksi Therefore, the secondary pressure stresses are as folloys: O'e = -2.31 ksi 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: O'e - 25.12 ksi ~ NEBG-a07A (6/80)
8-160 NUCLEAR ENERGY ,-, GENE RAL . , ELEe TRIC 22A74S4 SH NO. 159 BUSINESS GROUP REV 1 20.2.1.4 Peak Stresses, The peak stresses are obtained for both the pressure and thermal cases by comparing Table 4.2-8 and 4.2-9 to 4.2-17 and 4.2-18 respectively. The total peak stress is the addition of these t~o values, a e = -4.22 - 0.011 = -4.231 ksi at = -6.5 + 0.013 = -6.49 ksi The peak stresses due to mechanical loads are small and therefore are not included. 20.2.1.5 Index Calculation Stress Summary:
- +/- refers to locations 180 0 apart Primary Stresses at! (ks i) ar (X5 1)
Loading Mechanical Load +/- 1.081- --' Pressure 1.26 0.63 -0.111 TOTAL 1.26 1. 711/0.451 -0.111 Secondary Stresses Load ing at! (ks 1) Kechanical Load +/- 5.555* Pressure -2.31 2.73 Thermal 25.12 12.08 22.81 20.365/9.255 NEBG-807 A (6/aO)
8-161 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 160 BUSINESS GROUP "EV 1 20.2.1.5 (Continued) Peak Stresses C1 e = -4. 23 1 .k.s i C1~ = -6.49 ksi For SA-350 (LF2) Carbon Steel S = 27.85 ksi Y 3 E = 26.0 x 10 ksi RESID = 39.5 ksi Therefore, based on stress intensities SI = 1.822 + 22.81 - 4.231 + 39.5
. 27.85 27.85 + 52
= 0.0655 + 0.7274 = 0.7929 < 1.0 allowed 20.2.2 Section I 20.2.2.1 Xechanical Load Stress Thermal Sleeve Load Stresses Dead We ight m = 1.2 + 0.3(20.12) = 7.24 in-kip 7.24
= -mz = 29.07 Fz 3 C1 AX =~ = 13.78 = 0.218 ksi Thermal F
C1 AX = --! A
= 1.2 13.78
= 0.087 ks i NEBG-807A (6/110)
8-162 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 161 BUSINESS GROUP REV 1 20.2.2.1 (Continued) No%%le Load Stresses Dead Weight m = 21.4 + 0.64(8.47) + 0.15(1.8) = 27.09 in-kip 27,09 a BEND = 0.381 !! z = (0.381 ) 29.07 = 0.355 ksi F 0.15 all = 0.381 A% = (0.381) 13.78 = 0.005 ksi Thermal m = 275.41 + 4.42(8.47) + 1.37(1.8) = 315.32 in-kip 315.32 a BEND = 0.381 !! z = (0.381 ) 29.07 = 4.133 ksl-F 1.37
= 0.381 A = =
all (0.381) 0.038 ks i 13.78 Total Stresses Primary Stresses
= 0.827 ksi Secondary Stresses 20.2.2.2 Pressure Stresses Primary Pressure Stress 111(8.386) = 0.943 ksi 2(0.494) 0.471 ks i a
r
= -0.111 ksi NEBG.807A (6180)
8-163 NUCLEAR ENERGY G.E NERA L* ELECTRIC 22A7454 SH NO. 162
.BUSINESS GROUP IIIEV 1 20.2.2.2 (Continued)
.: ~ndary Pressure Stress. The stresses in Table 4 .2-9 are the P+Q pressure stresses corrosion not included. The primary stress (corrosion not included) is expected to be:
C1 e= 0.93 ksi 0.465 ksi The actual stress from Table 4.2-9 is given as:
"e = -1,381 pai
"~= 3,289 psi Therefore, the secondary pressure stresses are as follows:
"e == -2.311 ksi 2.824 ksi 20.2.2.3 Thermal Stresses, The thermal stresses given in Table 4.2-8 for steady state normal operation are:
"e = -1.322 ksi 12.77 ksi 20.2.2.4 Peak Stresses. The peak stresses are obtained for both the pressure and thermal case by comparing Tables 4.2-8 and 4.2-9 to 4.2-17 and 4.2-18 respectively. The total peak stress is the addition of these two values.
"e = 5.845 + 0.005 = 5.85 ksi "t a 7.323 + 0.001 = 7.324 ksi 20.2.2.5 Index Calculations Stress Summary:
* + refers to locati~ns 180 0 apart NEBG-807A (6/110)
8-164 NUCLEAR ENERGY GENERAL." ELECTRIC 22A7454 SH NO. 163 BUSINESS GROUP REV 1 20.2.2.5 (Continued) Primary Stresses Loading Xechanical Load +/- 0.827* Pressure 0.943 0.471 -0.111 TOTAL 0.943 1.298/-0.356 -0.111 Secondary Stresses a Loading r O::s i) Mechanical Load +/- 4.258 Pressure -2.311 2.824 Thermal -1.322 12.77 TOTAL -3.633 19.852111.336 Peak Stresses CJ s ... 5.85 ksi CJ t ... 7.324 ksi For SA-3S1 (CF3) Stainless Steel S Y
= 17.75 ksi E ... 25.7 % 103 ksi RESID ... 27.5 ksi Therefore, based on stress intensities, SI = 1.409 +
27.176 + 27.5 17.75 17.75 + 51.4
- 0.0794 + 0.7907 = 0.8701 < 1.0 allowed NEBG.a07A (6/80)
8-165 22A74S4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 164 PlEV 1
- BUSINESS GROUP 20.3 Conclusion, All stress indices are less than allowable of 1.0. The calculated indices are as follows.
S. 1. Location (1) (2) c 1.002 0.96S 11 0.7929 I 0.8701 (1) Based on Stress Intensities (2) Based on Positive Principle Stresses
- NEBG-807A (6/80)
8-166 22A745 NUCLEAR ENERGY GENERAL.; ELECTRIC SH NO. 1 BUSINESS OPERATIONS REV 1 I APPENDIX 30 RECALCULATICtiS REQUIRED DUE TO MANUFACTURING DEVIATIONS This attachment analyzes the effects of the manufacturing deviations on the feedwater safe end and thermal sleeve. The deviations are described in detail in DDR Numbers 15139. 26521. and 15127. dated April 29, 1981. July 2. 81. and Hay 7, 1981, respectively. 30.1 Deviations Due to DDR Number 15127. The deviations reported within DDR Number 15127 are two undersize locations on the thermal sleeve pads. The undersize amounts are two mils and four mils on the 7.812 inch diameter and the 7.625 inch diameter, respectively. It is noted. however. that the deviations are on the thermal sleeve pads. and within the previous calculations the pads are neglected conservatively. Therefore. these deviations will have no effect on the previous analysis and will not require recalculations of stresses. 30.2 Deviations Due to DDR Number 26521, 'The deviation reported in DDR Number 15140 is a tool undercut along the tapered surface bet~een the 10.84 inch diameter zone and the 12.00 inch diameter zone. However. this tool undercut was weld repaired. heat treated. and machined to original finish size. This type of repair is allowed per ASME Code, Reference 6.2. Post-weld heat treatment ensures stress relief and an acceptable metallurgical condition. The part is restored to the original specified dimensions. Therefore, these deviations will have no effect on the previous analysis and requires no recalculation of stresses. 30.3 Deviations Due to DDR Number 15139. This paragraph analyzes the effects of two manufacturing deviations on the Monticello feedwater safe end. The deviations are described in detail in DDR Number 15139. dated April 30. 1981. and illustrated on Figure 30.3-1. 30.3.1 Summary of Results. The results obtained for these deviations are compared below with those from the previous nominal calculations. NEO 10711. (REV.tO,It,
8-16": 22A7454 NUCLEAR ENERGY GENE RAL . . ELEe TRIC SH NO. 166
- IUSINESSOPERATIONS REV 1 TABLE 30.3.1-1 PRIMARY STRESS ANALYSIS (All stresses in ksi)
After Deviations Before Deviations All owabl es Case Loea tion Pm Pm+B Pm Pm+B Pm Pm+B Design F 13.96 24.38 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 1S.07 33.28 27. as 41.77 Level C G 4.25 23.37 4.11 22.77 27.85 41.77
- NEO 107 A ( REV. 10/111
Z
/II UJZ Cc £ ClIO o zr mm ~ " CII>> ...... CII:D CD g Glm
- DZ om C:D
-OGl G')
m z: 0.576 MIN - m 0.125R
- a 0.0045 l>
-i r-
,/
//'
/ - O.125R 12.045 4l m
I ., r-rn 1 n
-t
- xl 9.676 IJl ------L 10° MAX n 10.840 IJl 1
~ 8.413 + :0 N m N 9.390 IJl < >
1.50 - -L 8.318 IJl ""-J t-' t;
~
1 _L 1 4.06 ---~0.50 NOTE: DASHED LINES INDICATE REQUIRED BLENDING (I) J: Z
~
FIGURE 30.3.1 DEVIATIONS DESCRIBED IN DDR NO. 15139
8-169 NUCLEAR ENERGY GENERAL. ELECTRIC 22A7454 SH NO. 168
- BUSINESS GROUP IIIEV 1 TABLE 30.3.1-2 PRIMARY PLUS SECONDARY S'mESS ANALYSIS (All stresses in xsi)
After Devia tiODS Before Deviations Location P + Q P + Q. P + Q P + Q. Allowable F 62.3 37.23 61.68 36.61 5S.8 G 62.82 45.94 62.29 45.42 55.8
- Thermal Bending Removed TABLE 30.3.1-3 FATIGUE ANALYSIS After Devia tion Before .Deviation Location Fatigue Usage Fa Ugue Usage F 0.439 0.279 G 0.41 0.409
- NEBG-807A (6/10)
22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC BUSINESS GROUP REV 1 30.3.2 Primary Stress Ana hs is
- a. Section F (As Built) t = 0.576 in - (1/32 in + 1/16 in) = 0.4822 in
\. ~ I corrosion Do = 10.831 - 2(1/32) = 10.7685 in Di = 9.679 + 2(1/16) = 9.804 in 2
Area = 15.584 in I Z =C = 38.365 0.0045 0.576 1L .) 1 9.679 q, J-FIGURE 30.3.2._ SECTION F (AS BUILT) NEBG-807A (6/80)
8-171 NUCLEAR ENERGY BUSINESS OPERATIONS GENERAL 0; ELECTRIC' 22A7454 REV 1 SH NO. 170 30.3.2.. (Continued) (1) Design Conditions Design Pressure Stress 1250(9.804) as = 2(0.4822) = 12,708 psi 6354 psi a = -1250 psi r Stress Due to Nozzle Loads P=4.05kip M = 534.4 in-kip F z
= 2~28 kip M = 534.4 + 4.05(1.62) = 541 in-kip M
a BEND = Z = 38.365 = 14.102 ksi 541
~ .
= -X = 15.584 Ft 2.28 _ 0 147 a AX -. ~Sl Stress Due to Thermal Sleeve Loads P = 2.57 kip M = 3.124 in-kip F = 5.7 kip z
M = 3.124 + 2.54(26.97) + 5.7(1.8) = 82.7 in-kip M
= Z = 38.365 = 2.156 82.7 a
BEND ksi
= F z* _ 5.7 a
AX A - 15.584 = 0.366 ksi NEO .07A IREV.IO/")
8-172 22A7454 171 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. BUSINESS GROUP REV 1 30.3.2.&.(1) (Continued) Total Stress C1q, = 6,354 + 14,102 + 147 + 2,156 + 366 = 23,125 psi C1 = 12,708 psi 9 C1 r
= -1250 psi
( 2) Service Level C Conditions Service Level C Pressure Stress 1375~9.8042 C1 = = 13,978 psi e 2(0.4822) C1 C1q, =~ 2
= 6989 psi C1 r
= -1375 psi \
Stress Due to Nozzle Loads P = 6.44 kip H = 789.3 ill-kip F = 4.61 kip z - H = 789.3 + 6.44(1.62) = 799.8 in-kip 799.8 = 20.847 ksi C1 BEND
= 38.365 4.61 C1JJ.. = 15.584 = 0.296 ksi NEBG-807A (6/10)
8-173 22A7454 NUCLEAR ENERGY GENERAL., ELECTRIC SH NO. 172
* . BUSINESS GROUP ~EV 1 30.3.2.a.(2) (Continued)
Stress Due to Thermal Sleeve Loads P = 5.08 kip
){ = 5.39 in-kip F
z
= 6.0 kip H = 5.39 + 5.08(26.97) + 6.0(1.8) = 153.2 in-kip 153.2 C1 BEND = 38.365 = 3.994 lsi C1 AX = 1~:~84 = 0.385 ksi Total Stress C1~ = 6989 + 20,847 + 296 + 3994 + 385 = 32,511 psi C1 e = 13*,978 psi C1 r
= -1375 psi (3) Thickness Requirement for 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 (Smp /Smn ).
1I'here: t p
= Pipe Nominal Thickness S
mp
= Pipe Allowable (S )
m S = Safe End Allowable (S ) mn m
- NEBG-807A (6/10)
NUCLEAR ENERGY -- GENE RA L ' " ELEe TRIC 22A7454 SH NO. 173 BUSINESS GROUP REV 1 30.3.2 *** (3) (Continued) For our geometry; tp S mp
= ~.5405
= 18.1 in ksi J t p
(S I S ) mp DID
= 0.526 in Smn = 18.6 ksi Assuming Section F is the safe end thickne*ss, Safe end thickness = 0.576 in > 0.5405 in
.. CRITERIA MET NEBG-807A (6/80)
8-175 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 174 BUSINESS GROUP PIIEV 1 30.3.2 (Continued)
- b. Section G (As Built) t =
(9,39 - 8.378) 2 - (2
- 1/16) .. 0.381 in corrosion t
assumed
. = (9.39 -
2 8,413)
- (2
- 1/16) = 0.3635 in D
o
= 9.390 2(1/16) = 9.265 in Di = 8.413 + 2(1/16) = 8.538 in
- Area ~ 10.166 in z = ~ = 21.77 in 2
3
/ /
f I 8.4134> I 1 8.378 4> k-4.06 1 I
.50 9.390 4>
FIGURE 30.3.2-2 SECTION G (AS BUILT) 1 NEBG-a07A 16/80)
8-176 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 175 BUSINESS GROUP REV 1 30.3.2.b (Continued) (1) Design Conditions Design Pressure Stress a 222(8.538)
= 2(0.3635) = 2 608 psi 9
- a a
q,
=~2
= 1304 psi a = -222 psi r
Stress Due to Thermal Sleeve Loads P = 2.57 kip M = 3.124 in-kip Fz = 5.7 kip M = 3.124 + 2.57(23.87) = 64.47 in-kip
- ~ 64.47 = 2 962 ~S1*
vBEND 21.77 * ~ a AI = 1~:~66 = 0.561 ksi Stress Due to Nozzle Loads P = 4.05 kip M = 534.4 in-kip F =2.28 kip z NEBGoa07A (6/80)
8-177 22A7454 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. 176 INESS GROUP "EV 1 30.3.2.b(1) (Continued)
~~ same ratio used in the previous nominal calculations (Page 35) will be used here. This ratio limply accounts for the stiff nozzle influence.
M = 534.4 + 4.05(4.72) + 2.28(1.8) = 557.62 in-kip a BEND = (0.381)~ = (0.381)Sii:~; = 9.759 ksi F a AX = (0.381) ~ = (0.381)1~:i:6 = 0.086 ksi Total Stress a cp
= 1304 + 2962 + 561 + 9759 + 86 = 14.672 psi a
e= 2608 psi a r = -222 psi (2) Service Level 'C' Conditions Service Level 'c' Pressure Stress e = 2(0.3635) = 333~81538~ a 3911 psi as a cp
= -2 = 1956 psi a
r
= -333 psi
- NEBG-807A (6/80)
8-178 NUCLEAR ENERGY ---. GENE RALfj ELE CT RIC SH NO. IUSINESS OPERATIONS REV 1 30.3.2.b.(2) (Continued) Stress due to Thermal Sleeve P = 5.08 kip M = 5.39 in-kip F z
= 6.0 kip H = 5.39 + 5.08(22.87) = 126.65 in-kip CJ BEND
=!I Z = 126 1 65 21.77 = 5.818 ksi F
CJ AX = A% = 1~:~66 = 0.591 ksi Stress Due to Thermal Sleeve P = 6.44 kip K = 789.3 in-kip Fa 4.61 kip z The same ratio used in the nominal calculations (Page 35) will be used here. This ratio simply accounts for the stiff nozzle influence on the thermal sleeve. H = 789.3 + 6.44(4.72) + 4.61(1.8) = 828.0 in-kip CJ BEND = 0.381 Z M
= (3.81) 828 21.77 = 14.491 ksi CJ AX
= 0.381 ~= (.381) 1~:~~6 = 0.173 ksi Total stress
~ a 1956 + 5818 + 591 + 14,491 + 173 = 23,029 psi CJ .. 3911 psi CJ r
* -333 psi NEO .01A (REV. 10/'1)
8; ELECTRIC 8-179 22A74S4 178 NUCLEAR ENERGY GENERAL SH NO *
- BUSINESS OPERATIONS REV _.J._.,
30.3.2 (Con tinued)
- c. Summary Condition Section Pm_
__ P P + P P + P Allow m Allow m B m B Design F 13.96 18.6 24.38 27.90 Event G 2.83 18.6 14.90 27.90 Service F 15.36 27.85 33.89 41.77 Level ' C' Event G 4.25 27.85 23.37 41. 77 (All stresses in ksi) All condi tions are met for primary stress analysis.
- 30.3.3 Primary Plus Secondary Stress Analysis. This portion of the report discusses the detailed stress evaluation of the thermal stress, pressure stress, mechanical stress, stress ranges. and fatigue usage for selected locations of the geometry.
30.3.3.1 Thermal Stress. Sections F and G are thinner than their corresponding sections in the previous nominal calculations. The reduction in thickness for these sections is 1.65 percent and 2.3 percent (Section F and G. respectively). This small change will have no significant effect on the heat transfer coefficients and thus "ill be' directlt propo~tional to the change in the biot's number. For Section F, the biot's numbers before-and after the reduction are 1.067 and 1.0494, respectively. For Section G, the biot's numbers before and after are 1.1785 and 1.1514, respectively. For both sections, the biot number is in suc~ a range that the small percentage change in it due to the thickness change will have no significant impact on the stresses in the region *. This, along with the f.act that, as the thicknesses are reduced, the relative stiffnesses are lowered, causing thermal stresses.
"hich are secondary or displacement controlled stresses, to be lowered.
Therefore. the stresses obtained in the previous nominal calculations "ill be conservatively used here. 30.3.3.2 Mechanical Load Stress, The stresses obtained by mechanical loading are rec~lculated to accollDt for the reduced thicknesses *
- NEO 107 A (REV. 10/11)
8-180 NUCLEAR ENERGY --~~ GENERAL. ELECTRIC 22A7454 SH NO. 179 BUSINESS GROUP REV 1 Section F Nozzle Loads (P + Q) P = 4.42 kip P = +/- 3.136 kip
)f = 275.4 in-kip* H = +/- 390.9 in-kip F
z
= 1.37 kip F z
= +/- 0.26 kip M = 275.4 + 4.42 (1.62) H = 390.9 + 3.136 (1.62)
= 282.56 in-kip = 395.98 in-kip
=
282,56
= 7.365 ksi = 395,98 = 10.322 hi 38.365 38.365 a = 1.37 = 0.088 ksi AX 15.584 Thermal Sleeve Loads p = +/- 2.52 kiP.
F z = 3.7 kip M = +/- 2.333 in-kip F z
= +/- 1.5 kip x = 3.7 (1.8) = 6.66 in-kip M = 2.333 + 2.52 (26.97) + 1.S (1.8)
= 73.0 in-kip 6.66 73 ,0
~BEND = 33.365 = 0.174 ksi a = = 1.903 ksi BEND 38.365 1.5 a AX = 15~S~4 = 0.238 k5i . a AX = 15.584 = 0.097 ksi Total Stress a.BEND = 7.539 psi = +/- 12.225 psi
- a. =
AX 326 psi = +/- 114 psi NEBG.a07 A (6/10)
8-181 22A7454 180 NUCLEAR ENERGY GEN ERAL., ELECTRIC SH NO. REV 1
- BUSINESS OPERATIONS Section G (P + Q)
Noz.:z:le Loads P = 4.42 kip P = 1: 3.136 kip H = 275.4 in-kip M ". 1: 3 90.9 in-kip F = 1.37 kip F z
= 1: 0.26 kip
- z:
M = 275.4 + 4.42 (4.72) + 1.37 (1.8) M = 390.9 + 3.136 (4.72) + 0.26 (1.8)
= 298.73 in-kip = 406.17 in-kip
= (0.381) 298.73
= 5.229 ksi = (0.381) 406.17 = 7.109 ksi at 21.77 21.77 BEND 1.37
= 0.052 a, = ( O. 3 81 )
0.2
= 0.010 OAX =
a (0.381) ksi lu.166 ksi 10.166 AX Thermal Sleeve Loads p = 1: 2.52 kip F = 3.7 kip M = :t 2.333 in-kip z F z
= 1: 1.5 kip M = 2.333 + 2.52 (23.87) = 62.49 in-kip
_ 3.7 = 0.364 ksi = 62.49 = 2.871 ksi all - 10.166 21.77 all = 10~i~6 = 0.148 ksi Total Stress a = 5,229 psi = 1: 9,980 psi BEND all = 416 psi = 1: 158 psi
- NEO 107" (REV. 10/",
4 NUCLEAR ENERGY GENERAL. ELECTRIC SH NO. BUSINESS OPERATIONS REV 1 30.3.3.3 Pressure Stress, At Section F and G, the thicknesses are 0.576 inch and 0.4885 inch compared to thicknesses of 0.5855 inch and 0.50 inch, respectively, in the previous nominal calculations. This thickness disparity is expected to affect only the primary stress component, leaving the secondary and peak stresses the same. Therefore; the stresses in these sections are equal to the stresses obtained in the previous nominal calculations, plus a correction stress. This correction stress is calculated as follows: Section F Pressure = 1111 psi Thickness = 0.576 in; 0.5855 in Diameter = 9.679 in; 9.669 in Axis 1 ere m
= 2t =~
4t Aer Hoop
= 161 psi NEO 107A 'REV. 10/11)
e, 8-183 22A74S4 182 NUCLEAR ENERGY GENERAL ELECTRIC SH NO. BUSINESS OPERATIONS REV 1 30.3.3.3 (Continued) Section G Pressure = 111 psi Thickness = 0.4885 in; O.SOO in Diameter = 8.413 in; 8.38 in Axial PD O"s = 2t
~O" Hoop = ~CJe = 26 psi
- 30.3.3.4 Total Primary Plus Secondary Stress Ranges. Calculations of P + Q stress intensity ranges at Sections F and G yill be performed. to validate the subsequent fatigue analysis. These stresses are calculated in the same manner as those in Paragraph 4.2.5.
30.3.3.4.1 Thermal Stress Ranges. The thermal stress ranges are identical to those found in the previous nominal calculations (refer to Page 110). P + Q O:si) P + Q (ksi) Section (Membrane Plus Bending) (Membrane Only) F 28.78 *3.71 G 34.86 17.98
- NEO 107 A (REV. 10/11)
8-184 NUCLEAR ENERGY --- GENE RA L
- ELEe TRIC 22A7454 SH NO. 183 BUSINESS GROUP REV 1 30.3.3.4.2 Mechanical Load Stress Range P + Q (l:s i) P ;t- Q <<bi)
Section (Se ismic Inc 1) (Seismic Not Incl) F 24.68 7.87 G 20.28 5.65 30.3.3.4.3 Pressure Load Stress Range, These stress ranges are exactly the same as those found in the previous nominal calculations (refer to Page 111), plus the added stress calculated earlier. Sec*tion P + Q 0:5 i) F 8.84 G 7.68 30.3.3.4.4 Total P + Q Range. The total P + Q stress range is as follows: P + Q Range (ksi) Section P + Q Range (ksi) No Thermal Bending F 62.3 37.23 g 62.82 45.94 The allowable range is 3 Sm = 55.8 ksi. Both locations are acceptable with thermal bending removed. NESa..07A (6/10)
8-185 22A74S4 184 NUCLEAR ENERGY GEN ERAL. ELECTRIC SH NO. BUSINESS OPERATIONS REV 1 30.3.4 Fatigue Analysis, This section provides all the detailed fatigue analysis required to shoy an acceptable design. 30.3.4.1 Stress Concentration Factors Section F The' geometry of Section F is illustrated in Figure 30.3.4.1-1. To calculate the stress concentration factor, Reference 6.6 will be used. The concentration factors 'used prior, (refer to Page 124), are (inner) 1.59, and (Outer) 1.48. The deviaton affects the outer factor only. Assume: r = 0, hence Xo = 4.0 Then, using Paragraph 4.7.2.4 of Reference 6.6, (X' - 1) ex: o - 1)
=1 - ..Ji 90 Solving for X', usin~ ~ = {90 - 29)0 = 61 0 X' outer = 1.97 Section G The geometry of Section G is illustrated in Figure 30.3.4.1-2. To calculate the stress concentration factor, Reference 6.6 will be used. The concentration factors used prior (refer to Page 126) are (inner) 1.0, (outer) 1.80. The deviation affects the inner factor only.
Assume: r = 0, hence, Xo = 4.0 Then using Paragraph 4.7.2.4 of Reference 6.6, (X' - 1) (X - 1) =1 - ..Ji 90 o Solving for X', using ~ = (90 - 10)0 = 80 0 X' inner = 1.333 Since X' outer is larger than X' inner,'X' outer will be used in the fatigue analysis. X' = 1.80 NEO 107A (REV. 10/111
8-186 4 NUCLEAR ENERGY GENE RA L . . ELEe TRIC SH NO. IUSINESS OPERATIONS REV 1
~
3.214 0.585 --"l I I
\ /
t ....... -.-/ 1.81
""1 Ll I
I I I I I I I J I 9.669 cp 2.56 I 10.84 cp 1 12.00 -; 1- 1 8.38 4> e 1 0.015 1_ 0.0045 r - , 8.7oj~- _ __ 10.84 $ DETAIL A FIGURE 30.3.4 .1-1 SBcrI~ F DEVIAl'I~ AFTER BLENDING e,' NIEO 107 A ( REV. 10/111
8-187 22A74S4 186 NUCLEAR ENERGY . GENE RA L
- ELEe TRIC REV 1 SH NO *
- BUSINESS OPERATIONS I
- - - .25R ,I
- 50 12.00
- 1-1 10.75 <P 8.413 ~
1-i 8.378¢ 1
....l-
- 8. 413 I 8.378 ~
'1
~
DETAIL B 1
- FIGURE 30.3.4.1-2 SECTION G DEVIATION AFTER BLENDING NEO .07A (REV. 101.1)
e'
~-188 22A74S4 187 NUCLEAR ENERGY GENERAL ELECTRIC REV 1 SH NO.
BUSINESS OPERATIONS 30.3.4.2 Alternating Stress Range. To calculate ~he altenating stress range, the folloYing equation is needed: Yhere: SN xe = 1 + (1 - n) n(Jrl)
--1 3S m
For carbon steel n = 0.2; m = 3.0 X = stress concentration factor t SN = P + Q stress intensity range F1 = peak stress For Sections F and G, the peak stresses calculated in the previous nominal calculations are assumed to be identical. The deviations are not severe enough to significantly vary these stresses. Mechanical Mechanical Range Range Thuma! Pressure ~ ~ Location (Seismic) (No Seismic) Range Range (Seismic) (No Seismic) F 24.68 7.87 28.78 8.84 62.30 45.49 G 20.28 5.65 34.86 7.68 62.82 48.19 (All stress in ksi) NEO '07A (REV. 10'"1
8-189 22A74S4 188 NUCLEAR ENERGY GENERALe ELECTRIC SH NO.
- BUSINESS GROUP "EV 1 FINAL 30.4.3.2 (Continued)
I Salt Salt
~ ~ e Location ~Seismic} ~No Seismic}
I
-.:...L i {Seismicl ~Seismicl ~No Seismicl F 62.30 45.49 1.97 11.93 1.233 83 .0 50.78 G 62.82 48.19 1.80 6.45 1.252 74.8 46.6 Using the total alternating stress range, the fatigue ulage factors C&1l be solved. From the design specification (Reference 6.1), there are 1,500 thermal cycles and it is assumed there are 10 seismic cycles.
(No~_e: SA-3So-U'2 has UTS < 80,000 psi.) (Seismic) (ksi) Ebs ti c Modul us (ks i) Allowable Usage Location Salt Factor SA Cycles Factor F 83 .0 1.103 91.55 750 0.0133 G 74.8 1.103 82.50 1000 0.010 (No Seismic} (ksi) (ksi) Allowable Usage Elastic Modulus Location Salt Factor SA Cycles Factor F 50.78 1.103 56.01 3500 0.4257 G 46.6 1.103 51.40 3750 0.3973 Location Total Usage Factor F 0.439 G 0.407
~or these sections (F and G), the high cycle and existing fatigue usage are negligible. Therefore, the above is the total fatigue usage.
NEBG0807A (6/80)
- MONTICELLO ----. 9-1 EXHIBIT 9 REACTOR VESSEL RAPID CYCLING (STRESS R~PORT)
- REV 5 12/86
; t1~ LUI:.NI: KY \KAPLU l.Tl.}
9-2 GEN ERt~ l@,j ELECin IC REVI:IC~ STATUS SHEET 22A7227 2 1 MNTS. ) CONlON !. ... EEl S ... No. NUCl.~AJ'! Ef04ERGY DIVISION 80-0 .rtL DOCU!&BST TITlE R~o...L.IG~ v'E.5Se:L ( R/,P 10 CYCU:1G) DS1ECIFlcAno~ OOR:',;\1r:1G 0 OTHER
----------------- ~PE ___S_i~RE~S~'S~R~E~P~OR~,T~_________
F~F MONTICELLO LEGE130 O~ OE:CRiFTIO:J OF GROU;>S r£l'l rh. til A o DMH-1131 889R PS c.,. ,1-," GA BAYLIS * ~ G 1980
9-3 2211.7227 SH NO.2 NUCLEAR ENERGY BUSINESS GROUP REV 0 Certification of ~tress Report This certification for the Reactor Vessel (Rapid Cycling) Stress Report, ~nd accoopanying documents, constitute the basis for the Stress Report required by Pilragraph NCA-3SS0 of the ASHE Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Cooponents, 1977 Edition, with &ddenda to and including Suomer 1978. I hereby certify that this stress report was prepared under my direct snpervision and that I am a duly registered Prcfessiontl Engineer under the laws of the state of ~Un.nesou. I certify thn!.:, to the best of my knowledge and belief. the Stress Report for the Reactor .Vessel (Rapid Cycling) is correct Ilnd co~plete and in coopliance with the requirements of Article NB-3000 of the AS!*:F; Boiler and Pressnre Vessel Code, Section III, Nuclear Power Plant Compone.nts, 1977 Edition, with Addenda to nnd including Swcmer 1978. Listed Docu~ents Type of Doc=ent Revision Docu::lcnt _N~r.lber N1.l::lber Design Specification Reactor Vessel 22A7111 o (Rapil Cycling) Certified P.E. NU:lber /4372. State: Date: HEBG aOlA
9-4 NUCLEAR ENERGY
/!.~ ELI,;r: C-i nI uf't GE-,~\ E~n AL ~~jtl 22A7227 SH No.3 BUSINESS GROUP REV o
I: SCOPE 1.1 This rcport docu:::cnts a rapid cycling fatigue analysis of the feedwater nozzle replacement safe end and therl:lal sleeve asscoblj" for the 'reoove.ble'
°trpc sparger at Monticello. The analysis was perfor::ed in accordance "ith the ASHE Codc, Scction III (Reference 6.3). Thc 1etailed analysis is contained in Ap~endix 10 of this report.
- 2. SumtARY AND CONCLUSIONS 2.1 The calculations presented in this stress report for the feedwater nozzle replacecent safe end assecbly sho" that. "ith no !eakage. the fatigue usag~
factor due to rapid cycling is very low (less than 0.0001 during 40 years).
- 3. DESIGN 3.1 The feed"ater nozzle replacccent safe end llne! t~erc:al sleeve asse:::bly sholl.n in Fi1;urc 1 llre dr-Signed :lnd analyzed in accordance with the c!oc~ents referenced in Paragraphs 6.1 Rnd 6.2.
- 4. ANALYSIS 0 4
- 1 Thcr!':l:1l Strcss C\*cles. There l1re t'!o types of thereal cycles defined in the spe~.i.licntions referenced. in Paragraphs 6.1 and 6.2: (1) ,;ystcm cycles and (2) rapid cycles. Syste1:l cycles I1re thr result of operatic/nal trcnsients such as startup. illitiatioll of feedwater °flo;;, scr~s. etc. being icposed on the nozzle. Rapid temperature cycling results in the unstable turbulent mixing of hot and cold water aroend the feedw~ter nozzle at steady state operating conditions. Since the systel:! cycle transient stresses cause the lIIaxioum thermal stresses that can be pro:'oo,ced. thro rapid cycle stresse~ are not linearly additive to the ca:r.imUl:l systeo cycle stresses, ie, the stresses and fatisue daQase due to systec c~cles and rapid cycling oay be calculated independently of ecch other and the usage factors may be added.
4.2 ~oteo Cycle Flit igue Analyl'..ih The sy~tec c)'cIes defined in Reference 4.2 will be analyzed elsewhere. 4.3 Rapid Cycle Fatir.lle Analysis. The equlltions and procednres for determining rapid cycle fatigue, as define~ in Reference 6.1 were progre~med into a timeshare computer code. The details of this code are presented in Appendix 10. 111 this code. the u~ase factor is calculated for a design life of 40 years. as specified in Reference 6.1. "&8 .. 107A
9-5 NUCLEAR ENERGY 2211.727.7 SH NO.4
** ~SINESS GROU? REV C
- 5. RESULTS 5.1 Based on the date. obtained fro::! h_fe~er.ce. 6.1, the fatigue usase factor caused by rapid cycling in the nOlzle, safe end Lou: then:ul sleeve was calculated, as sl:own in dete.il in Appendix 10 * . ~n*:* :.::;~umption that no feedwater, or ~n insisnific~nt amount, lenks t!~~~~~ ~l.~ interference fit be-h.een thc spar .. er tee Il~d the rcplaceClent the '.:' :; ;'.i :'*~ve (inner tber::lal sleeve, Figure 1) WhS cade. For a design life o. ,~ years, the r~pid cy~ling usage factor will be less than 0.0001 fer all locations.
6* P..EFER£~CES 6.1 General Electric Cocp~ny, Reactor Vessel (edpid CyclinS), Desisn Specification 2~A7111, Rev. O. 6.2 Rea::tor Vessel (Syste::l Cycling), Desi!;n Specification 22A6996, Rev. O. 6.3 AS~ffi Doiler and Presscrc Vessel Codr. Section III, Nuclear Power Plact Coeponents, 1977 Editi0n with Addenda throuch 1978. NEaG .OlA
20.50 REIJIflN A FLUW ' CO) rn
~ ;:!
n r-a
~ ::;l l> -~
103.00R ~ ....
! ~
~ ~~"S-\.~
m "_
, .188 STN STL CLAD
- 0 C)
,\,.i",,_
I. ,"_
.~
27 ~ rn
- 3. 1 Q r-2.50~J' < m 121 C')
I ... ~ '1.50 o ...... Z ::J REPLACEMENT CARBON STEEL
~ (")
CARBor~ STL SAFE END FIELD HELD
..)
OUTER STN STL HER~lAL SL~EVE lO.R4 O. 720 ~. OlD £I L REPLACP1ENT STII :II
,., N I-llU 10.75 £I 10.88 ~ ::-:-:.. T11rW-I.,\L N SLEEVE '"-J o N
~~,OO £I I NOZZLE~ '"-J 1 1 1--- ----- -- '"
9.669 :!:..OlD £I
-' .013 INTERFERENCE FIT 1/1
- z:
NOTE: NOZZLE DIMENSIONS ARE REFERENCE DIMENSIONS ;,::
~
VI FIGURE 1 EXISTING FEF:DHATER NOZZLE AND REPLACEHENT SAFE END ASSEHIILY
\0 I
0\
9-7 122A7217 SH No.6 NUCLEAR ENERGY BUSINESS GROUP REV 0 APPEr-'DIX 10 RAPID CYCLIFr; FATIGUE CALCUL\TION 10.1 Introduction 10.1.1 The purpose of !!-;~ .-.nalysis is to determine the r:tpid thereal cyclintj effects on the fee~ws!~~.~*~ e and s:tfe end for the design life of the syste::l. 10.1.2 Rapid temperature cycling (on the order of v.l to 1.0 nz) occurs as a result of cold feedn~tc: aeing injected into-a hot reactor. lbe most dO::linant cause of this cycling in the nozzle bore and on the blend radius is turbulent mixing of leaka;e flow w~th Region A fluid (see Fig~re 10.1). Rapid cycling' is caused in the absence of leakage flow by turbulent Region A fluiu c:tusing the ther::lal boundary layer around the cold ther::lal !leeve to be broxl"n up and swept against the nozzle. Inco~pJete!y mixe~ sparse~ discharse fl"w and Region A fluid that is carried batt to th~ no~zlc al~o causes sooe ra;!d cicling. 10.2 Procedure 10.2.1 The procedure for ~~ter~ininB rapid cycle fatigue is given in Reference 6.1. A co~puter proura~ was develo~ed. b~sed on this procedure. The following includes a de.tailed de'!'cription of the method and a listinc; of the progr~ used to calculate the ~ffect of r~pid cycling. 10.3 Fatisue Evaluation 10.3.1 Stress Calculation. rne following infor::lation established the condition for rapid cycling:
- a. Amplitude and frequcnc)' irc:1 Table lo-J..*
- b. Feedwater flow, temper:tture, and time data from Table 10-3.
For etch 01 the 26 data points in Table 10-3, there arc 11 data points in
'Table 10-1 *
- HEBG I07A
9-8 Z-2A7227 SH NO. i NUCLEAR ENERGY BUSINESS GROUP REV 0 10.3.1 (Continued) The ectal surface temperature range is calculated to dete~ir.e the altern~ting
!tress produced by rapid cycling. Metal ~urface tecperature range is calculated accorcing to the forcula (given in the design specification, Reference 4.1):
AT = A [(C)(T p-p A where
= Metal surfece peak to peak temperature range, OF A amplitude coefficient for a given freque~cy of cycling, fro~
Table 10-1 c coefficient from Table 10-: feedwater teepera~ure from Table 1C-3
= Region A reecto= ~ater temperature fro~ Table 10-3 NCIiCO 107A
9-9 22A7227 SH No.8 NUCLEAR ENERGY BU5Ir':i::;S GROUP REV o TABLE 10-1 Al,:PLIThllE/FPJ:GUENCY DATA FOR f'J.PJD CYCLING (See Fisnre 10.1> Locations a to
- Index AI!Iplitude Frequency
_1_ A Cycles/llr 1 1.00 15 2 0.98 15 3 0.955 15 4 0.91 30 5 0.84 75 6 0.75 120 7 0.65 150 8 0.55 1 SO 9 0.45 450 10 0.35 1200 11 0.20 7S00 TADLE 10-2 COEFFIClr.~'T C FOR NOZZLE SUr.FACE DO\1NSTRIW*! OF TIIER:!AL SLEEVE J~NCTIO~~ (See Figure 10.1 and Not.es 1 :Lnd 2) Location a 0.10 b 0.09 c 0.10 d 0.10 e 0.10 f 0.12 NOTES:
- 1. Interpolate linearly bet~een defined points.
- 2. The coefficients are zcro for locations not specified.
"£8<; 107A
9-10 G~ rJ EAAL ~>:;;
,1~7l ElEe T~H C t2A7227 I
NUCLEAR ENERGY SH NO.9 BUSINESS GROUP
~/
REV 0 TABLE 10-3 FL0\7. TEMP£:p.An:m: A.:\rn TI~!F: DATA FOP. R.-.. PID CYCLU:G Feed . .*ater Feedwater Region A IYours Index Flow Temperature Teoperatl!re Per _J_ % nat c d of OF Ti:::e-'i Year 1 100 375 546 67.87 5945 2 100 360 546 7.37 646 3 82 345 543 .14.75 1292 4 46 300 538 0.98 86 5 36 280 537 0.45 39 6 20 260 540 0.36 31.5 7 6 225 S~O 0.32 28 fI 6 185 540 0.03 7 9 2.5 185 540 0.11 10.5 10 2.5 240 525 0.16 14 11 2.5 280 480 0.52 45.5 12- 2.5 265 450 0.12 10.5 13 2.5 210 420 0.16 14 14 2.5 185 365 0.56 49 IS 2.S 185 470 0.04 3.5 16 2.5 125 450 0.76 66.5" 17 2.5 SO 215 0.32 28 18 2.05 80 170 0.36 31.5 19 0 300 340 0.49' 43 20 1 350 360 0.005 0.4 21 2 190 350 0.013 1.1 22 2 125 340 0.009 0.8 23 2 70 330 O.OOS 0.4 24 2 190 400 0.01~ 1.1 25 3 200 340 0.002 0.2 26 0 70 70 4.17 365.5 HI[&G aOJA
9-11 NUCLEAR ENERGY DIVISION [ 22"", SH. ~jO. 10 REV. 0
--------------------------------------------------------------------~
Z~ oc
~~---t>
o c::c::: \.C
! /
J 9-12
~-------------------~
NUCLEAR ENERGY 2:*;;.7227 SH NO. 11 DUS!~JESS GROUP REV 0
~-.-
Alternating stress is calculated nccording to the formula Ea.iT n-;> 2 (l-,) where
}
E = Yonng's Modu!.'s = 30.0 x 10 6 Froe Reference 6.1
'J .. Poisson's Ratio 0.3 C1 Instantaneous coe':ficient of thermal expansion hiterp::>i.atetl between poin~s given in Yable 1-5.0 of AS~m Code. Section III. Subsection AN where C1 is evaluated at a tecperature of T TA - 0.5 (j.T p-p )
9-13 NUCLE.lI.R ENERGY GEr-J EnA L ~~.. ELEe T Ai C 22A7227 SH NO. 12 REV o
- --~
BUSINESS GROUP 10.3.2 Usare Factor.... The number of cycles and the a!lov.. able nUl:\ber of cycles are calculated to deter~ine the rapid cycling u~age factor. The n=ber (If cycles allowed is dett'n::ined from the AS~:E Code (Fig~re I-9.1) (Reference 6.3) for a given alternating stress, a
- nlt The n=bcr of cycles accUl:\ulated from rapid therDa! cycling is calculated according to the equation Cycles = t. f
- 0
- L where lycl e s cycle~ induced by r~pid cycling r frcquency of cycling from Table 10-1 in cycles/hr t . tice at each feedwater/!low/tecperature co~bioatioc frc~
Table 10-3 in '}. o design life of nozzle and safe end 40 years (froc t~e design specif~cation) L = 8,760 hrs/yr The usage factor is given by the following equation: 26 11 U = L i=l u.. lJ j=1 where U = usage factor due to rcpid cyclin!.
.th
*U .. = usage factor due to 'J. amplitude and frequency (Table 10-1) lJ h
and for t e J.th flow, temperature and time (Table 10-3 ) NEIIG I07A
9-14 NUCLEAR ENERGY 22A7227 SH NO. 13 BUSINESS GROUP REV o 10.3.3 Comnuter Progr:m L A compoter progr~o based on the preceeding procedure was developed. Table 10-4 is a listing of the program. 10.3.4 Output....: The tabulated results of the cO::lputer prograo, Table 10-4, are listed in Table 10-5. NlEaG a07"
TABLE 10-4 00200 MAMFIoo/S 003(10 00400 TH I S PROGRAM I t, USED TO CAtCIll.A TE RAP 10 eYCL I NO ON THE FEr:QI.*!A TER (FW) OO~OO NOZZLE AND SAFE END BY DErrlll*IINING THe TErlP[I~.'\T"r;or ORf,OIFNTS ([lELlA Tl 00(;00 AT 6 POINTS Ol~ TH~ S!\FE [NO. THE OEl.TA T's AI~[ CIII.CI'LIITFD AS A 00700 FUNCT I ON OF FW TEl-IP, n, FLO'J, FnrOllENCY or Hir 1:.n.F' I n GI'Cl.1 NO DIIE' TO ooooe TUROULEIH FLO'~, I::Te, AND TnlE IIr TEIW. MIO Fl. OF TilE SI\rE END. oo!:!OO THE DEL TA T' S A':E THEN REl ATEO TO CGIIPI~f'S!; I vr ~ rr"-:~s 1:,1 r..r:;It:n TO 0100:: DETER~IINE nlE N,IMBeR OF CYCLES 1\LLO,.,.I\I'L[ fl'ItJt'l TIIC .4~:1I:::: roo':"" CODE. 01100 0120 REAL RATFWF(26), lEMPFW(2G), FWTIMEI2G),' AM?(II) FRrQlll)
,1130 REAL ALPHAlll)., ALPHATlll), TH1I'I(;,2G), T[iIJ'PC(~~G)
(0140 REAL DELTATlG,2G), STHC!:,SIG,~(lI, ,\l.PHltITll;I, r.3(I;,:>G) ('1500 ':>IGOC RATED FW FLOW (RATFWF), !='\,' TEMPE nATURE (Tn1l'Fln, AfIl'LI TUDE I)F HIE 01700 RAPID f:YCLING (AMP.) AIIO FREOU[lICY OF THE n.\r'IO t;'(CI 1 JIG I "H::'O) Ar:E 0160e GIVEN ElY "OESIGN SI'ECIFICATIOI~ FOR FEEtM,'\1r.'l 'lOY/I.E MID SP.FE EtlO 0190e REPAIR", THER~1AL EXPANSION IA!.PH/I) , ~IOG"1 II., ('I' n.I\'~IICI-;-Y 02000 I EtIOD), A~1I'l Fa I SSOIl' S RI\ 1 10 (1'0 I ~')ON) /IRE r'l VC:N IW P.~.II:: r~JPV 021CO SECTION )1, TABLES 1-5.0 liND I-G.O, 0220C 0230 DATA DLIF[V40,OI 0240 DATA RATFWF/I .000,1,000, .~2, ,40, .30, .20, .OG, .on, .O~~, .025, .025, 0250& .C25, .025, .O?3, .025, .025, .O~~, .O::~.,n.o . . 01, .02, 0260& .02, .02, .02, .03,0.01 0270 DATA TEMPFW/375. ,3GO.,3*15. ,300. ,200. ,~(;O. ,2~'., 1(;5., lOS. ,2~0., 02e08. 2(10, ,2(,5. ,210. ,105. ,105, ,1::!5. ,00. ,110. ,:)('10. ,3~0., 0.:?90& I !'to. , 125. ,70. , I !'Ie. ,200. ,70. I 0300 DATA F~TIME/.G767, .0737,.1475, .OC06, .on45, .OO~G, .OO~2, .Dono, "310& ,0012, .OCI~, .0052, .0012, ,001(;, .no~G, .0004, .007n, .0032, 03208. ,003G, .0049, ,0000:5, .00013, .0ono~,on:Jn~, .00n13, .00002, 0330.'1. .0*1171 0340 DATA AMr/l,O,D,96,O.955,O.9I,O,~4,O.7~,O.G~,O.55,O.~5,n.30,U.?OI :t (T' N 0350 DATA FRE0/15., 15,,15. ,30. ,75. , 120. , I SO. , Inn. , ..I:;'). ,1 ;!Ofl. ,7:,,(10,1 ~ N
)::>
03(;0 DATA TEMPDC/54G, ,~4G. ,543. ,130. ,537. ,540. ,~10 ,5.10. ,~40. ,~25., ..., 0:;708. 460. ,450. ,420. ,365. ,4 ';0. , .150. ,?1:;. , 170 .. 3011'1. , 3eo, , o N N 03C05. ~"O. ,340. ,330. ,400. ,340. ,70.1 ...... 03~10 DAiA PDISSON/~.31 0400 DATA ALPHA.'G.07,G.20,G.44,6.r.7,G.aQ.7. 10,7.3:J,7.5~,7.7C,7.~S,O. 16/ U410 DATA ALFHAT /70, , 100, , 150. ,200 .. 2~In. ,31';0. , 35f). ,4C'1). ,*110 ,500 .* 5::;0. I 0420C 0430C CALCUl.A TE THE NUMIlER OF CYCLES FOR EACH [1;::L TAT ,'IS ,'I Fut'e T I 011 04400 OF RATFWF AND TfMPFW, 0.1::;0 PRI NT: " " OdGO PRINT: "MONTICELLO nAPID CYCl.E FATIGUE ANIlI.YSIS" 0470 PR! tH ~ .. VAL1JE!; n:OM NEW [:::1:1, PoT FLml r1*~.p " :x'" 0460 PRI'H: " DESIGN LIFE = ",OLIFf 2 0490 PflINT:" .. ~ 0:500 PRINT:" .. OS10e
"'" \0
.....I
\.II
0520C . CAU':ULATE FOR EACH OF G PO I MTS ON THE THERI1"\L r,I.Fr:E TH( OS30C APPROPRIA1E DELTA T WIIEH( K.I FOR POINT A; I';~:? f'IJ:: f'OI'IT O. OS*10C K=3 fOR POINT C; K=4 FOR POIi'll 0; K=:I FOfl ,.*I)INT 1':: Mli) F"0 FOR 05:>OC POINT F. O*St..lC 0570 00 1000 K~I.P 0500 SUSGF'A C= 0.0 0590 00 500 J=I,2G 0000 C3(1,J) " 0.10 0010 (;312.J) o*O.O() 0020 e313,J) * 'l.10 0030 C314, J)
- 0.10 0040 C3:~,J) .0.10 OG~O C3CO.J) = 0.12 OCC.O 2U~OFA CD 0.0 OG70C OC80 [10 I 00 I
- I; 1 1 0600 OELTATCK,J) = AI'1PIII ~ C3CK,J). (TEMPOCIJ; - TEI1P"'wIJII 0700 TEMPCK,J) = TEMPOC(J) - 0.5
- DELTATIK,J) 0710C 0720C L I Nt=.:ARL v I NTERPOLA TE TO DETEP.~II NE ALPIiA ".5 A FIINr.TI ON OF T EMf'E'~A TUf~E 0730C \,'IiEfiE TEtlPERA TURE ISO I VEil FROl1 HIE SPE C/\S
'0740C T= TDC-~.5aOELTA T 0750C 0770 00 10 L= I, II 0780 IF ITEMPIK,J) .LT. ALNIATILI .OR. TEMP(K,J) .GT. ALPIlATCL+I)) '0190& GO TO 10 0600 SLOPE = CALPI-'.'TCL+11 - AI.PHAHLl )/I"LPHAIL+I) - ALPHAlU) 10010 ALPHINTCIU : ALPHAIL) + CTEi1PIK,J) - P.Lo~IlATIL))/!)I.Clpr OCJO 10 CONTINUE OC50C 00;:;0 CYCLES c F~"IMEIJI
- FREQIII
- 0700 .
- OLIFE 0070 STI?ESSCK,J' = OELTATlK.J)
- Al.PhINTCK) <1O.0/C2.0 a ll.0-POIS50NII 0000. IF CSTRESS'C,J) .OT. 5.8E:I) CYAl.I.0W = CSTHE~SIK,JI/I .4ECI"(-~.23G~1 0000 IF(STBE3S0"JI .LE. 5.f.lEj .AND. STHESSCK,JI .GT. 2.0E51 0900& CYALLOW = ISYRESSIK,J)/1C82000. 1"C-2. IG~~41 0910 IFISTRESS(K,J) .LE. 2.0[5 .AND. STI?[S~IK,J) .GT. 0.4(4)
- r 0920& CYALLO\ol = CSTP.E3~IK,J)/II:l3787. )"(-2.G~'12{\1 IT' N 0930 IF(STRf.$:>C1<,JI .LE. 6.4(*1 .AND. STRESSIK,JI .OT. 3.7:> .. 4) N
- ):>
0040e. Cy,o.LLOW" ISTI':[S5(1<,J) 194"112.)"*1*2.C:;~H21 '-J IFISHlE55IK,J) .LE. 3.7::'E4 .MID. STR[SSCI<,JI .111. 2.0[41 o N 09:50 N 09C04 CYALLOW = IS¥R[~S(K,J)/4G34AG. )*'(-3.G6~~~) '-J 0970 IFISTRE!)SIK,J) .LE. 2.0E4 .Arm. STHESSCI~,JI .(>T. *1.~4F.4) 0!l004 CYALLOW" ISTRESSCK,J)/210309.1*'(-4.0167(;1 0990 IF(STRESSIK,J) .I.E. 1.24~4 .Arm. SlHE:';SIK,J) .GT. 9.4(3) 10004 CYALLOW ~ C::;TRE~SCK,J)/L5140.8)'.(*O.31~~70) 1010 IFCSTRESSIK,J) .LE. 9.4F.3 .MIO. STIlCSSCI<,JI .OT. 7.7(3) 10204 CYAl.LOIJ = I STRESSII<,.J )/37932.91" I -II. r.4?,101 '1030 IFISTP[SSIK.J) .LE. 7.7E3 .AND. SlRr:~S!K,J) .UT. 7.0(3) 1O*~0~ CYAl.LOh = (STR[~~CK.J)/IG50~.03)"(-24 I~nn~) iO!)O IFISTflE.";SCI(, J) .LE. 7.0[3) GO TO 100 ~ 106') USOFA C= CfCLES/CYALlOW
- z
~
l..
\0 I
I-' 0\
~-~~~ 1070 ZUSGFA C= ZUSOFA C+ USOFAC 1000 SUSGFA C= SUSGFA C+ usnrAC 1090 100 CON1'I NLJE 2000 PRINT:" AT Z?NE ",J 2010 PRINT:" STRESS (SIGMA ALTI ~ ",STRESSCK,JI/0.20 2020 PRINT:" SUM OF IJSAGE FACTORS =",ZUSGFI\C 2030 PRINT:" " 2040 ~OO COf"TINUE 2050 PRINT:" AT POINT ",K 2060 PRINT:" SUM OF USAGE FACTORS * ",SU~G~nC 2070 PRINT:" " 2080 1000 CONTINUE
*2090 STOP 3000 END 5000 S EXECUTE 5500 S n!(1JOE.',
;;: N
.< N l:o o N N
;t z
!=
0'1
\0
....I
9-18 GErJ ERA L ~~~ ELEe Tn I C 4:~ . NUCLEAR ENERGY 22A7227 SHNO. 17
- BUSiNESS GROUP REV 0 FU:AL TADLE 10-5 TABULATI::D RESULTS OF cmlPUTER PROGRJ~I*
Usage Factors 7.on e 0'0 i!l t I 2 3 4 5 6 1 0 0 0 0 0 0 2 0 0 0 0 0 0 3 0 0 0 0 0 0 4 0 0 0 0 0 0 5 0 0 0 0 0 0 6 0 0 0 0 0 0 7 0 0 0 0 0 0 8 0 0 0 0 0 .25 E-4 9 0 0 0 0 0 .3 S £-4 10 0 0 0 0 0 0 11 0 0 0 0 0 0 11 0 0 0 0 0 0 13 0 0 0 0 0 0 14 0 0 0 0 0 0 15 0 0 0 0 0 0 16 0 0 0 0 0 0 17 0 0 0 0 0 0 18 0 0 0 0 0 0 19 0 0 0 0 0 0 20 (I 0 0 0 0 0 21 0 0 0 0 0 0 22 Il 0 0 0 0 0 23 0 0 0 0 0 0 24 0 0 0 0 0 0 25 0 0 0 0 0 0 26 0 0 0 0 (I 0 SUJ:I 0 r 0 0 0 0 0 .63 E-4 Usage Fcctor~
-4 Total -= 0.63 x 10 NIEIIG .07A
*}}