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SAXTON - Loss of Coolant Accident Prevention and Protection TABLE OT CONTENTS Section Pan 1.
Introduction.
1-1 2.1 -1 2.
Reactor Coolsnt System 2.1 Description of Design and Fabrication 2.1-1 2.2-1 2.2 Inspection Program.
2.3 Leak and Leak Detection Experience.
2.3-1 3.
Emergency Cooling System Description.
3.1-1 3.1 Lxistir.g Saf ety Injection System Summary..
3.1-1 3.2 Proposed Modifications.
3.2-1 3.3 Electrical Facilities 3.3-1 4
Loss-of-Coolant Analysis 4.1-1 4.1 Blowdown Transient 4.1-1 t
4.2 Core Thermal Transient 4.2-1 i
4.3 Containment Transient 4.3-1
.2 5.
Conclusions 5-1 Appendix A A. O.
Smith Corporation Report on Saxton Tab 1
,) Reactor Vessel (1961)
Appendix B - WCAP-1391 Multi-layer Construction for the Tab 2 Saxton Reactor Vessel (1960)
Appendix C WCAP-1620 Supplementary Technical Informetion Tab 3 on the Saxton Reactor Vessel (1960)
Appendix D Description of Computer Programa Tab 4 t
1 i
i l
1.
INTRODt'CTION AND SL'01ARY i
The contents of this report are in response to the AEC (DRL) December 1966 i
Ictter requesting that the Saxton Reactor loss of coolant accident protec-tion be revaluated in light of advances made in the techniques for analyz-ing core behavior following such an accident.
The design and f abrication standards of the Saxton Reactor Coolant System are reviewed and compared to those currently in use in Section 2.1.
An in-spection program is proposed suppicmenting the existing program to give in-l creased assurance of system integrity (Section 2.2).
In Section 3 the safety injection system design and operati in is reviewed and a post-accident recircula-tion system is proposed.
In Seeidon 4 analysis of core thermal and containment resented utilizing current computational tools for pressure transients are e loss of coolant accident analysis.
It is shown tnat the - emergeny core cooling system capability exceeds its design basis and that there is margin in the I
containment design.
It is concluded that the existing systems supplemented with the proposed inspection program and recirculation system modifications provic'e the necessary protection to the public.
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REACTOR C00LANI SYSTIM 2.1 DESCRIPT10S OF DESIGN AND TABRICATIOS The code requirements for the Saxton Primary System are compared to those of RG6L in Table 2.1-1.
The design and fabrication of the Primary System was in general in accordance with the rules and requirements of both Sections I and VIII of the ASML Code, including the nuclear cases. A re-view of the materials, fabrication and non-destructive testing requirements has been made comparing cur *.nt Section III to Section I requirements for Saxton Primary System.
t Material, f abrication and non-destructive testing requirements were comparable to the present requirements of Section III and the Pressure Vessel in parti-cular was in accordance with rules and requirements of both Sections I and Vill of the ASME Code except that final stress relief was not performed and ultrasonic testing and magnetic particla testing were used in lieu of the Section I and VIII radiography. This is normal procedure for the type of vessel used.
The pittorial representation of previously obtained coolant system quality assurance data is shown in Tigure 2.1-1.
4 I
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'IABLE 2.1-1 REACTOR COOLANT SYSTD1 - CODE REQUIRD1ENTS RG6E Saxton ***
I Component Codes Codes l
Reactor Vessel ASME III* Class A Section I of ASME Code Rod Drive Mechanism Housing ASME III* Clast A Section I of /sSME Code Steam Generators 1
Tube Side ASME III* Class A Section I l
Shell Side ASME III* Class C Section I Reactor Cool'.... Pump Volute ASME III* Class A Section VIII Pressurizer ASME III* Class A Section I Pressurizer Relief Tank ASME III* Class C Section VIII Pressurizer Safety Valves ASME III*
Section I Reactor Coolant Piping USAS B31.l**
USAS B31.l**
System Valves, Pittings and Piping USAS B31.1**
USAS B31.1**
l ASME boiler and Pressure Vessel Code,Section III Nuclear Vessels.
USAS B31.1 Code for Pressure Piping.
Generally the Primary System was designed and fabricated in accordance with rules and requirements of Section I and VIII including nuclear cases, including nuclear rulings.
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4 2.1.1 Reactor Vessel Tabulation of Design Design data for the Saxton Reactor Pressure Vessel is presented in Table I
2.1.1-1.
The design pressure and temperature are seen to be the same as for the RG&E plcnt while the Saxton operating temperature and pressure have been slightly less. The hydrostatic test pressure for Saxton exceeds that for RG6E.
Design Codes The reactor vessel shell is of multi-layered construction while the bottom and top heads, vessel and head flanges and nozzles are of solid wall con:truc-tion. Vessel construction was in accordance with A. O. Smith high pressure multi-layer specification MLS-30A, rules and regulations of the Pennsylvania Department of Labor and Industry receiving a Pennsylvania special number and stamping and was in accordance with the ASME Power Boiler Code Section I where applicable.
Materials, Fabrication and Quality Control l
Fabrication and inspection of the vessel has been in ac;ordance with the rules and requirements of both Sections I and VIII of the ASME Code except that:
1.
Final stress relief was not performed, t
2.
Ultrasonic testing and magnetic particle testing were used in lieu of radiography on welds between the multi-layer shell and the solid wall parts-of the vessel.
A review of the materials, fabrication and non-destructive testing require-ments has been made comparing current ASME Section. III requitements to the
/
construction requirements of the Saxton Rea: tor Vessel. The main differences l
and' exceptions are:
2.1.1-1 s
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1.
Use of ultrasonic, magnaflux, and radiography at partial thickness to inspect circumferential welds to the main shell course and nozzle to main shell course welde, in lieu of radiography.
t 2.
No ultrasonic testing of the main shell course was made occ.ase it is made of multi-layers of 1/2-inch and 1/4-inch thick plates, li 3.
No magnaflux of the velds af ter final hydrostatic testing, and 4.
Welding of the spacer blocks directly to the roll-bonded cladding.
In general, the material, fabrication and non-destructive testing were equivalent to the present requirements of Section III. All A. O. Smith feiracation, heat treatment, and non-destructive testing procedures were reviewed cnd approved by WApD.
The A. O. Smith manufacturers report to Westinghouse on the pressure vessel component chemical and physical tests, list of qualified welders, hydrostatic tests and stress relief results and inspection procedures are presented in Appendix A.
Further Appendix B of this report (WCAp-1391) contains a t echni-cal description, description of manuf acturing procedures, inspection techniques, and weld designs for the Saxton Vessel. Appendix C of this report (WCAP-1620) l l
centains a discussion of ultrasonic inspections and cyclic stresses expected in the vessel and a summary of the vessel inspections.
Layered Yessel Justification, Advantages and past Experience i
some ot' the advantages of multi-layer vessels are:
- 1. -
The relatively thin plates used in multi-layer construction for reactor
. vessels exhibit better metallurgical properties than thick plates.
2.
The precompression built into the inner layers of the multi-layer shell by the combination of wrapping load and weld shrinkage techniques 1
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This precompression was not credited in the calcula-tions for establishing material thicknesses.
3.
By means of weep holes running from the outside of the inner barrei through the outer layer wraps, any fluid which might leak from the inner barrel is monitored.
A more thorough discussion of the advantages can be found in Appendix B.
Also, in Appendix B is a history of non-nuclear multi-layer construction application and a description of some of the nuclear applications oth?.r than Saxton, some of whien are:
kwJ 1.
High Temperature Test Facility (H.T.T.F.) Westinghouse Bettis-Pittsburgh, q
),
8 2.
Special Power Excuision Test (SPERT III) National Reactor Testing Station, Phillips Petroleum Company Idaho.
3.
Core Component Test Vessel (C.C.T.V.) Westinghouse-Lettis, Pittsburgh, Pennsylvania.
4.
Fuel Element Test Autoclave Westinghouse-Bettis, Pittsburgh, Per wylvania.
Pressure Vessel Cyclic Fatigue Analysis The thermal aspects of the Saxton Pressure Vessel start-up were determined by comparison to a similar size vessel which had a detailed thermal analysis.
Using the above results an analysis was made to determine whether the re-quiremente of paragraph N415.1 of Section III of the ASME Code were met.
v-Compliance with the intent of this paragraph exempts a pressure vessel f rom I
requiring a detailed cyclic fatigue anslysis.
2.1.1-3
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The analysis to de:< ninc compliance with N415.1 used the more conservative of fatigue curves, allowable cycles and allowable stresses taken either from the PB 151987 document which was ue.d on th9 original vessel evaluation or Section III of the ASME Code. The vattous criteria of paragraph N415.1 were checked assuming the vessel would be req sired to withstand three times the number of estimated temperature and pressure transient occurrences listed in Pigure 1 of Appendix C (WCAP-1620). The number of temperature and pressure transient occurrences estimated in WCAP-1620 was based on a five year life for the vessel.
This conservative approach shows that the vessel does meet the requirements 8
of paragraph N415.1 of Section III of the ASME Code neglecting the loss-of-coolant accident and assuming a maximum heat-up cool-down rate of 100*F/hr.
Hence a detailed fatigue analysis is not required for this vessel.
Table 2.1.1-2 shows the percentage of the new number of estimated transients assumed for the paragraph 1".15.1 analysis that have been used to date.
Nil Ductility Transition and Post Constructica Testing of the Saxton Reactor Vessel Materials As part of a W PWRPD evaluation of reactor vessel materials, a Saxton Pressure Vessel 1rradiatica m veillance Program was conducted.
This basically consisted af inserting actual reactor vessel materials in capsule inside the reactcr vessel to monitor the effects of radiation. The capsules contained-the inner shell material as well as the material from the wrapped plates.
Results of the evaluation of the capsule removed (No. 577-5) are shown in the Figures 2.1.1-1 through 2.1.1-3 for the Saxton inner shell material (A 212), the Saxton wrapped shell material (modified A 212 AOS 1135) and the ASTM correlation material (A 212).
These materials have a transition temperature shift similar to A302B steel. Figure 2.1.1-4-shows that the
=
observed transition temperature shifcs were less than the design values.
The curves on the figure are plotted for A 302 B steel. The expected maximum integrated fast flux (E > 1 mev) at the-vessel 1R is 1.3-x 10 g
2.1.1-4
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n-cm" at the end of the predicted Core 111 life.
Based on the design curve shown in Figure 2.1.1-4 for 500*F, the predicted maximum shift is 240'F.
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TABLE 2.1.1-1 REACTOR VESSEL DESIGN DATA RG&E
- SAXTON Design / Operating Pressure, psia 2500/2250 2500/2200 liydrostatic Test Pressure, psia 3100 3750 Design Temperature, 'T 650 650 Overall lleight of Vessel and Closure Head, f t-in.
39-1,3 18' Number of Reactor Closure liead Studs 48 36 ID at Shell, in.
132 58 Inl e t Nozzle ID, in.
27.47 12 Outlet Nozzle ID, in.
28,97 14 Clad Thickness, min., inc.
0.156
.125 Lower Head Thickness, min., in, 4.125 4.5 Vessel Belt-Line Thickness, min., in.
6.5 5.0 Closure Head Thickness, in.
5.375 5.25
- Information available from Ginna Station FSAR.
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TABLE 2.1.1-2 SLT1ARY OF CYCLIC STRESSES OF PRIMARY S!?TD1 Actual as a Actuai No.
Projected No. ** Percent of of Cycles of Cycles Projected March 1968 March 1973 s
Cold Start-up (heat-up) 40 150 26%
Cold Shutdown (cooldown) 40 150 26%
Hot Shutdown (from power) 115 750
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Normal Operation 100% power step 150 17%
+ 10% step 10% step 2 loss of flow tests 4 loss of load tests Scrap 11 pH tests 29 350 6.3%
12 unscheduled l
Loss of Flow 2 testa 15 26%
+ 2 stopped pump just after scram Hot leak tests 60 180 35%
Cold Icak tests 63 63 100%
No. of these cycles are insignificant per analysis in accordance with Par.1; 415.1 of Section III of ASME Code.
Num.ber of cycles assumed for Paragraph N.415.1 calculations. These do not represent the design limit but have been chosen as a conserva-tive extrapolation of cycles to be expected including the _ next five years.
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l FIGURE 2.1.1-4 1
2.1.2 Reactor Coolant Pipe and Fitting Table 2.1.2-1 presents the design data.
As in the case of the pressure vessel the pipe was designed for equivalent temperature and pressure to RG6E.
Codes Used and Comparison to Present Codes The Saxton reactor coolant pipe and fittings were designed to USAS B31.1 and also Section-I of the ASME Boiler and Pressure Code. This is equivalent to the design basis which is used for current plants (ASA USAS B31.1) and is directly comparable to the SCE reactor coolant pipe which was designed to USAS B31.1 and also met the requirements of Section 1.
Description of Pipe Size and Fabrication Practices The reactor coolant pipe consists of 12" pipe between the main coolant pump discharge nozzle and reactor vessel, and 14" pipe between steam generator and main coolant pump suction nozzle. The pipe was shop and field fabricated in accordance with USAS B31.1 and Section I requirements..The pipe consists of straight lengths of pipe, fittings, and pipe bends. The centrifugal casting process used for the pipe and static casting process use'd for the fittings have been used in industry for many years. -Prior to accepting bending of centrifugally cast pipe a development program was prepared and carried out to demonstrate that the pipe could be bent without any adverse effects on its integrity.
Ibterials, Fabrication and Quality Control s
The pipe used in the Saxton-Main Coolant-System is centrifugal cast type
-316 stainless steel purchased to the requirements of. ASTM _A351 and Code Case N-9. The static cast fittings are type 316 stainless steel and were-
_ purchased to the requirements are of ASTM A351 and Code Case N-10.
These requirements are identical to the cast fittings being procured for recent plants which are purchased _to ASTM A351,_ Code Case N-10.
The_ pipe and
[
'fittingsweresubjectedto100$_ radiography _perAST}1E-94andmet-there-quirements of AST[1 E-71 Class 2.
9 2.1.2-1
On current pipe additional tensile testc are taken to obtain material properties at 650'T with a goal of having the minimum acceptable yield strength equal to 1.20 X the allowable for wrought material and 1.25 X allowable for castings. Although high temperature tests were not conducted l
on the Saxton material, subsequent high temperature tests (650'F) were conducted on centrifugal cast pipe for the Selni Reactor.
The high temp-erature data (650'F) on the Selni pipe shows a yield strength which is 1.32 x the allowable strength at temperature. The Saxton pipe would have equivalent properties at 650*F.
Records on the Saxton pipe and fittings have been reviewsd and fo"nd to conform to the specifL'ation requirements. This included a review of the chemical analys.
' ensile tests, radiographs, fluid penetrant tests, hydro-static pressure tes _ and flattening tests. The records do not reveal any deficiencies in the specifications of' Quality Control for the pipe and fitting material.
Inspections to Date l
In April 1963 and again in April 1965 inspections were carried out to-determine if any changes had taken place in the Saxton pipe material.
Double wall radiographs and penetrant inspection tests were performed on a pipe bend.
These tests did not reveal any unacceptable indications and did not indicate any cracks-or propagation of defects.
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TABLE 2.1.2-1
,REACIOR COOLANT PIPING DESIGN DATA V
RG&E
- Saxton Reactor Inlet Piping, ID, in.
27-1/2 12 Reactor Oatlet Piping, ID, in.
29 12 Coolant Pump Suction Piping, ID, in.
31 14 Pressurizer Surge Piping, 1n.-
10-(sch-140) 3 (316 Cast S.S.
Design / Operating Pressure, psia 2500/2250 2500/2200 Hydrostatic Test Pressure, (cold) psia 3110 3750 Design Temperature, *F 650 650
- Information available f rom Ginna Station FSAR U-
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2.1.3 Steam Generator Tabulation of Design Data The design date for the Saxton Steam Generator are presented in Table 2.1.3-1.
The design and hydrostatic test pressure are seen to exceed those for RGE while the operating conditions are essentially the same.
Codes Used and Comparison to Present Table 2.1.3-1 presents a comparison of the codes used for the Saxton Steam Generator compared to current codes.
Futerials and Justification Table 2.1.3-1 presents a summary of materials for Saxton Steam Generator and Westinghouse (WAPD) current standard steam generators. Note that there are two major differences. Stainless steel tubes were employed in the Saxton
=-
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unit while Inconel tubing is used in our current units. Stainless steel tubing was used in the Saxton steam generator because it offered the corrosion-resistance that was required. The austenitic stainless steels, however, are known to be susceptible to chloride stress corrosion attack under certain conditions.
This has been minimized by maintaining chemistry within limits listed in Table 2.1.3-2.
Inconel is used in WAPD current steam generators
-~
because it offers the corrosion resistance required while at the same time it offars a very high degree of resistance to chloride attack. The heat transfer characteristics and the' physical properties _-of:Inconel and austenitic stainless steels are similar.
The other major change in steam generator material is the change-from carbon-silicon s' teel to mag-moly steel. This change was made because the mag-moly af fords a reduction in weight as a result of its greater ' strength.
1 Fabrication and Quality Control Table-2.1.3-1 presents a summary-of the non-destructive tests performed on f-both base materials and weldments-for-the Saxton and for current steam _ generators.
I-(-
2.1.3-1 1
L
. -. ~..
.m
-. ~.
- ~ _.
Allowable and Actual Stresses The material properties of the steam generator channel head, tube sheet,-
r shell members and tubing have been reviewed.
The mill specifications for minimum yield and tensile stress exceeded the minimum stress specifications of ASME Section I.
The calculated membrane stresses at design conditions-are consistent with Section I allowable design values.
The stress levels at normal operating conditions are considerably below those at design conditions.
O.
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__.,-.__ _..._,_.. _ ___ u_.-..,
_ _ _ _ - _ _ _ ~... _ _ _. _..... _ _ _ _
TABLE 2.1.3-1 Westinghouse Saxton Steam Current Standard Item Generator Steam Generator Design Code ASME - Section I,1959
\\SME - Section III, 1965 Special Ruling 1270N Special Ruling 1273N Mech. Design Report Channel Head 650*F/2500 psia 650*F/2500 psia Tubes - Internals 650*F/2200 psia 650/1600 (650*F/2500 act. psia:
Tube Sheet 650'F/2500 psia 650'r/1600 psia Shell 650*F/1800 psia 600*F/1100 psia Materials Tubes A-213, TP-304 B-163 Tube Sheet A-266, Gr. II A-302 B Channel Head A-212, Gr. B A-216 WCC Shell Plate A-212, Gr. B A-3C. B Shell Head A-212, Gr. B A-302 B Nozzles A-105 Gr. II A-302 B Partition Plate B-168 B-168 NDT-Spae Material Tubing PT UT + EC (eddy current)
~
Tube Sheet UT + MT UT + MT Channel Head UT RT + MT Shell Plate-None UT Shell Head None UT Nozzles MT UT + MT Hydro Tests - max
. Tubes (mill) 4730 psig (internal) 3135 psig (internal) i Channel Head 3750 psig 3126 psig Tube Sheet 3750 psig 3126 psig Shell 2700 psig 1370 psig
' Gas leak test Yes Yes NDT - Weld Shell, Longitudinal
- RT + MT RT + MT Shell, Circumferential RT + MT RT + MT Nozzle to Shell RT + MT RT + MT Tube Plate Clad PT + UT
-PT + UT-Channel Clad PT + UT PT After Hydro - All Welds
- None' MT
+-
l s
TABLE 2.1.3-2 SU) MARY OF OPERATING CHEMISTRY LEVELS Primary System Hot-500*F Minimum Maximum Normal Boron 3.0 ppm 2366 ppm varies with depletion
.005 ppm
.075 ppm 0.005 ppm Chlorides
.005 ppm
.1 ppm 0.005 ppm oxygen Hydrogen 0.0 105 cc/kg 30 cc/kg Lithium
.005 2.08 ppm 0.1 ppm Cold - Ambient Baron 272 ppm 2744 ppm 1700 ppm Chlorides
<.005
.075 ppm
.03 ppm Oxygen
<.005 Saturation (6-8 ppm)
Hydrogen 0.0 5.0 cc/kg 0
Chemistry of Secondary Side of Steam Generator Minimum Maximum Normal-pH 7.62 (cold) 11.1 10.2 Chloride
. 005 ppm 3.4*
0.2 Phosphates
< ].0 ppm 320 ppm.
50 ppm Hydrogen
.01 ppm 100 ppm (cold) 04 oxygen
.005 ppm
.005 ppm
.005 Total Solids 88 ppm 546 ppm (cold) 100 ppm On one occasion, during startup of the secondary system,-the chloride concentration in the steam generator varied from.06 ppm to 3.4 ppm-and back'to.3 ppm over a 19 hr. period.
- -. _ -..... ~.
. -. -.. ~. -.. - -. - -. - -. -... _..- -... -....
h
- .1.4 Reactor Coolant Pump The pomp design data are listed in Table 2.1.4-1.
A review of records oa the boundaries of the Saxton Reactor Coolant Pump has beca made. The design basis for the Saxton Reactor Coolant Pump is the ASME Loilet and Pressure Vessel Code,Section VIII, 1959, with applicable addenda as of January, 1960. The desigt, pressure is 2500 pei, and the design temperatu.e is 650*F.
The mater;als and non-destructive tests specified for pressure boundary parts cre summariaed in Table 2.1.4-2 where parts are identified by group letters in the tacle according to the following headings:
Croup 3
A Casing B
Stator Assembly C
Stator Cap Assembly D
Bolting E
Sealing Parts i
It will be noted in the table that the drawings call for chemical and physical tests and ncin-destructive tests. Quality Control records are still on hand as of February, 1968, confirming the drawing requirements for the casing,-flange bolts, and cap bolts.-
l-Additional non-destructive tests, not shown inLthe tabic, were given on detail machining, sub-assembly, and assembly drawings to insure sound welds and provide hydrostatic testing. A eview of these drawings shows one_! exception. This is the stator cap assembly which does not specify:n dye penetrant check.for the
~
welds at the vent piping. -Since the cooling tube sleeve at_the elbow-to cooling-coil connection did not have special material requirements, a-dye-penetrant check at'these externally-accessible sleeves furnished more assurance of good-material.
I t
2.1.4-1 Lw
.. _._ _.,, _.m. _.._ _...--.,, _ _.~...., _,. - -, -., -, -..
l i
l l
TABLE 2.1.4-1 REACTOR COOLANT PUMPS DESICS DATA RC&E*
Saxton I
I l
2 1
j Number of Pumps besign Pressure / Operating Pressure, psia 2500/2250 2500/2200 j
Hydrostatic Test Pressure (cold), psia 3110 3750 6, 0 650 5
Design Temperature (casing), 'F RPM at Nameplate Rating 1189 900-2010 252 60 Developed Head, ft Capacity, gpm 90,000 7250 Pump Discharge Nozzle in, 27 1/2 12 Pump Suction Nozzle in.
31 14 Motor Data:
Type Fingle Speed, Class H-Voltage 4000 220-491 60 30-67 Frequency, cps Starting Input (hot reactor coolant), kw 4000 150 Input (cold reactor coolant), kw 5300 190
- Information available from Ginna. Station FSAR.
l-
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E.4 d ? m.4 p*v.
1 PPET T! U:1rAPTE Of *>JT N PIACT3 C09IATP 8'W l
4.A. p.e we s Group Assembly 3. Porte
!vg. No.
f*sertyttoa
- iterist Spec.
T*ste Spe-fled
- per Dws.
C%P
- s. D. Tast Dye.
P* %
6 C"
Fa -d r-s. tw e
n l Notes 1, 7, o1,570 ff'1 N et ng 31$ SST ASTM A 3M1, I
8 I
A Cos:ng T m CF*M; *e I W 3-1 i
l l
2 a
- a l Note B
Stat. Flange 257F'A 701 Port' na yw SST 11' M -1, sal %3. SP.E0 x
f 52D Q1 tD3 l
Bar U
Nates 1, L.
se s
Ther m vel.1 Stat. Shell 2%73 0% 902 Ce nt. Ca st r4 SST 12M3-2, SA351. CM.C74!
l Cooling Colls 504D300 Gol TuW ng 30h SST * %1 1 Nctes b.
- 1 Tortie P n 9%D377 FPS I
Be r 3 4 SST k'J53-3 CL. Tube Siseev 50i D371 174 f
?>d! y 3Oh S3T l -
Mete
- l r
CL. T.:tw Elbev 257F415 W11 l
bb g 304 SST 5?51-1 Note ?
End Ring 257B e* ?e'2 j
forging 30k SST 0937-1, SA375. CR.r*
n j
C Stat. Cap 365Ch11 GT i
P v gtra 3OL S37 '3037-1, SA3%, CR.r" u
f.;bing 25W) 19 'J j
T2ty 3% STT $5!1-1 1
lb ehing 23 B M 703 1
Bar 3Ob SST h 4 3-3 Note s '. '
+
257Po09 ff0k Pipe 30k S3T 11*J!3-1,SA3 7's,TP yA
=
t l
e D
Flange Bolt l
160A % 1 tok fMt Alloy Stl. 6553-;
s 1
cap Bolt 771I683 109 Bol' Alley Etl. 741-7 1
i E
Cap o-Ring 1
53C2613 fr'b "0* R1 ns NJNA-N-121 h o15 Cap O-Ring 53C2613 frT2
- 3* Ring i
3 MA-N-12160-415 Fles. Osaket
$2D2053 FIS Ca stet
, 30h S*T & Astwetes I
y* SJT f Asbestoe 52D2053 Pl?
Gasket s
Fles. Gestet
+
I Job SST 12b93-1,SA3*,1,G?.CF" 1.e v
Ther. Berrier
}
%OhD3C6 101 O sting l
Kover j
160A085 501 I
Kover l
Kover Terminal 160A005 101 Terminal i
Alasdna ceramic Ters. 0-Ring 53C2613 705
- 3* Ring BUNA-3-12160-915 Ters. 0-Ring 53C2613 1125
- 0* ting E*NA-3-12150-915 As celled fer by dreving e.4/or present revision of material Specification 3"JIEFQ*.T?T TPAJ!*C1 gg
'I 1.
100$ Radiography Class 2 Col. of A5"M TT1.
2.
Ry11rostatic Test et ho$O poi for 30 minutes per T.S. 5*0320.
,_g
,,ecret Ase.mdie 71DPS3 ;c1 g
g,,ing g, n,c.,'
7713p2,. gi (This is shown on Rougn Mach. Casirs Drsving 771Dr*JL 701.
,-. W r1 calculatter.s for hydro value on EDst 2TT*49-J ere accertable 3, g 7, e.,
far Code III es well as code VIII).
g St gy7 r, a,. y.
<<4g372 og
?ta. r Aes'y.
% D371 CO2 3
Intresonte enemine.
Shell % Finage
- %D37? G71 k.
Fydrestatic test et la500 poi per T.S. 550%.
71,rg, m,en.
Sty.D3'1 Gil C
Stator Cap Ass'y.
35'Ch91 001 Inspect tubing per Quality Stander 4177TOO1.
Velv. Ass'y.
257PW - El yelse Cat. #13 v.6001 ( A.toelev-Eg neers In 6.
Intergrerralar Corrosion Test.
E Tarn. Ass'y.
'71 F 2 GO2 8.
Grain Site Determinetten.
Table 2.1.4-2
- m. _
?.1.5 Control Rod Drive Mechanism Pressure Vessel Housinc The tecords of the Saxton Pressure Vessel materials were reviewed for in-formation regarding the type of material used for each component, chemical and physical test reports, ultrasonic, radiographic and dye penetrant tests performed. The materials used compare favorably with those used in current control rod drive mechanisms.
The same material AISI 304 stainless steel, is presently being used for the rod travel housing, relief vent and cap.
The pressure housing on the current mechanism is made from aISI 304 stainless steel instead of AISI 403 (modified) stainless steel.
The complete pressure housing for the Saxton control rod drive mechanism was designed in accordance with Section I of the 1959 ASME Boiler and Pressure Vessel Code, including ASME Boiler and Pressure Vessel Committee Special Rulings #1270N, #'273N, and #1274N.
Fabrication, testing, and inspection were done in accordance with the regulations of the Pennsylvania Department of Labor and Industry.
All structural joints are mechanical joints rather than welded joints.
The canopy seal weld between the rod travel housing and plug.is 304 stain-less steel to 304 stainless steel. This weld was by the inert gas shielded tungsten are method using 308 stainless steel filler metal.
For the pres-I sure to rod travel housing canopy, seal weld, the 403 stainless steel pres-sure housing was " buttered" with 307 stainless steel and was dye tested.
I-The canopy then was welded to this " buttered" area in the same manner pre-viously described, l
i I
The canopy seal at the top of the pressure housing (for seal welding to the i
head adapter)-was formed of 307 stainless. steel and is welded to the head adapter by the manual shielded metal are method using 308 stainless steel filler metal.
l.
l l
l:
L 2.1.5-1 i.
l 2
__m...,,..--v..~,
m-..,-c
,mv,,, -,,,.., _ - <
,,.m
,,,,,,..,-,_i,__,,,,_,,3
,_,.,,.~,W-,-.s
~. - -. _ -
The only dissimilar metals weld joint on any of-the pressure vessel comp-onents is on the stationary gripper assembly. There are ten (10) joints of 304 stainless steel-to 405 stainless steel which are welded by the manual metallic are method using INCO "A" electrode. It should be noted that these joints are assembly type velds, and are remote from that part of the assembly which is subjected to pressure stresses.
T 2.1.5-2 l
I
'*4'
"'f"'tttw 1.y.m-pmy 9 p,_
2.1.6 Pressurizer Codes Used and Comparison to Present Design, material, and test data for the Saxton pressurizer is compared with the current pressurizer design in Tables 2.1.6-1 and
-2.
The pressurizer was designed with the requirements of ASME Boiler and Pressure Vessel Code Section I and the ASME Code Cases 1270N and 12735. These require-ments met the Section III requirements for fabrication, inspection and material requirements except for the following areas which are considered acceptable as noted:
1.
The shell and head material being less than four inches thick were not ultrasonically tested. The requirements for ultrasonic testing for materials less than four inches were established in December.
1967.
The shell material was a forging and was magnetic particle tested after heat treatment. The head plates were approximately-2 inches _in thickness; these comparatively thin plates did not receive ultrasonic testing or magnetic particle testing except for approximately 1/2 inch on either side of all welds, y
2.
The impact test requirements for the vessel material were 15 f t lbs at 10'F instead of 20 ft lbs at the NDTT as stated in Section III.
Prior to Section III, industry used an NDTT for the carbon silicon steels based on 15.ft-lbs.
Since the pressurizer is operated at temperatures in excess of 212*F, the 5 ft lbs increment would be.
accounted for, s
3.
The heater veld to the diaphragm plate which is a pressure containing weld is a fillet weld rather than a J groove reinforced fillet weld.-
In the Saxton design, this diaphragm is backed up by a cover plate bolted to the vessel. The heater weld as well as the adjacent center pipe support weld was dye penetrant examined both at the first pass stage and the final weld. These welds are similar to tube-to-tube sheet welds.
f.
2.J.6-1
_ _ _ _ _ _ _ _ _. _. _. _. _ _. _ _. _ _ _ _... _ _. _ _... ~.
4 The main nozzle velds were full penetration welds and vsre radiographed.
Present Section III requirements would require UT inspection because of the straight wall of the nozzle. The comparatively thin wall of this vessel and head, however, did not-limit radiographic interpreta-tion. These nozzle welds were magnafluxed af ter back chipping as additional veld assurance.
Materials and Justification The materials met the requirements of-Section III except as noted previously.
All cladding was deposited by manual metal arc, inert gas shielded arc, and/or automatic series submerged arc welding methods. In addition to a visual inspection upon completion of the first layer of deposited cladding, the final clad surface which was exposed to the primary coolant was dye penetrant examined compatible with the present Section III requirements.
Fabrication and Quality Control The fabrication and quality control except-as previously noted were equivalent i
to present Section III requirements.
Inspection to Date l
l The inside of the pressurizer was visually inspected within several feet-of the bottom of the pressurizer after removal of the heater package. No evidence of weld cladding failure as would be indicated by " rust" was observed.
This operation was performed prior to startup with the plutonium core, i.
j 2.1.6-2 4e l
~..
TABLE 2.1.6-1 PRESSURIZER Pressurizer RG&E
- Saxton Design / Operating Pressure, psig 2485/2235 2500/2200 Hydrostatic Test Pressure (cold), psig 3110 3750 Design / Opera ting Temperature. *F 680/653 668/636 Water Volume, Full Power, ft 480 **
30 Steam Volume, Full Power, ft 320 70 Surge Line Nozzle Diameter, in.
14 3
- Available from Ginna Station FSAR.
- 60% of net internal volume.
I li
TABLE 2.1.6-2 i
Westinghouse Saxton Current Design Item Pressurizer Pressurizer Design Code ASME Section I. 1959 ASME Section III, 1965 Special Ruling 1270N Special Ruling 1273N Mech. Design Report, Heater Bundle 668'F/2500 psia Individual Heaters 680/2500 Upper and Lower Head 668'F/2500 psia 680/2500' Shell Barrel Assembly 668'F/2500 psia 680/2500 Spray Ring 668'F/2500 psia 680/2500 Materials Upper Head A 212 Gr. B SA-316-WCC Lower Head D 266 Gr. II SA-316-WCC Shell Barrel Assembly A 212 Gr B SA-302 Gr. B NDT* Base Material Upper Head MT (in weld region)
RT + MT Lower Head MT (in weld region)
RT + MT Shell Barrel MT UT + MT Heater Bundle RT Individual Heaters UT+PT+RT l
Cover Plate at Spray Nozzle MT Not applicable I
Hydro. Tests Max.
Hea ter Bundle 3850 psi Individual Heaters 3106 psi L
Upper Head 3850 psi 3106 psi.
l Lower Head 3850 psi 3106 psi Barrel Assembly 3850 psi 3106 psi NDT - Weld Shell - Circumferential RT + MT RT + MT Nozzle to Shell RT + MT RT + MT Cladding 100%.PT Internal Surface 100% PT Internal Surface
- Non Destructive Test
{
L L
l.
.4
.. ~.. -. -.
2.2 INSPECTION PROGRAM A primary coolant system inspection program is planned as a supplement to the existing inspection program to provide increased assurance of system integrity.
The primary coolant system with the exception of the reactor vessel is reasonably accessible for inspection.
The program described below may, in some instances, be limited by accessibility prob %r s or radia-tion exposure to inspection personncl.
Such limitations ca, no practically be determined until the inspection program along with other maintenance work is planned out in detall or until such time as the progra ' is actually carried out.
However, it is the intent to carry the prsgram out to the fullest practical extent and to supplement it and expand it as necessary to investigate any unusual or unsatisfactory conditions fevud during the inspection.
Any
- gjg, unsatisf actory conditions found which ould af f ect system integrity wi31 be a34' evaluated by the licensee using such technical a(vice as may be necessary.
"I
- 4. i' t
V including a review and evaluatica by the Saxtor, Saf ety Committee.
m The program described has been inftlated and will be repeated at five year
~
intervals.
The steam generator and the pressurizer have flanged access ports which permit internal accessibility for visual examinations. The reactor coolant pump casing and motor ere available for internal inspection after the main flange is unbolted and the motor is removed.
P Since the reactor vessel is closely surrounded by the support skirt and this i
extends in a water tight casing over the nozzles between the vessel and the concrete shielding, the outside of the vessel is availan for inspection c.
only at the upper flange.
Limited accessibility to the inside of the vessel is obtained by removal _ of the head and the upper core barrel and instrument support frame.
It is possible to obtain sufficient access to primary system components to demonstrate that there has been no significant change in the materials over the previous operating period and that system integrity is adequate for 7
1 1
continued operation.
l 2.2-1 i
i.
A review of the original design, materials, fabrication practices and quality control records has shown that from a safety standpoint the comp-onent are equivalent in integrity to those constructed for more current
- plants, Operating history supports this point and the following inspec-tien program will serve as additional confirmation.
F7h d') Db " ~' '
s The inspection program will be directed by engineers experienced in th'e
- ^
~
design and fabrication of these components and non-destructive tests will be performed by personnel qualified to Appendix 9 of Section III of the ASME
/}
4 Boiler and Pressure Vessel Code.
ig )L v
?-
.i All aon-destructive test'.ng will be conducted in accordance with the ASME Boiler and Pressure Vessel Code Section III. Specifically radiography will be governed by peragraph 624, magnetic particle testing by paragraph 626 and dye penetrant testing by paragraph 627.
The inspection program proposed for each component is shown in Figure 2.2-1 and Table 2.2-1 and is as follows.
2.2.1 Reactor Vessel The Saxton Reactor Vessel has had the head removed four times during its opera 31ng_his t ory. At this time the available portions of the vessel and the reactor vessel head were given visual examinations. Each inspection indicated that the vessel and head were in satisfactory condition.. In addition, in 1967, all 36 of the reactor vessel studs were ultrasonically examined by the-Magnaflux Corp. - Nine of the studs were also.given a fluid.
penetrant-examination.
The results of these tests were satisfactory and-the studs were reinstalled.
The primary system and the reactor vessel have received 63 cold (86*F <
Temp < 212*F) leak tests at pressures in excess of 2000 psig and 60 hot
. leak tests, 41:of which were between 2400-and 2600 psig and 19 of which-were between 1500 and 2200 psig. The results from all of these tests were satisfactory.
No leakage has been experienced to date from the leak-offs-5 2.2-2
---__._--__--__.--._._n_
l which penetrate the multi-layer shell from a point just outside the inner layer.
To facilitate inspection of the reactor vessel, the head and upper core barrel will be removed.
The head will be pitced on its stand on the cperating deck and will be given a careful visual examination on the ID and I
OD.
In this visual examination as in visual examinations for other comp-onents optical aids will be used to reach unaccessible er high radiation l
areas and to obtain magnification of the surface being examined. The weld
- ~ ~ _
between the_ head dome and flange will be_given_afpenetrant test from the
's.4-u i
outside.
-m l
The react-r vessel cladding will be given a visual examination using a boro-scope from the flange down to the top of the thermal shield. The examina-tion of the cladding with the boroscope will be through several 1-1/4" holes in the core support flange which restricts the examination to the location of the holes. The bottom head will be inspected using a small television camera.
2.2.2 Steam Generator i
I j
The primary side of the steam generator has been previously opened and a careful visual examination made of the internal clad surface, the divider plate, and the tube to tube sheet velds. No unusual conditions were found and to date no leaks have been detected from the primary to secondary side l
of the steri generator.
t l
The steam generator will be given a visual examination in the area of the tube suest to channel head we,ld, the tube sheet to shell weld and the nozzle to channel head _ welds. These welds will also be given a penetrant inspection.
The man ways on the the secondary side and on the channel head will be removed to permit visual inspection, 1
l 2.2-3
2.2.3 Pressuriser The pressurizer has previously been given a visual inspection by removal of the heater b"ndle which penetrates the bottom.
During these inspections l
no unusual conditions were noticed on the weld overlays cladaing in either the steam or water pnase areas.
Inspection of the pressurJcer will consist of a visual examination and penet. rant test on the OD of_the shell to he.ad w.e.lds, the surge line nozzle
.. ~.
to shell weld and the surge line nozzle to pipe weld.
.. - ~
Internal examinations will be accomplished by removing the heate_r_, bundle and enking a visual examination of the ID.
In addition, photpg,raphs will be taken of the cladding in the steam phase, water phase and at the water to stean phase interface to determine if there has been significant corrosion or c:acking of the cladding and to serve as a reference for future examina-tions. The heater bundle will be given a visual examination prior to reinstal-Intion.
I 2.2.4 Reactor Coolant Pump The reactor coolant pump was removed from its casing on March 6, 1968, to effect replacement of a leaky flange gasket. When the pump was removed, the casing flange and interior was given a careful visual examination and no defects were found.
The motor flange was examined with similar results.
All flange bolts were measured before and after removal from the casing.
l These readings were compared to readings taken prior to the last installation l
l and indicated that no significant relaxation or creep has taken place in the bolts. Additional maintenance inspections were carried out on the l
oump bearings and the pump was reinstalled.
!Ao additional inspections
\\
l are n.ow planned for the reactor coolant pump.
l l
2.2.5 Reactor Coolant Pipe and Fittings in 1963 end 1965 the 30' bend on the 12" reactor coolant pipe at the coolant pump outlet received penetrant and radiographic inspection. The results of these inspections were satisfactory.
2.2-4 5
Y l-
_ _ _ ~ - _. _ ~
i I
4 t
i Radiograg ic and penetrant tests will be perforrt,ed on the 30* bend and ra the results will Le compared with previous inspections. The bend, a fitting and the nozzle,,to, pipe velds at the s:eam generator and rea,ctor coolan.
i pump will be given visual inspecti?ns.
In additfor the nozzle to pipe welds at the steam generator and reactor coolant pump and a pipe to fitting weld will be given penetrant inspections.
Radiographic and pen g nt tests will be performed on a pipe bend at the ouYEtofthesteamgeneratorandonthepipeinthevicinityofthecharg-
)
ing line connection.
l In addition, the pipe to fittir.g ar d fitting to loop welds on the suye line, auxiliary system return to loop and spray line to loop connections
)
will be given visual and pen,etrant inspec.tions.
2.2.6 Component Structural Supports Support brackets integral with the components and structural steel weld-ments from these brackets to the foundations will be given a careful visual examination.
i e
4 2.2-5
i l
TABLE 2.2-1 l
l f
Method of Inspection Planned Inspection 1.
Reactor Vessel cladding between Remote visual closure flange and thermal shield 1
2.
Reactor vessel flange and flange Visual to shell weld l
3.
Reactor vessel outlet.ozzle, 1.D.
Visual 4.
Reactor vessel studs Visual and ultrasonic or penetrant j
5.
Reactor vessel top head Visual and penetrant 6.
Reactor vessel bottom head Visual (TV camera) 7.
Steam generator tube sheet to shell, Visual (internal on primary) l tube sheet to head, nozzle to head welds and penetrant (external 8.
Steam generator secondary side Visual 1
9.
Steam generator channel head, l.D.
Visual 10.
Pressurizer, 0.D.
7 Shell to head, surge line l
nozzle to shell and nozzle l
to pipe welds Visual and penetrant l
l 11.
Pressurizer, I.D.
Visual and photographic 12.
Reactor coolant pump and casing Visual flange l
13.
Reactor coolant pump flange bolts Visual and dimensional i
14.
Reactor coolant pipe Visual. penetrant and radiographic
- 15. Reactor coolant pipe fittings Visual, penetrant and I
radiographic i
16.
Reactor coolant pipe weldments Visual and penetrant i
17.
Component supports Visual r
The methods for the visual inspections are being developed and I
will-be incorporated into written procedures.
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l 2.3 LEAK AND LEAR DETECTION EXPERIENCE I
The Saxten Leak Detection System is described in Section 210 of the Saxton Final Safety Analysis Report.
Leakage past the inner shell of the reactor vessel would be indicated in the control room. The reactor has leak off connections between head gaskets which is indicated in the sample roon.
Leakage from steam generater manways, pressurizer manways, and main coolant pump flanges is collected, indicated on a pressure gauge and alarmed in the control room.
The existence of pressurdeer relief valve leakage is indicated by sensing temperature increase on the outlet of the valve. The control rod drive room has a sump which would collect water leakage from rod drives and alarms in the control room.
Gaseous leaks from the steam phase of the pressurizer and valve packing leaks have occured and have been observed by gaseous activity increase in the containment. Larger leaks such as the pre-post filter gaskets have caused both the particulate and gaseous activity in the containment to increase and also have caused the sump pump to operate occasionally. Up
\\
to the present, leaks due to tubing fittings have been taken care of by the charging system and have been indicatgd by loss of level in the pressurizer.
Leaking safety valves in the charging and letdown systems have actuated the discharge tank high level alarm-and have caused-the discharge pumps to operate. Leakage which has occurred has been readily detected because of the presence of boric acid crystals.
I To date there has been no indication of leakage into the steam system (as determined by Tritium analysis) or past the inner shell of the reactor vessel. There has been no evidence of leakage through system piping or velds.
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3 DIERGENCY COOLING SYSTEM DESCRIPTION i
3.1 EXISTING SAFETY INJLCTION SYSTEM SDD1ARY The flow diagram for the fafety Injection System is shown on rigure 3-1.
The system utilizes the refueling water 8torage tank, two pumps, and the piping, fittings, and instrumentation shown.
All of the system except for the two injection lines and the two check valves is outside the containment I
building. The safety injection pumps are motor driven, centrifugal type
'l g
pumps.
Luch will supply 375 gpm at 723 paig with a shut-off head of 915 piig.
The pumps take cuction from the 80,000 gallon refueling water storage g
tank which contains a minimum of 70,000 gallon of borated refueling water during operation.
Injection is directly into the reactor _ vessel through two spare nozzles which are 120' apart. One injection noerle discharges into the inlet downcomer region thereby to the bottom of the vessel the g
other injects into the upper vessel plenum.
The safety injection derives from a low pressure (; 1000 psig) signal from detectors in the pressurizer dd-or the main coolant loop.
Both pumps start automatically and each of the
,gf, 0.
two isolation valves in the injection lines open automatically. Water flows by gravity feed from the refueling water storage tank to the pump USUd suction. The piping and check valve artangement results in series injection for the expected case of both pumps starting.
However, if either pump fails to start, the arrangement allows unimpeded injection by the other.
A flow controller channel is provided on each injection line to detect i
the abnormally high flow that would occur in the event of an injection line rupture and close the associated isolation valve thereby diverting flow to the remaining line.
The safety injection pumps continue to run until stopped by the operator.
Control switches for each pump are located on the main control console.
I 3.1-1
t 3.2 PROPOSED MODITICATIONS Recirculation System The quantity of water required to balance decay heat by boil off is small r
due to the low thermal heat rate.
Assuming operation at a steady power at 35 MWt prior to the loss of coolant. the following water addition rates are needed for boil offs.
Time After Decay Heat Required Water Accident Rate,(KW)
Addition Rate (CPfi) 10 min 1120 7.70 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 666 4.60 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 385 2.60 2 days 149 1.00 1 week 109 0.80 10 weeks 80 0.60 A fully automatic spilled coolant and injection water recirculation system will be provided to maintain the core flooded after the initial refueling water inventory is injected. Two canned motor recirculation pumps will be located adjacent to the axuiliary compartment sump and the piping, valving and instrumentation modifications indicated on Figure 3-2 will be made.
Design parameters for the recirculation pu*aps are tabulated in Table 3-1.
A recirculation signal (marked R on diagram) derived from redundant low water level signals in the Refueljng Water Storage Tank will automatically open the isolation valve in each recirculation path, start both recirculation r
t pumps and stop the injection pumps. Each recirculation pump will deliver
/g f7 into the reactor vessel via connections to the safety injection lines up stream of the existing check valves as shown. The capacity of either pump
-is more than sufficient to supply makeup water -to accommodate-boil-off at the earliest time recirculation would be initiated.
Spilled coolant from a rupture occuring in the auxiliary con.partment will drain directly to the sump.
Since a break could also occur in the reactor i
3.2-1
%N
~_ -
compartroent, which would result in spillage into the r torage well w.ater, 1
a flow path will be provided between the reactor compartenent and the auxiliary 3
cottpa r t rnen t.
This will be accornplished by extending two existing pipJng connections (presently capped) between the two compartments to an elevation
(),
[
slightly above normal reactor cortpartrnent water IcVel.
Either a valve
- ' O '. 94 0
or cap provision will be included in each line to permit isolation when
, g, '
the water level is increased during refueling. During operation the connections '
will be left open. The head of spilled injection water over the pipe connec-tions and the flow area of the 3 inch pipe is sufficient to gravity feed the recirculation pumps.
The pump and valve motors and the instrumentation will be suitable for the post accident containment environment.
Remote indication of the posi-tionoftheisolationvalvesont/futletoftherecirculationpumpswill be provided in the control rc,om.
Each pair of a recirculation pump and its corresponding isolation valv.)
will be powered from a separate bus.
A local pressure indicator will be provided on the discharre line of each recirculation pump.
If necessary the suction line of each pump will be provided with a connection fitting so that a small we ter tank can be positioned adjacent to the pump and connected to the suction for testing shutoff head.
[
The flow path for the test will be from the tank through the putnp and to either the sump or back to the tank via a small bypass line from the pump discharge.
t Flow Controller Channel The logic in the existing circuit for detecting an injection line tupture and closing the associated injection'line isolation valve is based on-a---
twoto one flow ratto in the two injection lines.
If the flow ratio exceeds two to one the valve-in the line with the higher flow is automatically isolated.-
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l To preclude isolation of an intact line in the event the isolation valve in the remaining line should fail to open, the circuit vill be modified
]
to require indication of an open flow path in the line with the lower flow prior to isolation of the valve in the line with higher flow. The modification
]
will include a position signal indicating the valve is in an open position to ensure that the flow path is open.
Safety injection Actuation Circuitry i
i The safety injection actuation circuitry has been analyzed frem the standpoint i
of an active component failure preventing automatic operation. Failure of the master actuation relay, operating coil, or its fuse could prevent l
automatic safety injection. With coil or fus3 fallore the operator would trip the relay manually. Mechanical failure of the relay would require b*
s st:a.
the operator to manually start the pumps ar.d operate the valves. To increase the reliability for automatic initiatfor the circuit will be modified by addition of separate relay, operating coil, and fuse in parallel.
1 sI i
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3.3 ELECTRICAL FACILITIES Three 115 )N transmission lines emanating from the S.S.G.S. are connected to the Pennsylvania Electric Company 11$ IN transmission system which has numerous other power sources. Also three 23 }N lines can feed the site.
These lines leave the site in three different directions. A house service generator with a capacity of 937 INA lu availabic for station service power.
The approximate time for bringing the house generator up to speed and sup-plying statJon service is approximately 45 minutes.
1 Power from the S.S.G.S. is fed to SNEC by two independent undergrourpfeeders.
4 Loss of the normal supply allows the emergency supply to close in automatically.
With the above variability of off-site electrical supply, complete loss
(
of power is in itself very unlikely.
With the low (less than 5 Mwe) electrical output of the nuclear unit, the' loss of the unit as would occur for a loss of ecolant accident has an insignificant effect on the system.
Hence there is no reason to expect a loss of power concurrent with loss of coolant.
Considering this and the backup afforded by the house service generator the existing facilities are adequate.
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l TABLE 3-1 RECIRCULATION PD!P DESIGN PAP &lTTERS Number 2
Type Subttersible Chem. Pump i
Design pressure - casine 150 psig Design temperature 250'r Design flow 60 gpm j
i Design head 235 ft Material 316 stainless steel NPSli Required 10 ft l
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LOSS-0T-COOLAN7 ANALYSIS hiloduction The core behavior following a postulated rupture in the Reactor Coolant Syr. tem has been analyzed using the latest techniques that are used on the current generation of large pWR plants.
Specifically the computer codes used for the various phases, of the transient aret 1.
Blowdown - FLASil Code 2.
Core pwer - CHIC-KIN Code 3.
Core Thernal - LOCTA-R2 Code A brief discussion of each of these codes is presented in Appendix D.
The accidents were analyzed for a core power of 35 MWT and peak pwer at the hot spot of 19.1 KW/ft.
Three break sizes were analyzed.
These were:
1.
Double Ended Reactor Coolant pipe-- 14 inches (1.28 ft )
2 2.
25 square inch hole - 0.173 ft 2
3.
Surge Line Break - (3 inch pipe, 0.037 it )
4.1 BLOWDOWN TRANSIENT The primary pressure and volume for the three break sizes are presented in Figure'4.1-1 to 4.1-3.
The blowdown portion of the transient was calculated by the FLASil Code.
For the refilling portion of the transient it was assumed that one of the two available safety injection pumps was operating.
An eight second delay time was assumed for starting the pump. The delay time has been verified by repeated system tests.
t 4.1-1
Tor the double ended break the reactor system volume is completely voided before the safety injection pump starts. The total delay time bcfore the refill is started is based on signal delay and pump start up delay.
The safety injection pump refills to the bottom of the core at 63.5 seconds.
At this time the refill is retarded by the boil-off from the hot core.
The core is completely refloeded at 105.1 seconds.
For an intermediate size break between the surge line and rea'.or coolant pipe,.173 ft the core also becomes completely uncovered, but the vest,el is not empty when the safety injection pump begins to inject.
The bottom of the core is recovered at 42.1 secends and completely recovered at B3.7 seconds.
For the surge line break, which is the largest connecting pipe to the reactor coolant system, only the top 25% of the core becomes uncovered in the transient.
Smaller breaks would have improved volume transient.
Core Power Transient During Blovdown for the surge line break, the initial subcooled decompression, about 0,5 seconds in duration, causes a decrease in nuclear power to approximate 1v 23% due to the conservatively assumed positive density coefficirnt-of reactivity of 0.235 (k/Lo (corresponding to the minimum negative moderator coefficient of -2 x 10 6k/*F).
Following the decompression, the pressure rises slightly and the nuclear power increases to 50% at 2.2 seconds at which time trfp becomes effective.
For the 25 square inch and dovble-ended cold leg breaks, the initial subcooled blowdown and rapid continued decompression introduce voids very rapidly and extenaively causing the reactor to be shut down within 0.2 seconds.
I h
I 4.1-2
SAXTON, DE-CL BREAK, AREA = 1.28 ET l
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1 4.2 CORE TilERMA1. TRANSIENT Core thermal transient curves are presented in rigures 4.2-1 through 4.2-6 for each break size analyzed.
These are a.
21 clad temperature vs. time b.
Stainless steel clad temperature vs. time i
Both stainless steel and circaloy clad rods are p7esent in the core. The high power density rods are rircaloy clad and operate at a peak power density of 19.1 kw/ft. The maximum power density of the stainless clad rods is 70% of the peak.
The f ollowing table summarizes the results of the curves:
Total % Cor+
Break Size Clad Melt.
Totai Zr-il 0 y
~
2 Double Ended 1.28 ft 2.3 4.6 Inte rmediat e.173 ft 1.1 2.4 2
Surge Line.0375 ft 0
0,15 The values quoted above for total core clad melt are those associated only with the Zr clad rods. No melting occurs for the stainless steel clad rods.
Thus it is concluded that no aciting occurs for a break of the largest connecting line to the Reat: tor Coolant Systems when the core is operated at 35 MWt. This is the break size for which the Safety injection System is designed.
For larger postulated ruptures up to the hypothetical double ended severance of a reactor coolant pipe, clad melting is limited to 2.3% of the fuel clad area.
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4.3 CONTAINMENT TRANSIENT l
The buildup in pressure in the containment building following a loss-of-coolant is calculated by the COCO Code. A description of the COC0 Coda is contained in Appendix D.
f The pressure curve in Figure 4.3-1 is the result of the transient for the double ended severance of a reactor coolant pipe.
Included in the transient is the mass and energy from the complete blowdown of the Reactor Coolant Dymtom, all of the initial stored energy from the core and internals, all of tae decay heat of the core generated during blowdown, and the energy i
from the Zr-H 0 reaction including the energy from the hydrogen oxygen y
recombination. The peak pressore 19.4 psig which is.censiderably below the containment design pressure of 30 psig.
l In Tigure 4.3-2 the blowdown from the steam generator-is added to the energy for the core in the previous paragraphs. A peak pressure of 24.5 psig is reached which is also considerably below the containment design pressure.
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CONCLUS.!EF, Analysis of core cooling system capability indicates that clad melt is prevented for credible break sizes (up to largest connecting line) and is limited to about 1% for bres.ks as large as 25 in. The protection afforded
(
by the safety injection system exceeds its design basis.
For breaks as large as a circumferential rupture, the system has capacity to limit clad melt to less than 3% of the fuel clad area.
Even for this case it is felt.
the lattice would support fuel and allow adequate cooling to prevent meltdown.
The analysis of the containment pressure transient shows that even for the double ended coolant line break, the pressure would stay within that for which it is designed.
It is our opinion that with the proposed insp.ni er program and recirculation loop modifications the system is.Acquate to protect the public.
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4 APPENDIX A.
_A. O. SMITH C010' ORATION. REPORT ON SAXTON REACTOR VESSEL C(9(;j) i.
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., IJ6~506'~7 c;5 ~.nen - 65.6 i
2 I
~j i
4 l
l l At (62% ~.~
l ~~~Mi 95,120 ' ~ f lj7,650
~
~ ~ 18. 5 ~ ~ E o fA - 60. ?
'~
i i
i I At T21~5
-~
~~
' - ~.T15,MO i
i
- Austenitized - af~,15007 ~0il-QuentEed -l 1bmpered 6 h'ours At 'lU21W ' ~ ~
~ ~ ~ ' i~
~
f
- ^
~ ~ ~
- sY,ress Relieved h(E6urs'at 9FPF I ".
f 0Fa~in 31:o 6-8
~
~
, Ultrasonic Tests satlsfactesyC~ Electro li1E #77-5.0002 to.0003" th2 by Pyro f.:ataf finishers i
.j Rockwell Cl Readings After @ellilWdmacIdriLiga i;_
i i
1 1
}__
Stud No.
Rock C
$ 33-
- 10. 32-32-32 !l$ 33432-32i 20 33-33-22 25 32-31-31 30 33-29-10 35 32-33-32 LO 32 '2-20 l~~TT 31-33-jf-~ ~ 6~ 33;31-33ji 3i~'i1C~ 32-31-32_,[1_6j)432-3ff ~21 ~ 27-132_-j) 26,(-33-32_31_27-3
[_2_\\_))3-2 - L 33-3h33.12 32-33-32J1L30A33 33L22 35-33-33 21 29-33-33 32 33-D-n n 33-h
'?
V '-
I 3
. 32-33-33
. 5 33-31-32 13 33-37-3? '18 33-32-31; 23 33 33 32 28 33-3;-;2 33 32 J-3 la 3;-32-n 1_ h_i 32-i-33__1_.S__32_3h33 6 30-33-33 :19 33-32-3Oh 32-3p33; ^29 29-31-30 ; 3h )3-23.n 39 3 M.t-n i
]
,a::.
4
.i
l MERIA( RE,COBD FOR F1.ATES, HEADS AND FORG'NG5
.[
c.ssu d_J%stinghan=a Flact Pie Cecp. A.P,D:
- 19. Main Nuts l
ShR-8026 20.
Spherical Washers l
Di t t.shuwah, Dannsylysnia
{
A,-ony.,
t a
i~.-....r rvec Rasetor Tenne.J.
t Sarton. Pennsv1vania l
ee:2Tewion CRAW 1HG NO.
l *l 4 O. S SERI AA., NO, m m. -._
1 8'f1CO'SS EOUt?setNT O'v10sora j
67-360-h5 SHfDPING ORDER NO
,,m,,,c y,,
i C4TE _ __IIthPH4PF 27 19bl
.Taden f.
Tmsv sem f
n i
4 mu Ttsr r.u o_a _r_.
.h
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L, -
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4 4
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p CHEM tC,.t e.e sscM.
I I
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l l
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l l
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'1 WL in0163at5.1 F codtinyg I
19,_ J Standard._. __38213 f__,.38. __. 72. J13).ni s '
.11.._L 78 '
77:_.91 ? 1rf..nnn
- 122 7tn g 7 A 1_f.c 4 An,g' i
'Iisid !stre_npad 6%oce. o1.onn t
h 1
I-i t
_l_.___.
riva Natch f.gact. Teat.=_ eat._ in??_ 35 l35 5-36 5 etk 18...
!-- __n
___a I_Avar,anj t1 zed _at.1<50'F I- 01Llonantched l
Temered_ 3 fira.a.t QO80 w i
0 i
_ _. f.._.
f_ttitras eie tem f - =et9'afae_id'y-(__.dlactri film '!.77-5 h m03_ Thb'.hy Brq yetaLvd shara
)
.___L H ackhe1L_C_Fe a din g_a rtir._prilirdndry _Jea dhi n ini. _ _ /
l l
'_Mut_5c. Rock C_5 133-27-32__10_.33-10-31 1 r in-32.in 7tt32-21-30_25_29 _28-30_30 27-??- 25_35_.2%2S31_lg.30- 3
_1_..'3L & 29 6_17-2031_ll_32-33-?3_1618_1d,?L7JL_31-2129_ 26. in-2%29 m m '>6. m M n' Tn 7't L1.31_ n-3n f9un i
30-32-32__B_ 2-3.b32_23_28e3M33.3 8_31-31-26.2.'L27-2{. _?? _26 -3i-33_32 33-?
27-27-29 7 RS-29-27_12 11-28-25_11_.10-23.- S_22_2S-? 76
>7 99 u 1.?
1 1.
2
_c
. 2-._.
i 28_26.33-??,?S_33.31-31-32.3A ?%3%?5 ha.I.??-2M7
.1 3
h..
29r29-29.._9 2-33-29 1h 33-31-30192R45-?? ?! ?7-26.?? ?? ?6 91-?R % ?u ?*_ vi 39. n.y y.;i e.
e,,
ey
_~
J
_..htandard__.7 _.382?3.
_.. 3R L.7?_f...n11_j.n15(.31 1,7817_7_J _.23 311,rvvy 138,g<o ho < it. ce 3.37 c j
' m ld Siro,6+w a !Acn"r 37,<na
.l j
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/
__ _ V".Notchlspect_Yeata_4t In'i__35-19,5-33 Ft. b _
l i
L
- l. Austenitised at 15100FJ _ Oil'Ouenched__lTerrefed_1. Jour 3_dt_10960,
l l
F
_LUltra s_prLi c._Te pts _- _.s ati s factd ry___ _ !
I L _ _.__ !
'I _
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. Roelore.11 C. Reading.s afteLoreliminarr Maghin1ncr_ _ _ _..'
294L,_32-29 _.J5-28, f 3%31_ L1-30. _ _
% e Washm___. l2dL 127_._8.c9_1148_;A-31.17-30_2048 2F27 2648 l
1 F
R Rock C jJ-28 6-29_
9-26_12-26_il549 i_lB-28_21-as 2h-]o 27-27 3c,29_.33 26. 3649J3948 Lo29 i
l.
!..t_
.. 2 S_
b g9 7-28 10-30 ',t-30__'1638_L_17.-32. 22-Jo_2y7 _28-27_ 3148 3:147 37,30 3 48 h_* T3 'W.
[
- p. jade. - Washers L_
- _ __ __I I
i
{ 'Jo, : Rock c IL27 6-29 N 31 1'J.28 815-29 _1%?ff 2L,R,_ph-yl,_17_12 ym 7p 11 23 1A_27 39_2C 'L2 2S
[
. _ _ 2 7.
h-31_.7-30__.
10 2p_13]31.__f16 10 2k 22p27 25 78, 2 A 3
38 l'
3, 99 77,37 s,g_yn L 3_ ^;
{
4
' I 1
j
-.. 2?___. 5sfl_S-12_.
11-30 4 48.
, 7-28 20--31 23-28_ 26-29 2947 32~.10 35-? 33 27 haa2 Ope _:
'1 i
L l
t t
(
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MATERIAL RECORD FO' FI.ATES, HEADS AND FOCGINGS j
Meetinchouan_Dectric_Cor L__A.P.D.
cuerourn g
22 Nozzle Neck Plate (Item 47)
I 4:.c a....Pittsburgh, P m ay3xania 23.
Sbield Support Bar F
24.
Sbield Suneort Pad Re..a.c tor. _Yessel o_.....
iv._.._
- 25. Shield Support cussets Sexton,_Perrisylv inia and Shield Lifting h2gs r,:
r,,,y,m
- 26. Shield Cmtering Pad end Ceter!ng Pad.
1 0 00016
- 27. Shield Suwort Rinr: and Shield locatine Lucs Pace 5 af 8 j
. _. _. _.. _...~ 117 Rno16-1~
28 Outlet ?!ozzle Shield and Support Clips
- o 5"'
c" "67-360 (It~ 157)
" " ~ ' " " '
s,,,,,yc onor, no,
- 29. Inlet Nozzle Sbield Febulazy_R7,1%1 30 Cmemtric Redacers John L. Iverm one l
S4tt_1 TEST REPORT h
b C"
"
- C"'
wamsreensncn us17: clan mNo w=.
l e
s Si L Ni
___"__._._._____I
= =_Co__4-. et war l'n.' = 'ar><l c'.e u r ! m e
- ' 5=1 5
.Cr i
I itY 8001 6 fin t 81 G - ASDf AN 1 TypeIJ04_-fatarial rwk 1* +o_ha_k.ali _.sy 9.a11 M-- + a - -, !s_
_I i
J.. _ __
.',11trasonie testin er s.*sived on _1 t,* is suppnr_t_ skin 1 A ha,
.a,4. ca4 +m tog a, I
. 21. f I_A31eEberty_1._11699-3
___j038l_1.50_L.016_p23._.43i_9,43_[18.46nonn4 37 500 ' 80.000
'61.0iR.orAJ-2Los i
Prihell_$rdasr_. _J4t2-1/,9 *_39,en
~_rn,nen _ _ fo n_ i;e nra_lM.%
__._ _._Ludlite I
i i
Ultrannic_.Iests - 5ptisfaltterr I
i i
i J
l e q ie_7.e.
'av %,i - av
.__22__!_A31gumhenI_L.79]?0 d_.0702.1.58,!_.025__.015 j _51 _'__9.63 3 8.40I _.060_35,900_ _.82,onn
- 65,0_ R aea A n]
I L i
6 i
.LDrinell_ Fardoces - In
_ oral!c rent __og rand.!- a6 !__ _.J i
l Etrasonic_Tes't,sawilty_A.[.,_ Smith ___-leatisfcctor._ _i__. _h__ _
0
+
e 23 i Alleg any 345fz60-9
~,_055l.1.43_e_.921_._.012_L40 _. _9 Of 12.2a,o30_ __gi em : 9,3no__'
_oo..,
,s. o1 4. -74.c,
i
~ ell Ibrdness -136 i
breliel Test - ok LBend -iok
~~
i Ludlum i
i.
i Br
_[I _ AlWheny_j 326682-4
_,064_i 1.561.,027 f Oi7._.353 ~~8.8315.17. 06c_36 400 ! 86,700 J *63.67P.qfA a 2.o:
3 3
i
! Ludlum
~
Dritiell Hsrdnese -163 5 Oxalic 'b'
- ok Beid - oh l
I
' 8 ~.~~ Allegheny 1 32iS6_9-11
' 054_*1.71$.025i.012.61 9.14 18,RTif~~i9fof.-
e1,300 _
'7f.6 ; _RRA3Kl.
l
~-
. __. ___I_ _L__ ditan l
?
f~ Brinell_ Hardn_ers_-14't Oralic ! cst-o!: P e d - ok u
26
- Allegheny _l 23868-1 L063!_1.57!_.0211,018_L.50_!__9.7s._18,48_.10'._42,/00 81,5io0_
'62.0.:R. ora.-69.5 r
i i
___ ___ __Ltdlum
)
Prine 11 iRardnesa -143_.
_ _. _Oxelle ~ cst _
..ok Scid t-ok l
27_ j Allegheny,!
23_610-2 M (l.71 L 021 L O101.26301 !_) 8_,50,0/.O q 41. Pro,
85,100__. "60f R e(* A -7f. 7-Ludium I
i
_'____ L _ _.__...___ Brin ell ;Ihrdnec.1-149
.Oxa11: Test - et Pend - ok.
L
. _ _ _28_ l._U.s, steel 7I2235 _
i so,5no. *5s.o ok Enbritt1Em.03_ Ll.?_6. '. 028_'.,007)_.59.f_9,36_1.8,32 _.14 '__L8,6no,
t.
l
_ __ _ J L
ent...!Testj_ok_Jioekvell_Ihr6 ness 464 J1ex 4'VnO
'M,000 _. *(Ef' ok u_
l 129. LO.o Carlson1 60177
'_.059. I l
_1,11 _,024_i_A14.'_.64 ;_101_18.49 11 40,1to 81,ecc
- 9.0 et In 80016_Jat'l._J_- 1921_4403 {iGr,_WP 316 I'._ Rock *ll-E-R9,34 I
2 4
I_
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'51.6 n.orn.-72.'2i Forge i.
1 l
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[
1;r" MATEaIAL P2 CORD FOi! Pi.ATES, IEADS AND FORGNGS i
cusrouch Westinghouse Electri; Corp. A.P.p-
- 37. Nozzle Necks (Item 132) f 54R-8026 38 A1110 ment Pins (Item 54)
Pittsburgh, Pennsylvan1*
- 39. Cuide Ftnnels (It.ms 167 thnt 174)
[
r.carces l.
- 40. Cenning Step (Plate, Su, port t
j
- =s::r rvrc_Manetor Yessal Plates, end Inner Cover Plate I
- 41. Cem1:;Co P1tc
'Saxton Pennglvani_a 42 Canniig Inner Shell and Interacdiate
[
j nesrmarion t
Shells
_e I
IN 80016
- 43. Canning Oute-SI-ell Pc 7 of 3 f
oss.wim m i
- 44. Cerming Shell Relves
.. n s -. wa _IW. 80016-1 z.;,
g g.g m 46, Nozzle Tubes'(Itens 159 and 160)
- ro: css c =w ow's:==
r l
.m:,,:= oneen no __67-360-45 m vuser wie.cw t
)-
February 27. 1961 Jofra L. Ivarson j
w mye t
t f
i utu. TEST stront i
c-i
~~
~ ~,fp_),
l arem 1 Si
{
l I~~ ~ ~
~ ~ ' 'N1_~ Cr t'e-e e n' a cr cra emai. aa g
o.
I_.
u w
e ASIM A240 type
~LO44_ J,561[316T3191,01A_L,31_
_13.25 A36_.12 L35,500_j f"T--~~MY 15016 !!a t 'l 1{
l
- _00,500_
'52.0_iP,ofAJ-7t,.0 -
i.37 !c.0, car _1 son.
30030-1A 1.mme_-tok_ _I r
I!!olv-Q 26 i Brjaell' Ha
.s
_ 143_ _t
_Ultrebonig IIest l satisf2c I
3 i
t i
i~
1-MY 80016 ltst'l L AS1H A?<76_T3lne_._304I 1
4 I
L i
Lo22 _ 012_
_55 9.1$_18.6'-
_17[ 50,500 j__BS,000_ __'52.0_fR. aft-71.0}.
i i
1g___G,D c.rlmah 18101
'. _071
_79 l
.i i Brida11 banead -159 i i:
Bad _- mk 8
19
.c_o_carimanL.30068 8: 015 1 21. n22 !.0161. 51. 1 9.1'L18.5d,12 i_ 34,000____80,500__ '62.Qa R.ofA< O!
~
1 I
l_ErinM11_Rnrchesa_l-326._l I
I l-L i
1 ' astm m2Asi ce_ a Flanse h=15tv - M U _Staal I
I I
I
[.___l 2 5. 0__ !_ok_ L_
j 28.0 _ '_ nkJ -
- __ in n s_stani con 642-7-1
' !.11
.zs Lnnn mzl-I i
_j 15,7nn s6_n2n i
15,61n !s1;680_
._ 41
_._1LS.Staal -
400652-6 1 6,11 i 45_.I nns
- mzI-
- 11
- 100__.> 53,890t._._!_.32.0.dok__ L _ _q l
L 42 I U=s_ steel 10 % 11 1 yzip i 12 i _n1n _ M1 1 -
p 2,too m,soo__;,i a = n u t _,
n
- u. s _ se.,.
<n9sas-u
!.u un r_n1n
-mi t -
i-19,250-4 'il,Z7a___L33.02 ok_L _
r i 11 l n_s_staal !
100A11-6-1 H 12
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l_ _
i t
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I i
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~
ll (no_I 1.7%l__. nit 1 _n11.1.12 110.4*L 1 R,M L-J__LB,000_._ L._90,400__i.66.0__J _ _ __ I __ d j-j i
IF1.esl Test l-ak 'lydro '
12500_sei-ok 1.Jhekde11 n _75-78.
i I
{
t I
i a J1.a.et,e h.inding.,Jf4_hi_mt.Janner kuntaEquenebed,_-_.____ !
l i
dialdd vers ! dye-gaatrat_tedted -belder's.qtmllfied in 40E Code i
I-t 0
Fermino. Ram theat,_Weldi'ng._and _ dye _dmetralitLpedforsusd_1y_ GEinn c11. 4ro. _ _
I I
'I L
I astMf a264 Twpa 101 -c_bnt Mer cowit 1 AA i n-w aev s, 2047 L.066_ l.56 t_.015.',017
,61 in-4d 18.63 066 '_ Elare.Teht - ok E_
J Wilcar li l
_ _j l __ _._ _ _ _ ' Prirt, te:.:t -1000 poi - nk I.____j 2
i t~
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MATERLE RECORD
~
Westity; house Electric Corp. A.,P.D.
TCi! RfJEE lO,DS AtG FO.iGUGS cusTm en-W L7.
Pittsburgh, Pennsylvania Stud and liut Clooix Scrwe amre s h8.
Bl.codor and Frensuro Connection Tubing h9.
Inaula tion m.=w n vree _ Reactor Vessel 50 Ccepanion Ring I!alves Saxton, Pennsylvanis y,,,,
117 80016 unas m no. -
117 60016-1 p.g, 6 c' d
. o s seniai. no suunm aer.>cn no 67-360-b5 February 27,_1961_
w ew=
v.ex am oart
,, John I,. Iversyn1_
-h 'dANtFACTtp8tEse u s u. r en u n n.
)
cr em2.
MC3.T; St.A4 A*tO mue
_,c j
J h7 Eilliobrn.
i s
=
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ut-
= ~ _- w. m. y y_
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=
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Univ. Insula tion Glass F!ool
_AST! ( SPmCL A26L Twok J0h oor i Fed.' Unec! S3-C 'i66A Tvm ? Fed.' SoectifH-I-$63 T 00_1. Cinn.R tinbe=WL wl Per i
~
~ Comp ~any Asbestos Cloth 7
~
ASTM A212'Gr.B P.Dx.~Qualiti,'~
-_? _.29 T~.83 r.oII~r.022
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_ _x; fires october 72,1%17 I-r
A. O. SMITH CORPOR ATlGN M.l.a.a en,. Win enein v.,, e.au, n Reperi on. unii,ed tren.,e v. net
.ufri. Shop Job Nn E NOIS.. 7 u..<.f r.n, e d t,,
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Maa !wm J in,vestity;home Lleetrie Corp,(APD)Pittsbur Pmnney v, Pop,,4,,,,.
Ord., No.h6F
- 2. b_. _.
i ania 8 ial FVf t
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i
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4 He.. rr.31 test reports be.on decled ca ell the plates or seemless vessel forgingi ente <ing th; unfived preuvre venetne..a I*48 D o e s Pr e te r;el conf u.m to spe cif.r etion ?.._
. _IM. _....,.
5.
5hn't or Druni-N o._.1._ _D;a meter. h Ft.. 1 k._..~ln.
Length over d_20s,2 J/16. i.
- t. - Stempi on SAct P.ces _ _,
_..Se.....e Adde rg) _, _ _,.
_._.,ea,, ASTM alp) Or.4)kO_.. _,_.
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v.b.
_f 6.
Jcic+i Re d:og,eph.d _.S*4.12.
....._ _ Vessel Stress Rel.eved. $*1 ll. (Yes or No) Hiir.iency of Joint 70,,,,,.___ _.;
9 G;.M Jo;at.doub,le_yeld but,t_ joint.._
_. No. of Courses.__..
1
- 10. Ou+e,Shed _
- n.
!.tvi, of Se it Leag:e d;a.l
.. " G:rik
..._ Length of Sectinn or Course.
u L0a r al dish,bottcei 30= e a.$..
. i>.a k..N2 55 to c.
in.
k divi o lauckle__. -. J0" h _ _
- 11. Heads (th. laeu)
A,
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Redivi o 5 m g ua,,,Q
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g t pem.ip **icalopu 1/ *anar depaait. bgnc wood engle il conical." "=' '. ~. _ ".
Bot, or opp. end. 8 gM.
te i
Il remevabte. heed botti used_ M.
- k"..!_N
.oc method of festoning NM II188 M....i.c.m.6.=iM F
- o....,
4 e...
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12.
Red ~q.aphic Inspection All oc Fer ( nnt Thickncti longih.dme! Jo;rai 888 adders (a_,,_
_ _,,,;n.
e b
Circi,mf ru.t!al Jo;nti 989.tdd8Et!b ___
..._,.,._..in, streu Rel. ving Headi Ring Ncas.
Contro!!ing Thednen Temp. of Yeisel Time Temp i Held w
If pa,e of venal only __. p_ adde n(g_, _,,_,,,.
- n.
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If entire veuni
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I3. Nonic OudeYin bh2Q%,t,9_gg gEc.ist. or Reinioreeb...
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No.... h_.5M 1.II_
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4
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- 16. Mn*nnd nf 3.,ppe.rting wesiel
._D%RPll8 T1DE Maldad. ist lap S.a.i. i.. is n. i>
ag4L u.... 6,s.e n.e n..
4.
- 17. e Allowable w orking pronura et etmosphere tem, location of yield if yielding occured _
~ " _
e poveture 15ee W. R and F 525) _.
2500,_ p,;
I Hydroitette tuit i+reis in lennitu.
b Hydroitene test prenu o.
.).Ik _Pii dinal joints (W. ennels onl )._2kaMO
.Pil y
Hydroostic tco pressure when hemmor tested
._ Psi c
g Allowable operating stren (Two.
d Proof test penoure d applied
_.. P41 thirds stren obtained in () 10.100
~
Pd
- 18. Constructed for p'. m.re of 250Q._ P6. With specified operating temperature.6_50e._p, w;th c,,,,;,, a. ance
_,,,3,,
Remeeni. 4944%.9r.It#991..._#.N' Allig alad.12Easr All d a4 i ^ --- %d* layers 4
WE CER11cY the above Jeta ta b.. correct end that eft defeile of meteris! cometruction and workn ansh*p on,tbis unfire assure vessel conform 66 A. O. Smith Drew;ng No. W 80016 Dete _.Januar7 30, _.19 61 5;9n,d A. O_,_(>in, h co,po, on "
- Manufacturer By January 30 8... 19.6 1.
Cheded by.
.M _. inspector Fo b IN8 g nr1 l' N +y C e.
M20. IIAlv)
l A. O 51 A.Tli COM.
<W A10Ml; & Fa006%
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I D M 8a i W +"a Manufac t.aroi Report < a on Unfired Procoure Vescol Nrm. p3 rania Special IW lL67 i
Renete Vcnael A. O.S1M r.T n.
IN 80316 A. O. E : 4
't :,orict l'o ;.N 60016-1 l
aint in.:'. mt:se slectric Ctep. (A.P.D.) 5s. 0026 6.
Stampe on shell platee r Inner Shelli 1/2" ASTM A212 Gr.D 4'00 - ritJ 1/C" rim f.26L Gr.3 (7301 ) Cla:.d'rz; 70000.11 1 15 4
Shall Ig erna lA" ASTU A212 Moc3 fled - 75,000 pel Mrs Vecael Flan ;os:
a ASE A105 Or.II por ASn.1350 f:odifieri (.26.1 hx Carbor.) 65.CD0 pai 13 l
Top and Dottcra lloedne l
AST!' A212 Or.B FDG - 700CG pai.n
- 12. Radiographic Inspectior l
l Longitudinal i
l IWt bontt Thickneen Inner Ltell la 1/2" (cfter alloy dnpaof t)
!!:rzzlo (Stainlare !".12,i " /L2A) i Extennlorv4 on l'Q's,, d]A) 1 3/8" l
!a2,1:13.lflh,1G
.. 3h l. 1hA) l
! L15,Ill5A)
Thormal Sn1cid
- i. >
3n Skirt
,i j f..a Se l Flary,e hiry
- 11 J/l. "
a Cin:unferential Top L ori in Flarg' C f 3 /6" be ore alloy deposit L L/' after alloy dcposit Shall ta 'le.ry:. 5%TH CORP.
l l ETE lp2((rlg? '
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ue wAwF., WiscoNs!N MJ M j'M" I
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Auditional Nondoctractive Toeto Ultrasonic Clad inner chcIl in"p'etod for bond.
Voscel for,;in;;a inapoator for soundnoco.
Vessel Head Platos ?.n1po.:ted for nouroneen.
l S t.ainir es steel. platos o.or 1" thick inspocted for soundness.
Plates for aup sorts and lif ting lugo inspected for noundrnes.
Studs,, ruts arut n whera 8.nsamted for eatmdnoss.
Vessel aca=a Ci. and C2 inanocted far t.oundness.
Nozzlt. to Sheli swis innpc ted for noundnesa.
!!agnaflux Outer nurfact n! : oa,;itu ti nta Soa.1 (1,1) af ter strino relief.
All shell lay 3r aco '.a uf;ct ench laye r of weld.
Outer rurface of C3 cunf.tront$nl tiaras C1 and C2.
Nozzles in Shell (til2,04711%U15)
Inside turface of attachnant viteld (before al.lo donosit).
Outoide nieface of attachment trid.
- r., after aomoving r. eld dan.
b., arter voldint; fillot.
Head Seam 1C1 t..ttaiz.c and liftina lug welds after stre n relief.
l Root Pass arr! f tnished mld of skirt ring to end "latt.;e beforn etres:1
- rolief, Root Pann and fin.i.cred attacit.ont welci of cen1 flinae to er:3 flan,;e beforo stross roller.ini r :niit nt tnr ctress rolio!.
l Root l' ass wrj cocm): tod vit 11 of rion.1 flan,;e asser.Ny. 'lemy,naflux after ntroen reli<f.,
Skirt lanc, mean in ; ronvn arer ruging and final wold insido and ou t:1do.
Root pasa un t c.onpj < t.o 4 e la of okirt, sho.11 to base and skirt base q2arter se.w.
All conms of M:itt at. :o"1 b' stol atrcsa relief.
Root pass and d A 1 n.:: 0.' :1! irt to ring, skiri ccraent vol de and noal fluri,;o cu.;
f r. of vornel nesecbly.
&fo Pemtrant Inner Surface of Lon.:itedn.al Inner Si ell ream (L1) after streus raliif.
Alloy dennnit heiida of c -
- c. inh..*ontiti act.ms C1 and C2 No zlea in Shel; (bli,IIIN'S. !"LS) z Insido alloy depcsito.' mo'feco and nock l ort;itituin d and circtnar r.ntons Mn LW. t t.d.,uu *, r%r stress I olief axt n T:hM15, cif ir.ci de
.jienetor.
Noenln sleek to educt r nr.. - t t 1 tc)ci n,t outeide.
Alloy c%ni t. n.'
nar.. In v:.%chr e t welria after depacit.
All de mri.ed n' vfr c
- t. of *np,u! bottoa heads and pad violds in bottoa hett riNr
':ront rel All ncrm:ie..d o rf. cc of
..g rl:av;o af tor rtress.rnl$mf,
1
- .iv S0016-1 A1d2nda
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MH CORP.
If S70f ?!!
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I.*'m'n M4 NAva;IE. Wl',CONstu
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are 1%netrant (Cont)
Er.d flani:e and insido projes t i oi n o..1 :
- lil2 and N15 aftor rac n: 1:
All welds of t.horml thicid sor. Aly af t.or strees relief and c.acninity;.
Attacnment wolco o; centering i.de in vespel shell and ndapter:, in trip and bottom heads.
Stress I(elieving Part of ves::al ent/s Item C ~'tr allim thk.
Te:p T!rn Helo Inner Shall 1/'"
H50 F 1h h a Top Head Asnembly 5 1/2" H500 F 6
Top Head Assernbly(n ' o. apnsj '.) 1/8" coposit 11500 F 1
Bottom Hoad (after acpnoit )
3/8" daposit )
and welding support luco) 1/2" and 3" fillet)
H500 F 1
relds
)
Top Flange (after deponit 1/8" deposit )
and support ring nr.lc.)
3 A".1" fin e t
)
1150 F 1
acide
)
Shan !;ozzles i 1/?"
n50 F 1 1h Skirt Aener.hl) 3/h" 1150o F 1
Seal Finnge As. !mbig 3/h" 11500 F 1
Thormal Saield Anact.bly 3"
U50 F 3
Nozzle Outlnte L, Lo'J 2 - 10 1/2"la ');12. J1.)
!!o ic.. nt A330 per A105 Or.Il taodiriod c
(.26% llax.Cs,rbon, 65,000 poi IIIS) ra.bmcien -l 14 A2ho Type.'.16 2 - 101/2"10 VA,0.La.cr Po:
- /...Ti4 A350 per A105 Gr.II lbdif tod to 2 1/8" ID (',13,116)
( 28% IJax C4.rban. 65,000 pai 1/U) t;t4 nw :1 -icTh a2h0 Type 3Oh he i mer s.hT..; Ah03 Type 3M -
Tie cortify the nao.e chu.
.o ao c-
,M taat all dotails of catm ial,. conatructirn and wcrkmanship on this un.' ired p: o.m vem:n1 conform to A..
O, Snith Drnrin., %
W 80016 4
By (A. ~ 'Miz& m.,,
/ //
Date January 30, _1761 Signsa A.U.nith Corp.Efgr.
Jolin L.
Iverson January 30 1961 Chocken b;[~h4,M8@ctrIr f:2r Mar riand cacualty Cr.,
y EC 16400 t
+
W 80016-1 Addenda m
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VESSEL WELD TEST PLATE REPORT. A n
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o_us..e Electric.. Corp. A.P.D.
9,,, - L" p- - -
pru..ng Nebe' N $4 _. _. __._.
... __Ve selType. jaas,at.su Yeseel bul Nm,,,
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Movin e el Spe n hc ot.on ASTM MM,.Nedif1d
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=-.
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Weld; Ijpe RamAng]A 1 ne Now _,....Longitte = 1 Imy (2sps......
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Stress Relieved
... Wo REDUCED 5ECTION TEH510H TEST ( Aciens Weld)
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C. l>]Vn 3571 2m Irenus Gellingo Fb 7 30-56 j
C. Pleu 357:
2(D 11enus Collinen F5
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P1ms 3571 5
_ John Mvyaombourg Fh b2-56 C. Plun 3571-
?'G John Miyaecnourg FS 31 517
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P71 titurica Chrletonner Fh 0- 15-56 C. Plun 3671 l
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C. Mtus 3!.71 v
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C. Pitus 3673
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Frank itadomski Fh 9-7-56 C. nun
?,371 J3/
Oocar I, loo Fh 7-31 C. P2.m 3571 3%
Sdv:in 't'esch Fh 9 6-56
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- s.lter Daten.n Fh b?2-56 C. Plun 3672.
l Tialter Reinhardt e77 Eh 3-23-59 C. Pita 3671 i
b79 Jn1ter 11oinhardt F5 3-25.-59 C. I' liv 2 3671 i
' hiR Donald Lucc a Fh 422-$6 C. - F1te 3671 l
h:1.1 Donald Lucore Q21b N13-55 C. Plun 3671 t
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A. O. Smith urug No.
W 00016 A. O. Smith Sorial No.
W 60016-1 Rocctor Tiestin;; house Eloetr$c Corpo A.P.D.-5hR-8026 U tEEr~ ~~
We'.dar Clia~u~ of 7 te Qualify.nn
~
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I Nt.t A'.ca T llo Qualification Qualified Nat'l Board Ctrn !ie.
Inspectcr h!/2
'Diaman Lailnr Fh 1-9-59 C. Plum ifu1 i
hV' Hobart Krug Fh 1%15-59 C. Plua
'tfdl SN I Sigfred Inndtplist rh bl%56 C. Plcn 16n
!40 Sigfred Lundquist F5 ll-19-5h C. Plun 1%~1 537 George Ploo '
Fh S-D-53 C. Plum 3671 5*;"
Gocrrge Plotr' F5 7-23-56 C. Fa um m il Fh0 Thomas Hobro F5 1-28-59 C. nus 3611
$h9 RusseD Cete tney Fh h-23-58 C. Ilum j671 I
Sh9 Russen Cwtrwy F5 10-2-58 C. Plum l
.16 71 5FP drnest Kellne Ph S-28-56 i
C. Plum J67J 560 laroy rober Fh 2-20-59 i
C. Plum 36n
$60 Leroy reber F5 2-2h-58 i
C. Pluu 3671 591 aussel steldt Fh 10-5-54 C, Plum 3671 591 Russel Stelitt F5 9-h-57 C. Plum 3671 5P7 John Knpas Fh 9-U-56 C. Plun 3671 5?7 John Knpas F5 9-3-57 C. Plum 3671 519 3dnin skowron Ph 7-27-56 C. Plum M71 559 ddwin intowron F5 7-28-56 C. Plum j
3671 1
6*d.
2ef or 08Kimosh Fh 8-16-56 C. Plum M 71 6%
. Peter O'!{inonh F5 1
11-h-57 C. Plum
% 73 6;i George Peterson Fh i
3--22-56 C Plum 3671 6;C.
George Peterson F5 i
3-28-60 C. Mum 1671 g
6:1 George Peterson Q21b 7-3-57 C. Mun 3671 1
6.*5 Andrew Brodaeller Ph 1-15-59 C. Plum.
3671 6.'$
Andrew Brodaeller F5 t
7-12-56 C. Plum M71 6Hi
.Jamoe t. doll Fh l
7-20 C. Mum
$71 6t'h James liden F5 7-2-57 C. Plun 36n l
- nf; J. D. clark Ph 9-30-5h E.E. Fritz 1095
??.0 J. D. Clark F5 B-.18-56 C. Hun 3671
- '. o scrl Oltman Fh I
D-29-57 C. Plum 36'il
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"; o Ralph Limmons Ph l
l-9-59 C. Plum
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'J F o Rciph himmous F5 1-15-59 C. Plvm
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A. O. SMllH COPP.
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ATOMIC & 080( 6p P.LIOFI l b.
3 CQUlFMfjli DIVl%!4 LIST 0/ ni.:.D4n QUALIf1d4 nOL FAG
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1 it tW/ Ut f E. WI$00N$lti l
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A. O. Smith Derg; No.
MV 80016 A. O. Smith Serial No. M7 800 M-1 Reactor Tiestin ; bouse Electric Corp. A.P.D.-$hR-8026 Tiolder Y! alder Class of Date Qualifying ifat l~7 isaac i
No Qtuli fica tion Qualified Nat'l Doard Ccm. i?o.
Inspector 787 Edgar Bowera Fh 9-29-5h E.R. Frita 1 tin 787 Edgar Bowero F5 8-29-57 C. Plum 3671 802 Samuel Jackson Th 8-1955 C. Plum 36R 806 Densel Woodward Th 12 55 C. Plum 3671 82h Joseph Stoltz Ph 6-20-60 C. Flum 3671 825
'liillie Adama a
h-22-57 C. Plus 3671 825 Willis Adams F5 2-31-58 C, Pl=
36n 8h3 Inster Dagmr Ph 7-27-56 C. Plum 3671 8h3 lester Dogmr P5 3-22-60 C. Pim 36n 81 0 Imster Degmr Q21b 7-1557 C. Plum 36R 85h William Sterard Ph 5-15-57 C. Plum 3671 861 Ralph 'l11 son Ph 5-22-57 C. Plum 3671 l
895 Darrin S,joborg Ph h-26 60 C. Flum 36n 92h Leo Dockstader Ph 3-20-58 C. Plum 36R 928 Cleatua Kennor Ph 2-1&-58 C. Plum 36n
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928 Cluatus Kenner F5 10-2-58 C. Pitan 36n We certified that all of the above listed welders have boon qualified in accordance with Section H of-the ASMS Code and that all testa have been witnessed and approved by the above listed National Board Comunissioned Inspector.
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l-46hn'L. Iverson i
fiaryager, Quality Control hginserin; Atomic and Process Equipnoot Diviuion u
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VCAP-1391 (Bevised Edition)
M uch 1, 1960 MJLTI. LAYER CONSTRUCTION IDH THE SAXTON REACTOR VESSEL Prepared Dy:
L. P. Katt February 1, 1960 Approved: 8//h[(S/d/
7 E. A. Goldomith, Sectier. Manager Primary Systems Section 1
WESTINGHOUSE ELECTRIC CORPORATION ATOMIC POWER -DEPART &TT PITTSBURGH 30, PENNSYLVANIA
A0310VLtr#Iffra The author wishes to acknowledge the invaluable contri.
bution made by Mr. J. J. Maurin of the A. O. Smith Corporation in the preparation of thin report. The nu' hor also viches to thank Mr. A. C. Martin of Westinghouse Atomic power Department for hte help in review and interpretation of the analytteal data supplied by A. O. Smith Corporation.
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TAllLE OF CONTENTS Pace No.
ACKNOWLED3ME!CS....................................................
11 LIST OP PIGURES........................
4...........................
Y
\\
1 vaRODucI10N.................................................
1 II ADVANTAGES IN ETLTI LAYER CONSTRUCTION....................... 2 III - PROBLEMS IN MULTI LAYER CONSTRUCTION.........................
le l
A.
Thermal Streusec in R11ti-Layer Shelle...................
1 B.
Thermal Stresses j n the Saxton Vessel....................
(,
C.
Deviation From the ASME Code.............................
7 D.
Pennsylvania "Special" Approval of the Saxton Vescel..... 8 IV
SUMMARY
9 APPENDIX I APPLICATION OF MULTI-LAYER CONSTRUCTION TO THE SAXTON REACTOR VESSEL......................................
11 A.
Technical Description of the Saxton Vessel.............. 11 l
l B.
Manuf acturing Procedures for the Saxton Vessel.......... 13 C.
Inspection Techniques for the Saxton Vessel.............
15 D.
I Weld Designs in the Saxton Vessel....................... 16 APPENDIX II - HISTORY OP KTLTI-LAYER CONSTRUCTION.................. 1 A.
Current Osmmercial Non-Nuclear Applications............. 17 t
B.
Curre nt Nu cle ar Appli c at i on s............................ 18 C.
l Expected Future Appli c at1ono............................ 20 t
I i
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LIST or F10Ulms Ficure No.
Title 1
Thermal Conductivity in Stacked Disks 2
Thermal Conductivity in a Malti-layer Shell 3
Saxton Reactor Veccel Outline 4
Effect of Onama llenting Rate on Maximum Tensile Strenc in Wa.11 of Saxton llook-on Reactor Multi-layer Vessel - Cylinder "A" Precompression Effect of Temperature Difference Across Wall on Maximum Tensile Stress in Wall of Saxton }{ook-on Reactor Malti-layer Vennel. Cylinder "A" Pre-compression 6
Effect of Gamma }! eating Rate on Maximum Tensile Stress in Vall of Saxton liook-on Reactor Malti-layer Vessel - Zero Precompression 7
Effect of Temperature Difference Aeroes Wall on Maximum Tensile Stress in Wall of Saxton llook-on Reactor Malti-La,rer Vescel - Zero Precompression
. iv.
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1 I firiu >t e rlotl The Satxton Reactor lo an experimental fric111ty to attain lo.ter nuclear power cocts by conducting experimento which vill lead to h16her thermal efficienelen rtnd lower capital costo. This reactor is designed to produce 26 thermal megavutto and vill be hooked on to an exit: ting turbine. generator which in part of the S!txton Stew Generating Station of the pennsylvania F.lcotric Company at Saxton, Pennsylvani a. One of the cpeeled featuren to be incorpora.ited int o the Saxton plant is a renctor veonel of multi-layer construction, a type which prominec cignificant advantages and capital cost sav-ings for large size plants.
Malti-layer preccure vesselc, which are a specialty of the A. O.
Smith Corporation of Milwaukee, Winconnin, differ from conventional reactor vescels only in the method of fabrication of the main cylin-drical chell course. The bottom head, removable clocure head and main nozzlec are identical to thone uned in conventional reactor vesselc. The main shell of a multi-layer veccel in comprised of many layerc of relatively thin plate, each septtrately formed into a barrel, vrapped and velded one to another until the required total thickness is at tained. The inner barrel of the multi-layer shell is made from a plate, approximately 1/2 inch thick, which han been clad with a thin layer of stainleon steel prior to forming. After forming and veld!ng, the inner barrel is ctreoc-relieved and the veld seam ic completely X-rayed in accordance with the ASME Code. Subsequent layers, which are 1/4 inch preformed plates, are applied one at a time in three equal segments.
Each layer is velded intermittently along the three longitudinal ceams to the layer belov vhile the assembly is mounted in a special vrapping machine. After removal from the vrapping machine, the longitudinal layer seams are completed, ground fluch with the outside diameter of the layer and are magnetic-ally inspected. The next layer is then applied. The vrapping and velding process employed cruses a pre-stressing in each layer result-ing in a net compreceive streso in the inner layers of the completed shell. The completed multi-layer shell in bored for nozzle insertion and scarfed at each end for velding to the bottom head and bolting flange in the same manner as for solid vall vessels.
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II ADVANTAr.ES IN MU!41-LAYl:h COTTTBUPTION The uce of multi-layer conc',ruction for reactor veccels offers several dictinct adyttntages over solid vall conctruction with respect to safety, overall econory and manufacturing flexibility.
The advantage in multi-layer const ruction are ao follows :
1.
The relat ively thin plateu uced in mult i-layer construction exhibit better metallure,Jeal properties than thlet platen for the same application. Thln pinteo can be manufactured without the danger of loclunionn and risvn whleh often occur in thich platen.
14etter lov temperature impact propertico recult because a more un! form quench after heat treatment la posolble.
2.
The precompression built into the inner layern of 4 multi-layer chell by the combination of vro}> ping load hnd veld shrinkage techniques employed in t he manufacture of multi.
layer chella resulic in a favorable ntrecc dist ribution through the vall under j nt ernal preanure.
In destructive pressure tents on full cize multi-layer vecceln by A. O.
Smith, 100 per cent of the calculated strengths were developed in the chelin before failure. The multi-3ayer shell test failures were in all cacen of t he duct 13e-type, without shattering or fragment ation. Similar tests conducted on solid wall vesseln by A. O. Smit h choved that the chdle developed only approxjmstely 70 to 80 per cent of th 711 calculated strengths before fuljure. Likewise, frapc. 4 tion at failure van found to be typical of solid vall vessels.
3 In multi-layer construction, only the inner clad barrel serveo as the fluid cont aining port ion of the vencel, and therefore is-the only layer of the multi-3ayer chell vnich must be leak-tight.
11y means of a c1mple vent hole running from the outside of the inner barre] through the outer layer vraps, any fluid which leaks from the inner barrel can be effectively monitored. Thun, any defect in the -inner barrel is detectible at a time when the vecnel itself is still structura]Iy cound. A camilar defect in a solid val] vesse3 would not be detectible until a complete structural failure of the vessel vould occur. Thic. feature in especially impor-tant to multi-layer nuclear vence]c since irradiation damnge which may be imparted to the ve esel valls wovld most likely occur in the inner shell only.
14 Multi-layer construction in reactor vessels offerc definite manufacturing flexibility an to overall size and shell thick-ness an compared to atandard solid val) construction. Shell
( d
,mm___
_m_._____._m_
thickness, limited by roll nize and roll enprwity in colid vtal construction is conentia]Iy unlimited in multi layer constr' action since the chell is built up of many layer vrais of relatively thin material.
Also in future large vessels which present a handling and trancportation problem because of weight and size limitations, multi layer construction pro.
vides a solution.
In multi-layer construction, the " thin plate" metallurgy of the layer vrap material precludes the necessity for strece relieving the cettpleted vessel.
- Thus, it to possible to ship multi. layer reactor veccels to the site in sections and }erform a field veld of the sections without the complications involved in colid vall construction.
The welds connecting the multi-layer shell to the bottom head and flange for61ngs are relatively free from veld shrinkage stressec and do not require atreso relief. The elimination of shrinkage stressen_in thid veld resulto from-the fact that the individual layer platen tee able to expand longitudinally, thus, stress relieving to un nececcary.
1 3
f 9
7_
III TRCELEMD IN MJLTI-lMEh r0!MIRUCTIOf[
The une of raulti.Inyer construction for reactor vesselc imp:icen several important } roblems wh!ch kre not prevalent in colid vall vessels. These problemn are related to the heat trannfer tend related thermal otresces in multi-layer shells and ASME Code a;-
proval of vessels using this type of construction.
A.
Thermal Stresses in Hulti-Layer Shells The neutron and gamma bombardmer.t of a multi-layer reactor vessel causes a temperature gradient in the shell. This temperature gradient cauren a higher temperature at the outer layer vraps and a lover temperature at the inner vraps, beenuse of the inner vall cooling, resulting in a reduction in the interface }ressure between layers. The reduction in interface pressure between layers reduces the effective over-all thermal conductivity of the shell and thun increncen the thermal stresces.
Several years ago when the need for multi-layer effective thermal conductivity data for reactor vecnels became upparent, A. O. Smith conducted a series of testo to determine this parameter.
A laboratory device van designed and built for the determina-tion of the thermal conductivity of multi-stacked discs for various contact pressures. The attached Figure No. 1 gives the thermal conductivity results for contact pressures between dises up to 2000 psi. Withthesteelintheasreceivedcondg-tion, the thermal conductivity varies from k.0 Bt or x ft x F at zero contact pressure to 11 3 at 2000 pai, picti ng the layer steel improves the thermal conductivity to 4.? nt zero contact pressure and 14 3 at 2000 pai. The thermal conductivity of solid carbon steel is approximately 26. pickled steel is used in multi-layer vessels for nuclear applications where thermal conductivity is important.
A simple test was also used to determine the thermal conductivity of a multi-laler shell made from pickled steel. Heat was trancter-red through a k in. thich muJti-layer shell from hot oil inside to cold water outside with both fluida at atmospheric pressure. The thermal conductivity of the vall was calculated from the measured temperature drop through the vall and the amount of heat transfer-red. The attached Figure No. 2 gives the experimental results, the thermal conduegivity of the multi-layer shell varies from 6.0 Btu /hr x ft x F to 1k.5 as the temperature drop through the vall increased. This curve illustrates several points.
1.
The thermal conductivity of the multi-layer shell is approx 1-mately the same as the thermal conductivity of multi-stacked dises. The comparison is for pickled steel.
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2.
Since multi-layer nhella etre renufactured wit h conciderable precompression, there it hich cohtact } reocure between lay.
ers of the manufactured chell. The initial thermal cchduc.
tivity of multi-layer starts at 6.0 versus 4.0 for multi-ctacked discs at zero contact pressure choving the effect of this precompression.
When the resistance to heat conduction between layers of the above mentioned multi layer shell is uced to calculate an air gap between layers with an equivalent resistance, the calculated air gap is very small and averades.00008 in. The average gap vas eniculated from the measured initial value of thermal conductivity (8.0 Btu /hr x ft x F).
The Franklin Inctitute in conjunction with A. O. Smith has made nn analytical study of the multi-Inyer dec1gn for precoure and thermal stressen found in nueient reactor veccelo, in setting up equationc for multi-layer, it was necessary to take into consideration pre-comprescion during manufacture, internal operating preocure and the terperature distr $bution in the vall. The temperature distribution is affected by the amount and type of internal heat generation and the thermal conductivity of the vall. These equations have been programmed on the A. O. Smith IlW 704 computer to acive multi-layer vall problems for steady state conditions, j
Vherever possible, the theoretical assumptions for the analyses are realistic or on the conservative side. The value of thermal condue-tivity for multi-layer used in these equations is the lover line on Figure No.1 for the "as received" material even though pickled steel is used for layer conctruction. The amount of precompression used in the equations can be varied but a conservative ve,1ue called " Cylinder A" precompression is generally used.
" Cylinder A" precompression represents values of precompression determined by actual strain gage measurements on the inner shell by A. O. Smith during the shell layer vrapping process on multi-layer shells. These values of precompres-sion can be built into a completed multi-layer shell by close control during the vrapping process.
Two multi-layer reactor vessels, SPERT III and IUTF (non-nuclear) previously ranufactured by A. O. Smith have been in operation since early 1960.
A. O. Smith has conducted a cooldown test on the lifTF simulating the temperature gradient in the shell similar to that which will be expected in a reactor application. The experimental data from IITTF 1r presently being analyzed by A. O. Smith. This data vill serve as a positive check on the accuracy of the analytical
' methods employed by A. O. Smith in the design of the Saxton vessel.
Similar tests on the SPERT III vessel vill also serve as a check on analytical methods employed by A. O. Smith. Details of the design, operating conditions, and test results expected on both IUTF and SPERT III are given in APPENDIX II - III!Tf0RY OF MULTI-IAYER CONSTRUC.
TION. %v
=_
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B.
Thermal St recren in t he SW on Ventel The resultc of the upplient mn ut t hese cent ral eput ions to the multi-layer veccel for Saxt on lor bot h 20 Ms and pot.rible 40 Ms operation of the Saxton plemt are 2ndicat ed ca Figurec k, 5, 6 and 7.
The ASME Code in itc ruling on Cat e 1273N, ctat er that the combination of preocure and Ihermal ntrencec at any level of nteady power orerotion shall not exceed 1 1/2 times the allow.
able code design streco. The allowable code design ctrecc for layer steel (A. O. Smith VMS W1150 Special Grade A) le 16,750 psi at 6LO F design temperature. Therefore, t he tot ta shell streopen chould not exceed 20,125 poi during the ute4tdy ntate power operation. The calculated nt rearco for the multi layer chell of the Saxton reaet.or vercel are within code limit s.
The maximum Jnternal heat general.lon at the inner nurfare of the veccel vull is 15,600 Blu/hr x rt i for 26 Ms (thermal) power operation in Sar. ton. The linear absorption coef ficient -
for ateel in 0 50/jn. Vainn t hene valuco of internal heat generat ion at the denign preneure of 2900 pel for preosurized water operation, the maximum nt recc in the vence] vall JB 19,500 poi and the calculut ed t emperature drop throuch t he vall la 260F no chovn on Figuren 4 and 5 Figures b and 5 alco in-dicate that for future operation at ho MJ (thermet1)3the maximum internal heat generation vil] be 20,300 Btu /hr x f t The mroximum streno in the vessel vall la t hen 20,$00 pei and the calculated 4
temperature drop through the vall is 3kOF. These stresses are well below the allowable of 26,125 pot. It vould be necessary to increase the gamma heating rate to 5k,500 Btu /hr x ft3 to reach the allovable stress of 28,125--then, the temperature drop would be 95 5 F.
Therefore, there is a safety factor of 3 4 at steady state on the analysis used for the multi-layer design at 26 Ma (thermal). The allovable cooling rate vill be set to permit an appreciable safety factor even during trancient, conditions.
If, for some unknown reanon, the multi-layer vall chuuld lose its precompression, the vesse] vould still operate natisfactorily.
It vould be nececogry t o inerence the gamma heating rate to i
52,200 Dtu/hr x ft to reach the al. lovable stress of 26,125...
then the temperature drop would be 970F.-as shown on Figures 6 and 7 The effect of local gaps on the temperature distribution in the Saxton reactor multi-layer vennel vas also eniculated and was found to have no detrimental effect. Extreme or limiting ascump-tions were made to simpiiry the calculations and obtain the worst possible temperature drop acrono ihe gap.
I
. (,.
A complete circun.ferent lui enp ol' 00P in. between the inner chell guid t he firnt layer rerui t.a in a t erg erature drop of lb.bor herocs the gap. Thic ic ent uhted at t he et ecified maximum heating rate of 26 W (thermal) for the Saxton reactor.
The corresponding t emperature drap acroro x.002 in, car between the two outermost layers is.2o F.
An areu 6 in. In diameter tu necumed to have a cup to thick that no heat is tra.norerred aerona the r/sp but munt flow t o the edges of the loane area. Thin in an impornible and t herefore a con-servati ve accumpt ion.
For auch a gap between the inner chell nnd the first layer, the vulculated temperat ure rice at the mid-paint of the first layer above i he cup 10 only 7.tFF. A nimilar euleu-lation for uurb h enp bet ween ' he out ermart 3 ayer given a tempern-ture rise of only 0.0"F.
The aren of an annumed et reulur cup belwee n t he inner ahell and the firal layer wrap under i nt erna t prennure Is limit ed by the flex)bility of the luner ahell.
Ihr t he Saxt.on upplichtlon, under 37$0 pol int erned prenuure ut. hydro. t ent, the 1/2 in, thich Inner chell would yield tu4d clone any built in cap whlnh 10 over le i n.-
in diameter. Thic ertimale le bnced upon the concervative accump.
tion that the Inner chell nbove the built-in gap behaven ac a circular plate with cJumped edrea.
C.
Deviation From the ASME Cnde The techniquec employed in mult1-1syer construction impose two specific deviations on the rules and requiremente of either Section I or Section VIII of the ASME Code These are the inability to interpret radiographs of the elrth seam velds, and eitmination of final stress relief of the finished vessel. Since the definitive radiographic examinstion of girt h seam velds for Sect ion I or Section vin Code approval is not possible because of the influence of the reun between layers in a multi-layer shell, A. O. Smjth is currently developing an ultranonic technique. Thic developmer.t is currently in progress on t he N'fV mult). layer vessel being msinufac-tured for Westinghouse Det t ic.
In case the ultrasonic inspection technique for girth seam welde does not prove satisfactory, radio-graphic examination is pocalbie using a betatron.
Alt hough A. O.
Smith does not have a betat ron at this time, use of the Allis Chalmers betatron has been arranged for the Saxton vessel if required. The ultrasonic technique for inspecting the nozzle to multi-layer shell velde has been previously perfect ed by A. O. Smith, and has been found to be more accurate than radio-graphy.
The final stress relief of all precsure velds required in Section I and Section VIII of the ASME Code, is purposely eliminated in multi-layer construction. This stress relief would tend to remove the desiratle precompressive stresses built into the shell layers, with
(
7 A
the resulting loss of interface pressure between layers which reducen the overall thermal conductivity of the chell. The elimination of final utrenn relief on the mult1~1tAyer shell to forging velds doec not, however, compromise the narety of multi-layer construction. Ac the veld paccec are applied in the veld groove, each layer plate is able to expand longitudinally, re-sulting in a veld which 10 free of built-in stressec. Strecc relief, therefore, is not required.
All componente in the Saxton vescel, including the multi-layer chell have been designed for site and thickness in strict adherence to Section I of the ASME Code nacuming no credit for the pre-stressed condition. All componente in the Saxton vessel except the multi-layer chell vill also be manufnetured and inspected in accordance with Section I of the Code.
D.
Penncylvania "Special" Approval of the Saxton Vessel The description of the Saxton vencel, includinc design calcu-1htions has been transmitted to the Pennsylvania Bureau of Inspection, Department of Labor and Industry by A. O. Smith Lorporation for approval for operation in the State of Pennsylvania.
State approval in the form of a "special" prescure vessel flk69 han since been issued for the Saxton vessel after a detailed review of proposed fabrication and inspection proceduren. The design of the Saxton vessel has also been reviewed and approved by the Saxton Safety Committee.
This committee has responcibility for review of all Saxton systems and components with respect to overall plant safety.
4 a
]
l i
IV
SUMMARY
A comprehensive review of the engineering analysis, fabrication methods and quality control techniques used by A. O. Smith Corp.
in the manufacture of multi-layer vessels provides adequate assurance that this type of conctruction is both safe and reliable for the Saxton application. The following listing summarizes the conclusions on the safety of multi-layer construction for Saxton.
1.
The use of thin plateo in the manufacture of the ru2 ti+1ayer shelle insures metallurgical soundness and improv4 1mpact properties of the shell material an compared to tuja platen for the same application.
2.
An independent analysis by Wectinghouse APD of the streccec due to gamma heating and internal preenure in the Saxton mult.1-layer reactor vessel indleatea safe operation at both 28 Mw and k0 Mw operation. This analysis given in WCAP-1$06 " Investigation of Service Stresses in Multi-Layer Nuclear Reactor Vessels", pre-sents results which compare favorably with the original analysis c;nducted by Franklin Inctitute in conjunction with the A. O. Smith Corporation.
3 The determination of the heat transfer through multi-layer shello, used both by A. O. Smith Corporation and Westinghouse APD in the stress analysis of the Saxton vessel are based upon reliable laboratory experiments conducted by the A. O. Smith Corporation.
4.
All deviations from the ASME Code required in the manufacture of multi-layer vessels are substituted by equally reliable alternate techniques and procedures to assure that the safety, integrity, and quality control are preserved in the final product.
A complete listing of the manufacturing procedures and non-destructive testing techniques to be employed on the Saxton vessel are given in APPENDIX I - APPLICATION OF MULTI-LAYER CONSTRUCTION TO THE SAXTON REACTOR VESSEL.
Malti-layer vessels for commercial, non-nuclear applications have been used safely in high pressure service for many years. A list-ing of typical currently operating multi-layer vessels in the same temperature and pressure range as Saxton is givcn in APPENDIX II -
HISTORY OF WLTI-LAYER CONSTRUCTION.
l 6.
The detailed design and fabrication techniques proposed for the Saxton reactor vessel has been approved by the Bureau of Inspection of the Department of Labor and Industry, State of Pennsylvania.
This approval was confirmed by the issuance of a "Speelal" Penn-sylvania pressure vessel number to be stamped on the completed Oaxton vessel. The vessel design has also received the approval of the Saxton Safety Committee.
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7 Operating data from the ITPfF wid SIERT III multi. layer nuclear reactor vessels, previous 2y maraufactured by A. O. Smith Corp.,
have and/or vill provide a confirmation of the analytierd deter.
mination of thermal gradienta and thermal r,trer.ses in mult1-layer vessels.
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APPDiDIX 1
- APPL 1 RATION OF MJLTI-LAYFR CONF 1Tf:UC'rION TO THE DAXTON REArrok VLSSEL A.
Technical Description of Saxton Vescel 1.
Punctional Specification The Saxton vessel shown on Figure No. 3 to a nuelcar reactor of the combined pressurized water and boiling water ty;e.
During normetl pressurized water operatlon, the reactor will Iroduce 28 Mw (thermal).
At a Inter date, a redenighed cure may be added to produce ho M.i (thermaj).
Denign Specificat ton Pressure 2$00 poig 0
Temperature 650 F llydrostatic Test Precoure 37$0 poig Heating and Cooling Hate 2000F/hr (denign objective)
Internd Diameter
$6 in.
Malti-Layer Vall Thicknesc
$ in.
Incide Vessel Length 15 ft. 7 1/2 in.
2.
Materiale of Conctruction Inner Shell AITTM A-264 Orade 3, clad plate with ASTM A 212 B FBQ backing material and 1/8 fr.. nominal, Type 304 stainless steel cladding.
Shell Layers ASTM A-212 B modified per Code Case 1056.$ which is equivalent to A. 0. Smith VMS W13$o speelal crade A material.
Physicals Tencile Strength, Min. 75,000 poi Yield Point, Min, 41,250 poi Elongation in 2 in., Min. 26.0%
Chemistry Carbon 0.22 to 0 30f Manganese 0.6$ to 1.1$%
Phosphorus 0.045% Max.
Sulphur 0.0$0% Max.
Silicon 0.10 % Max.
( 5)
~ _ _ _
Hendo A!IrM **.-PIPh H Q plhte "1 a.1 w ' t b i y] e sob L; wejd delsinit etlldled by'M rut rwrgul au torat i c 'T. I n Are" rnt i tnd.
j Depocj led elhdding to hhve h PkO MG surfhee finish.
Forced Planges ASTM A-350 Grude LF-1 whleh is equive-and Nozzles lent to A. O. Smith VMS $O00.
Physichl0 1encile Strength, Min, t;$,000 1 ni YJeld Point, Min.
33,000 1 01 Elongation in 2 in., Min. 23 01 Beduction of area, Min.
3 4.0'f Cheml nt ry Carbon 0.21 to 0.261 Mangunene 0.61 to 0 901 phocphorus 0.0W Mu.
Sulphur 0.0$% Max.
Silicon 0.1$ to 0 301 All primary fluid enntact surfaces to by Type 30f: L deposit clad and have a 250 RMS finish.
Studo and Nuts AISI-43ko to ASTM A-193 Threads to be treated with magnesium phosphate to prevent galling.
Bolting material shall have Charpy V notch im-pact value of 35 ft. Ibs at + 10 F.
Control Rod Parts ASTM A-2kO, Type 304, stainless cteel Access Ports ASTM A-240, Type 304, ctainlean eteel Thermal Shields ASTM A-2ko, m e 304, atainless cteel Support Skirt ASTM A-212 B FDQ carbon steel Insulation Canning ASTM A-283 D or equal Closure Oasket Flexitallic, Type 304, stainless steel NOTE: All pressure containing carbon steel plate and forging material including the support skirt shall have a Charry V notch impact value of 15 ft lbs. at + 10 F.
Material that is too thin for sta%rd size Charpy specimens vill be tested using subsize Charpy specimens.
k
- I F. -
s,
3 Nutzlcu The chell haa two 10 J/P in, 1 D.
Inlet,and tv: 10 1/P in. 1.D.
outlet connentionn attached to.he vull with ful1 penetration velon. Each connection ic integ: ally reinforced.
4.
Head Penetrations The reactor has twenty head pesietrat ions, nine 21/2 jn. I.D.
control rod mechanism adaptors in tha bottom head and eleven in the top head. Of the eleven acceup and instrumentation connections in the top head, cix are 3 in. I.D.,
and five
.e 2 in. I.D.
All reinforcement for thene connections are calculated to be in the exceso head thickneco. The ports for the control rod drive mechanism are capable of withstanding of momentary im-pact load of 2000 lbs, per rod due to dropping of the control rods.
Due to the difference in expancion of the head penetration connections and the het.d material, the connections tire velded on the inside face of the head only and are allowed to expand along the axis of the connection.
5 Design Criteria The vessel is to be constructed in accordance with A. O. Smith High Pressure multi-J ayer specification MLS-30A, Rules and Regulations of the Pennsylvania Department of Labor and Industry and vill receive a Pennsylvania "Special" number and stamping.
Design is in accordance with the ASHI Power Boiler Code.Section I, where applicable.
Design stresses reculting from a combination of thermal and pres-sure stressen during steady ctate operation chall be in accordance with ASME Code Case 1273N and chall be limited to 1 1/2 timen the ASME Code allowable value.
B.
Manufacturing Procedure For Saxton Veacel 1.
Clean 11neco of Layer Platen a.
Layer plates to be pickled and alkaline rinced on a bacio of two layerc at a time. Care to be exercised by covering the layer plates to protect them from grease, dirt and other foreign material.
b.
Clean the inner shell of all grease, dirt and foreign material before commencing the vrapping operation.
t' 13 -
l
2.
Fabriention or Layer Chell a.
Install layer plate on inner shell in vru] ping machine.
b.
Veld layer long seams in secordance v!th veld procedure, c.
Contour grind the long seams to template, d.
Hammer test layer for tightnesc.
e.
Proceed with the balance of the vrapping operation under the same procedure as above, f.
Scarf ends of shell section for circular seam velds.
Etch inner surface of veld cearf to be sure all alloy to removed from carbon oteel velding curface.
g.
Layout shell for connection openings.
All nozzle open-ingo to be laid out for machining to final diameter.
h.
Inspect for veccel uanembly.
3 Strecc Relieving Stresc relieving of the clad inner chcIl shall be performed prior to vra} ping the multi-layer shell. The completed re-actor vessel vill not be strees relieved.
Heads, nozzles and flanges shall be stress relieved before they are attached to the multi-layer shell.
k.
Welding A11 velding vill be performed in necordance with procedures
- Jr
.y veldera qualified in accordance with Section IX of t?t A ME Code.
5 Surr6'l, Finish All surfaces in contact with the process fluid shall have a 250 RMS surface finich or better.
6.
Tolerances The vessel tolerances shall be in accordance with A. O. Smith i
Draving AES-Q573, Revision 2.
Tolerances that are not specifically stated shall be ir. accor-dance with the ASME Code.
- ik -
C.
Inspection Techniques For Saxton Vessel 1.
Radiography The following veld nenmo nhull l>t- :udjographed in accordance with Section I of the ASMP. Code for neceptance or rejection, n.
The longitudinal neam of the clad inner nhell before application of the layers and after alloy depocit, b.
The circumferential seam in the top spherical head.
c.
The circumferential seams attaching the stainless steel nozzle otub ends to the ctainless steel lined connections.
d.
The vessel circular ceams joining the 11.ar barrel to the bottom head and top flange forging.
2.
Mngnetic particle Incpection A.C. ma6netic particle inspection shall be performed on the following --
The longitudinal seam velds of each applied layer plate after contour grinding.
l 3
Ultraconic Inspection a.
All forging material vill be ultrasonically tested for information purposes only.
b.
All integrally clad plate vill be subjected to ultra-sonic inspection prior to rolling or forming in order to determine the degree ot' bond. Ultraconic tecting is for information purposen only and major defects vill be repaired prior to rolling and minor defetcs vill be rechecked after rolling.
c.
Layer to nolid vall seamo (heado and flange f,,rgingo) chall be ultrasonically tested at full thickness in accordance with a technique that A. O. Smith to st present developing. In case the ultrasonic techn.,ue for layer to solid vall veld seams does not prove a,.' tis-factory, radiographic inspections vill be done on thece velds using a betatron.
d.
All nozzle attachment velds in the multi-layer chell sectior shall be inspected ultrasonically by the immer-sion method develcped by A. O. Smith.
I
- 15
____________._____________.m__._...._______
k.
Dye Penetrant Innpoetion Dye penetrant examination shall be },erformed on the following areas:
Stainless steel deposited surfaces after the last layer a.
of veld deposit. This includer, stainless overlay of carbon steel velds.
b.
All gasket surfaces after final machining.
liydroctatic Testing S,
Th? completed reactor including head chall be given a hydro-static test of 37k0 psig for a duration of two hours.
D.
Weld Designs in the Saxton Vescel 1.
All longitudinal and circular ceamo on the velded inner nhell vill be of the double butt velded type.
2.
All longitudinal and circular cerut, on the multi-layer chell vill be manual are velded to A. O. Smith production engineer-ing design standards.
3 All nozzle attachment velds in the multi-layer shell vill conform to A. O. Smith production engineering design standards.
4.
All other velds vill be in accordance with ASME Code practice.
P C
4
%u (
ArfENDIX II HISTORY OF MULTI-LAYFH CONSTRUCTION A.
Current Commercial Non-Fu: lear Applications The use of multi-layer shells for pressure vessels es.ra ahjut through A. O. Smith's efforts to find a safer and more economic construction that vould be particularly effective where thick valls were invo]ved.
The early development work van carried out in 1930 and a patent vac granted Mr. R. Stresau, the inventor, in 1933 Many patenu have been granted since that time and some are still in effect.
In the same year, a thick walled multj-layer vessel was built for testing to destruction--the first of many full scale vessels built and tested to destruction in the most comprehencive and thorough research progra:n of its kind ever undertaken. As well ac deriving a vast amount of design information from these tests and ver ifica-tion in act'ial operation, multi-layer vessels draw heavily on the accumulated knowledge of materials and methods of fabrication gained by the A. O. Smith CorporatJon from building come 8,000 multi-layer petroleum and chemical vessels up to 35,000 psi design pressure of every variety and description.
The following listing ic typical cf the non-nuclear multi-layer vessels manufactured by A. O. C'nith Corporation which are currently in service.
Included also is the customer, service, approximate size and operating temperature and pressure for each installation.
ID Press.
Temp.
When Customer Service (in)
(psi)
(OF)_
Instedled Hydro Press, Inc.
Accumulator h2 2850 So 1954 Imperial Cheia. Co.
Converter 59 h800
-650 1940 Calco Chem. Co.
Autoclave 46 2000 462 1941 Atmospheric Nitrogen Company Converter k5 3150 650 19h2 Rohm and Haas Co.
Reactor 36 5000 650 1950 Solvay Process Co.
Converter h5 3150 650 1951 Spencer Chem. Co.
Converter 36 h000 675 1953 M. W. Kellog Co.
Converter 50 5000 475 1955
- Borreguard Co.
Reactor 71 2300 665 1956 Escambia Bay Chem.
Company Converter h5 5/16 5200 482 1957 Texas Eastman Co.
Rt. actor 70
-2000 630 1958 l
o' s
(
The demands for multi-laye vescels in recent years have teer.
very great and the des 1 n flexibility of layer construction has 6
lent itself ade.irably to the solution of specific vessel designs for many purposes. The designs of such vessels has met the re-quirements of safety in accordance with regulations of Federal and State Laws, U.S. Armed Services, Lloyds Register of Shipping (Board of Trade Requirements) and other vorld-wide regulatory bodies.
B.
Current Nuclear Applications To date, the following multi-layer vecsels for nuclear applica-tions have been cupplied by the A. O. Smith Corporation:
1.
High Temperature Ter. Facility (H.T.T.F. )
5festinghouse-Bettic Pi t_t c burgh, Pennnylvang The H.T.T.F. Reactor is designed for operation at 650 psi and SOOOF, with no gamma heating.
It has been in operation since November 1)S9 The vescel is 82 in. 1.D. With a con-siderable number of shell openings in the vessel vall. Using multi-layer construction, the valls are 21/2 in. thick, made up of seven 1/4 in, thick layers with a 3/4 in, thick inner shell. Layer material is A. O. Smith VM5 W1350 Speelal crade A, heving an allovable stress of 18,750 pai at Soc F, and the o
inne 6 hell is clad with 1/8 in, thick, Type 304, stainless steel.
The top hemispherical head is 2 1/2 in, thick, ASTM A-212 B FBQ material clad with 1/8 in, minimum thick, Type 308 L, stainless steel deposited by the automatic submerged " Twin-Are" method. The bottom ellipsoidal head is also 21/2 in.
thick of the same material and cladding.
The design specificatlog calls for 2000 cooling cyclec, each cycle from 472 F to 200 F in two hours (136 F/hr) which is a moderately severe cooling rate. However, with the strain gage readings, it may be safe to exceed this cool-down rate, thereby improving the veasel periormance and increasing the rande of strain gage information.
2.
Special Power Excursion Reactor Test (SPERT III)
National Reactor Testing Station l
Phillips Petroleum Company, Idaho The SPERT III Reactor is a pressurized water reactor desi ned 6
for2500psiat700Fmataltemperature,vitg)expectedgamma 0
heating of.56 vatts/cm3 (54,100 Btu /hr x ft
. This vessel has been in operation since late 1958.
Using multi-layer construction, the vessel I.D. is L8 in, and f
shell thickness is 31/4 in (eleven 1/4 in, thick layers) with a 1/2 in. shell clad with 1/8 in. thick, Type 304 ELC i 1
stainless steel. The shell material ic A. O. Smith VMS 1146 having an allovable strena of 22,600 psi at the de-sign temperature. The bottom head is a layer head, the same thickness as the shell but using A. O. Smith VMS W1350 Special Grade A material, which has an allevable stress of 17,800 psi at the design temperature. The top hemispherical head is forged steel per A. O. Smith VMS 5002 mod., 3 -1/2 in. thick, clad with type 309 CB stain-less steel veld deposit.
Various tests vill be performed on this reactor to deter-mine its nuclear stabil$ ty and performance--for example, one test is a " cold water accident" test during which a sudden flow of 4000F vater will be introduced in the reactor while it is cperating at 6000F.
In addition to the high thermal shock stresses induced into the reactor vessel and its components by such an experiment, there vill be a major surge in pressure due to increaced nuclear acti-vity.
3 Core Component Test Vensel (C.C.T.V.)
9 Westinghouse-Bettis Pittsburgh, Pennsylvania Core Component Test Vessel is designed for 2560 psi at 650 F and is used to investigate feasibility of varioue core component designs under actual conditions of prescure, tempera-ture and flov.
The vessel vill not be used as a reactor vescel and vill not be sub 'ected to nuclear radiation. It le at pre-sent being f abrie -
Using multi-live
'ar, Atruction, the vessel I.D. is 47 in, and r
the shell thiu;est. la 4 1/8 in. (thirteen 1/4 in, thick layers) with a 7/b in. inner shell clad with 0.109 in. mininum thickness, type 304, stainless steel. The shell material in A. O. Smith VMS 11350 Special Grade A material, having an allowable stress of 18,750 psi at 650 F.
The bottom hemis-0 pherical head is ASTM A-212 B FBQ material clad with 1/8 in, minimum thick, Type 308 L stainless steel deposited by the automatic submerged " Twin-Are" method.
Tne removable top head has a bolted closure with a 15 in. I.D.
vessel mounted on the main vessel head. The 15 in. I.D. top vessel has a quick-openir.g closure with a control rod align-ment mechanism.
The design specification calls for the vessel to withstand as a minimum requirement, a heat-up rate of 3 3 F/ min from 700F to 550 F and a cool-down rate of 3 0 F/ min from 550 F to 70 F vith a corresponding pressure cycle during these transients from O psiB to 2000 psig.
4 <
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h.
Puel Element Tent Au t.oc l uve Westinghouse-bettin -
P l t.tnburgh, pera:nyl v %1 h This is a small multi-layer autoclave designed for ?,000 lain at 700 F,16 in. I.D. vith a shell thickness of 2 in. (.o x 1/4 in, thick layers) and a 1/2 in. inner shell clad with 1/8 in. nominal, Type 304 stainless steel. The uhell naterini is A. O. Smith VMS 11350 Special crude A, which hv an n1!ow-0 able stress of 17,600 psi at 700 F.
The top and bottom heads are both flat. The top nead is (1 in, thick, ASIM A-212 B FBQ material with 1/8 in. mininum thi ch, type 309, stainless steel veld deposit. The botty head le 9 1/k in. thick, ASTM A-105 Grade 2 mod. material with 1/b in.
minimum thick, type 309 stainless steel veld deposit. Thin vessel has been in operation since the spring of 1959 Reactors 2, 3, and h have State of Pennsylvania "Special" design approval and are stamped necordingly.
C.
Expected Puture Application 1.
Commercial Application Unprecedented extensicn of the frontiers of chemical engineer-ing processing is calling for equipment to resist h16her prec-suree and higher temperatures of operation.
A. O. Smith haa fabricated vessels for pressures up to 22,500 psi for the process industries and is at present designin6 and will fabri-cate multi-layer vessels for 3$,000 psi on an order recently received. Commercial multi-layer vessels have been fabricut ed -
up to 120 in, in diameter and 14 in, vall thickness.
With continued research and development of high ctrength nteeln, these pressures and dimensions can be exceeded to meet induntry increasing demands.
2.
pelear Application the current and partit larly the future application of multi-layer to nuclear reactor vessels is for high pressure and/or large diameter vessels. The extension of the present boiling water reactor technology to the supercritical temperature region vill call fqr ten foot diameter vesseln designed for 5000 poig pressure and elevated temperatures. Thin is a renaible design using multi-layer.
Economical power frem nuclear sources appears to be poonible for large power plants. If there.is sufficient demand for large reactors, it would be feasible to shop fabricate large diameter multi-layer vessels up to 12 ft. in diameter and
{
larger vessels h the range of 16 ft to 18 ft. diameter by a combination of shop and field assembly to overcome shipping limitations.
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BUPPL1!NENTARY TECHNICAL INFORMATION ON THE SAXTON. REACTOR V5TCIG 7tepared Ey:
L. R. Katz /f f' August 17, 1960 i
Appmved: k,
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b A.
.Xt1"?.i th, Manager Wna r-Ujotems Seet3on I
WESTINGHOUSE ELECTRIC CORPORATION ATCMIC POWER'DEPARDGNr P. 0. Box 355 PITTSBURGH 30, PENNSYLVANIA S
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TABLE OF CONTENTS Fuge No LIST OF FIGURES.....................................................
11 I
INTRODUCTION................................................
1 II MULTI-IM ER HIST 0RY.........................................
.2 III QUALITY CONTICL ON THE SAXTON VESSEL........................
3 A.
Summary of Inspections..................................
3 B.
Ultrasonic Inspection - Genera 1.........................
3 C.
Ultrasonic Inspection of Nozzle Welds...................
4 D.
Ultrasonic Inspection of Girth Sear Welds...............
h E.
Tightness and Contact Between Multi-layer Wraps.........
4 IV SERVICE STRESSES IN THE SAXTON REACPOR VESSEL...............
6 A.
Cyclic Stresses.....................................
6 B.
Thennal Stre sses at the Main Nozzles....................
8 C.
Stress-Baiser Effect of Weep Holes......................
O D.
Neutron Irradiation Effect..............................
9 E.
Comparison of Strength in Equivalent Multi-layer and Solid Sections................................................
' C F.
Ductility in Multi-layer Shells.........................
1C i
V OPERATING LIMITATIONS.......................................
16 A.
Weep Hole Imakage.......................................
15 l
B. - Heat-up and Cool-down Rates.............................
I6-C.
. yd ro static Te st Tempe rature............................
13 APPENDIX A A 12 STING OF MULTI-IMER VESSELS CURRENTLY IN SERVICE.
19 l.
APPENDIX B SUM 4ARY OF INSPECTIONS FOR THE SAXTON REACTOR VESSEL.
23 i
REFERDiCES..........................................................
30
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LIST OF FIGUIE Figure No.
Title 1
Saxton Reactor Vessel - Cyclic loading Conditions.
2 Macro-Hardness Survey of a Typical Non-Stress Relieved Weld Between Forcing and layer Plate Material.
3 Saxton Reactor Vecsel - Weep Hole Sampling System Schematic Diagram.,
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I INTRODUCI' ION This report summarizes technical infomation on the Saxton Reactor Vescel re-quested by personnel of the Reactor Hazania Evaluation Branch of the AEC for their use in preparing a fomal reconsnendation to the Advisory Committee on Beactor Safeguanis on the use of multi-layer construction for the Saxton appli-cation. The information presented specifically covero the questions raised by Meecro. N. Grossman, E. C. Miller and E. G. Case of the AEC based upon their review of Westinghouse AFD report WCAP-1391 " Multi-Imyer Conctruction rer the Saxton Beactor Vessel," (Reference 1) previously submitted to AEC.
We ansvers to the various questions by the AEC have been catagorized under four main headings, namely, Multi-layer Ilistory, quality Control, Service Stresses, and Operating Limitations. We feel that the information to answer the specific questions raised by the AEC are included under each of these four main heading.
he data presented in this report represents a,)oint effort by the A. O. Smith Corporation and Westinghouse Atomic Power Deyartment.
. 1 D
II MULTI-IAYER IIISTORY Over the past 30 years, A. O. Smith Corpo n tion has designed and manufactured approximately 8000 multi-layer petroleum and chemical vessels of every variety and description with de.eign pressures up to 35,000 psi. Many of these vessels which have been in service for many years are similar to the Saxton vessel in design and are operating under similar conditions of temperature and pressure, but differ only from the standpoint of the gamma heating tsfrect which is not present.
Appendix A is a listing of multi-layer vencelo manufactured by A. O.
Smith and currently in service.
This listing includes the operating conditions and diameter of each vessel and the number, size and proximity of location of the nozzle penetrations for each. Note the similarity between these veccels and the Saxton vessel, expecially with respect to nozzle design and proximity of location.
This listing then indicates that the decign and location of the nozzles on the Saxton vessel does not represent a departure from decigno applied to other multi-hyer vessels which have operated safely over many years.
4
-2
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,III QUALITY CONTROL ON THE SAXTON VESSEL A.
Surr.ary of Inspections Appendix B of this m port is a complete and comprehensive listing of the manufacturing procedures which require a non-destructive test or inspec-tion to assure the quality of workmanship and material in the Saxton vessel. Also included rare references to the procedures employed and the acceptance standards for each test or inspection which have been agreed upon by both A. O. Smith Corporation and Westinghouse Atomic Power Depart-ment. Of special interest are the ultraconic 'inopectionc of the nozzle attachment velds ard the circumferential velde joining the bottom head and bolting flange to the multi-layer shell, cince these procedures reprocent a deviation from the radiography requirements of the ASME code. Paragraphc III-B and III-C (below) discuss the-procedures and acceptance standards for the special ultrasonic inspaction of girth volds and nozzle velds in greater detail.
B.
Ultrasonic Inspection n==*ral Since the definitive radiographic examination of nozzle and girth seam velds for Section I or Section VIII ASME Code approval is not possible becauce of the interfennee of seams between layers in a multi-layer shell, A.- O.
Smith has developed an ultraconic incpection technique which Westinghouse concidere cc reliable ao radiography for detection of cignificant imperfec-tions in velds. The ultrasonic testing method for both nozzle attachment and girth velds on the Saxton vessel is covered by A. O. Smith's Non Destructive Testing Pmcedum MV-80016-UT-3, dated 7-15-60 (Reference 5).
This procedure outlinen the testing technique and alco definec the applicable acceptance standards agmed upon by A. O. Smith and Westinghouse APD.
The acceptance standanis agreed upon are based upon those defined by ASME Code (1959),Section VIII, Par. W -51 (m). Specifically, veldc which are found to have any of the following types of imperfections chall'be judged unacceptable:
1.
Any-type of crack or zone of incomplete fusion or penetration.
2.
Any elongated slag inclusion which has a length greater than 3/4 inch.
3 Any group of slag incuulunu in line that have a aggagate length greater than T in a length of 12T, except when the distance between the successive imperfections exceeds 6L where.L is the length of the longest imperfection-in the group.
The ultrasonic test is not considend as precise as radiography for detec-tion of porosity. This condition m sults from the fact that the general cpherical shape of individual porea causes scattering of ultrasonic waves.- o
l l
t The larger size imperfections vill be detected as discrete indications while closely spaced imperfections win cause attenuation of back reflee-tions. The agreement between A. O. Smith and Westinghouse APD on acceptance standards for porosity in as follows. A veld win be considered unacceptable if it contains porosity exceeding the following limits:
An individual pore size larger than that shown in the ASME Code (1959)
Section VIII, Appendix IV, in the porosity chart marked "large", for plates over 2-1/2 in. thick, and a pore distribution spaced more clocely than that chown in the ASME Code (1959) Section VUI, Appendix IV, in the porocity chart marked " Fine", for plates over 2-1/2 in thick.
C.
Ultrasonic Incpection of Nozzle Weldc During the week of August 15, 1960, the results of the firct incpection of all fcur nozzle attachment velds were reviewed. jointly by A. O. Smith and Westinghouse representatives.
Included among the Westinghouse reprecentatives was an ultrasonics specialist from the Materials Engineering Department.
The standard reference indicationo for these nozzle attachment veldc were reprecented by 1/4 in. diameter flat bottom holes drined into the outcide diameter of each nozzle immediately above the veld. The ultrasonic response from each of the holes van thus recorded on each Brush trace as a zvference for comparison with all indications picked up. With the aid of magnetic particle probing, it was proved that good correlation between those indications and actual dercetc was obtained. This correlation applied to imperfectionc both larger and smaller than permitted by the at:ceptance standards. On this basic, it van concluded that A. O. Smith has developed a natisfactory ultrasonic test procedure-for the nozzle velds.
D.
Ultrasonic Incpection of Girth Geam Weldc The cirth veld inspection vill be cimilar to the already cuccessful nozzlc weld incpection. Two flat bottom holco,1/4 -inch in diameter and 1 inch deep, will be drined radially into each circumferential ceam of the Saxton vessel. -The ultrasonic responce from the side of each of these holec~ vill thus be reconied on each -Brush trace, as a reference for comparison with an indications picked up.
All test holes used as reference indications in both the nozzle and girth veldo vill be repaired by velding and incpected by both magnetic particle and ultraconic techniques.
E.
Tightness and Contact Between Multi-Layer Wraps In the manufacture of multi-layer shells, the tightness and contact betweer.
layer platen becomes significant from the standpoint-of themal conductivity -
of the shell and the themal strences which may be developed due to a temp-erature gradient in the shell.- This fact la especially significant in the manufacture of the Saxton vessel multi-layer shell because of the gamma
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t heating which vill occur. The results of an analysis of the Saxton vessel for both pressure and thermal stress conditions as-reported in Reference 1, indicated that the operating stresses are vell within code limits. The thermal conductivity values of the Saxton vessel multi-layer shell for this analysis was chocen based upon laboratory testo on multi-layer chells manufactured and inspected to standard A. O. Smith specificationc. The inspection of the Saxton vessel multi-layer shell for Inyer contact and tightness likewise conforms to A. O. Smith's layer Tightness Inopection Procedure for Multi-layer Construction, MV-80016-LT-1 (Reference 6).
This procedure defines the inspection method and the standardo of accept-anee sgreed upon by both A. O. Smith and Westinghouse APD. The inspection technique for checking layer tightnecc and contact concicts of a method in which each layer wrap in methodically tapped with a suitable. tool af'ter the vrap is applied and welded. Any loose areac between layer vraps represents a distinct change in the ringing cound produced by the tapping procedure. The effectiveness of thin tapping technique van demonctrated by A. O. Smith to Westinghouse engineering and inspection preconnel on the incpection of the Saxton vessel shell. This inspection procedure also requires a " feeler" gauge measurement of any gaps which can be found between layer wraps as measured at both ends of the shell after each vrap is applied. The standards of acceptance agreed upon are the same as those previously established by A. O. Smith for commercial, non-nuclear vessels. Specifically, loose amas or gaps between layer wrapa which exceed the following limite shall be judged unacceptable:-
1.
A loose area greater than 12 in, circumferentially and/or 24 in.
longitudinally. In the event of more than one loose area circum-femntially in any 24 in. length, the total of.such-areas chall not exceed the ama prescribed by the limit specified above.
2.
A maximum single radial gap betve.en any two layers ac measured at the ends of the shell courses exceeding.020 in. or an area of such a gap
. measuced at right angles to the vessel axis in excess of.120 sq. in.
In ortier to justify the use of this comercial acceptance standard for vrap tightness to a nuclear vessel, A.-O. Smith _has performed an analysis to determine the effect of the permissible loose area on the thermal' stresses in the Saxton vessel shell under the conditions of gama heating at full power operation. This analysis was perfomed using the general equations for strescen in multi-layer shells previously developed by A. O. Smith in conjunction with Franklin Institute, as described in Refennee 1.
The resulto of this analysis indicates that if a loose area of 288 sq. in. as aH oved by-A. O. Smith Procedure MV-80016-LT-1 existed between the inner barnl and the first layer wrap of the Saxton vessel, the resulting tempera-ture gradient and themal streso in the multi-layer shell vill be well within safe limito. At 40 Mr (thermal) operation the temperature gradient at the loose area in the natlti-layer shell between the inner barrel and the.first layer wrap win be approximately h0*F resulting in a thermal ctreso under 5000 psi. The same loose ama between-the outermost layers would yield a temperature gradient of less than 5*F.
=.
_ _ _ _ _ _ _ _ _ _. - - _ _ _ _ _. _ _ _ _ - - _. - _ - - - -. ~ - - - - - - -
IV SERVICE STRESSES IN THE SAXTON VESSEL 1
The problems accociated with the preocure and thermal atmunen in the multi-layer shell of the Saxton veccel under full power steady-state operation are covered in Beference 1.
This par tion of the report covero the stressen in the multi-layer shell resulting from trancient operation including the cyclic ctmsses and the effects of streca raisers introduced by nozzle penetrations and weep holes.
A.
Cyclic Strences The effect of cyclic ctrecoco on the Saxton vencel have been analyzed with respect to all expected cyclic loading conditions and found to be well vithin safe limits. This analysis van conducted by A. O. Smith baced upon j
the expected cyclic loadings including those reculting from trancienta-introduced by plant start-up, shut-down, load changen and certain credible accidento, ac supplied by Westinghouse AFD and listed in Figure 1.
The j
calculations on the cyclic stresses in the Saxton vessel vere perfomed in accordance with United Statec Dept. of Commerce report " Tentative Structural Decian Bucia for Reactor Prescure Veccela and Directly Accociated Components", December 1, 1958.
(Refe mnce 8).
The inherent ability of multi-layer construction to safely withstand the effects of cyclic loading has also been demonctrated in actual operating experience and chop testing of multi-layer vencela. An example is a multi-layer accumulator vessel built by A. O. Smith for the Aluminum Company of Am2rica in 1941. This veccel is 32 in, incide diameter and van designed for 4500 psi operating preocure. The mul_ti-layer shell vac manufactured from VMS-W-135 Special Grade ~ A layer plate, the.same material an used in the Saxton vessel shell, and the shell vac designed using a safety factor-of five on the ultimate strength of the material. There are no penetrations in the multi-layer chell. While in the service of the customer the vessel.
was subject to approximately 4-1/2 million pressure cycles from 4100 to 4500 psi. This vessel vac returned to the Milwaukee plant-of A. O.-Smith-after approximately 12 years of customer cervice-for further cyclic testing.
The shop test on this vecsel van conducted by cycling thel' sel through-a pressure range of 4000 poi to 5500 pot which exceeded the design pressure by 22 percent.
Fatigue testing equipment produced 16 cycles / min. and 20,000 cycles were obtained per day. After 2,368,530 tect cycles vere-completed, water was~ noticed at one of the veep holes indicating a leakage in the inner shell. The leakage rate at 4000 poi was small. It_vas decided to continue-preasum cycling the vecoel._ A-periodic incpection_vas-maintained on the leakage rate through two veep holes located adjacent to one of the circle veld; seams. After 5,002,500 test cycles, some ten months after the leak was detected, the fatigue test vac concluded with no appreci-able increase in the leakage rate over that period of time.
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, Max. beatup rate 200*F/br f
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' Figure 1 --Saxton Resetor vessel Cyclic Inading Conditions
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Next, an attempt was made to burst tlP vessel. The burst test was not successful because the leakage through the weep holes at 13,500 psi was too great for the capacity of the pumps.
The test acetcu3ator was returned to the shop and one head was cut off to permit access to the interior of the vessel. Weld repairs were made to the inner shell.
The shell end and head were scarfed and re-velded.
Another hydrostatic test was conducted on the vessel at 9000 psi for a duration of one hour without any 3eakage.
Pressure was then increased gradually, and at a pressure reading of 15,500 pai there again was leakage at the weep holes. It was not possible to burst the vessel because the leakage exceeded the capacity of the pump.
This test demonstrated several important features of the multi-Jayer construction. The failure of the inner shell (as indicated by the leakage at the veep holes) did not propogate through the multiple layers of steel after some 2-1/2 million additional pressure cycles.
leakage through the weep holes can be a safety feature indicating when the inner shell has failed. This should be especially important for atomic reactors where radiation damage can occur in the steel.
B.
Thermal Stresses at the Main Nozzlen The thermal stress effect in the Saxton multi-layer shell at the main coolant nozzles was analyzed by A. O. Smith and found to be insignificant. This analysis was based upon the use of the general equations for stresses in multi-layer vessels, described in Reference 1, and a gama heating value of 350 Btu /hr/cu. ft. as calculated by Westinghouse AFD at the inner surface of the multi-layer shell, adjacent to the main coolant nozzles for 28 W (thermal) operation. The calculated temperature rise due to camma heating at the nozzles is less than 1!F for 28 W operation, and causes a negli ible 6
thermal stress.
l C.
Stress Raiser Effect of Weep Holes i
l In the history of the manufacture of multi-layer vessels by_ the A. O. Smith Corpuration there never has been a failure of a vessel due to a defect originating.at a veep hole. This is true both for the tests to destruction which have been conducted-by A. O. Smith and operational experience on vessels in service.
l Tables of allovable stresses for vessel materials were originally established for the purpose of comparison vith stresses calculated on the assumption of-perfectly homogeneoue materials of construction, free of. internal imperfec-t-
tions such as slag inclusions. The deviations fzom calculated stresses due to small internal imperfections which actually exist in practice are commonly ignored. The stresses around the small veep holes in the Saxton vessel vould -
seem to be of the same type end order of magnitude as those which would result at any local imperfection, and could therefore be ignored for the same reason, namely that they would be relieved by local yielding in a ductile material.
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We vould also call attention to the fact that multi-layer construction has the inherent advantage that local yielding propagation in the radial direction ic effectively prevented by thc layer interfaces.
The effect of the weep holes ac ctress raisers hac been analyzed for both static atal the most ceven dynamic loading conditions which v113 be imposed on the Saxton veceel. The resulto of thic analycic indicated that the vecp ho3ec have a negligible cffect on the acryice stresceu in the veccel.
The otatic streca raiser reculting fmm the 5/16 in. Veep holec in approximately 2.63 (Heference 31). This streou raicer recultc in a maximum tencile streca in the veccel which wouhl be 51,300 poi except for the fact that the metal vill yle3d at the actual yelld point of the material ( hl,250 poi).
Although thic value ic above the yield point of the chell plate material, local yielding at the hole vill relieve the strecces in the plate material to a cafe value.
Ao explained above, a ductile member loaded with a cteady ctress dnec not cuffer loss of strength due to the pmcence of a notch or hole, becauce of the ability of the material to yield at the localized ares of higher strecc.
The effect of the weep hoje an a streca raiser under cyclic 'oading vac analyzed for the three most ccvere conditions of dynamic loading to which the vencelc will be cubjected. Thece loading conditionc could result fmm three plant operating conditions, namely, a 10 percent loco of load during normal operation, normal startup, and an accident condition resulting in a complete 10c0 of coolant. The fatigue ctreco concentration factor recu3 ting from the S/16 in. veep holes in approximately 2 55 (Reference 9). The results of a fatigue analycio of the three loading conditionc outlined above are given in the following table.
Actual No.
Temperature Preocure of Cyclec Dare No.
Variation Variat.lon During plant of loadinc Condition
- F Poi Life Cyclec 1.
lose of load (10%)
534 - 528 2050-1975 1500 Infinite 2.
Namal Startup 50 - 530 15-2000 50 90,000 3
Ioco of coolant 530 - 120 2000-50 1*
800
- loca of coolant ic considered a credible accident condition.
D.
Neutron Irradiation Effect The effecto of neutron flux upon the Saxton veccel are identical to the effectu which would be imposed upon a colid vall veccel for the came applica-tion. The problemo accociated with thoce neutron flux effecto, - therefore, are the came for the Caxton veccel ac for arc reactor vencel being opecified by Westinghouce APD, and the same design approach and testing programs vill be -
followed.
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The nil ductility temperature of th( ot Lual ch(ll plate uced in the manufne-ture of the Gaxton veacel vill be deterwincd on aampleu prior to Ini tial ope rat i on.
Operating Irradiation tecto of chell plate ntunpleo will be conducted in the Saxton vessel. These camples vill be encaced in water-tight utainlenu nteel cano and aucpended in the operatind veccel between the thermal chield and the vessel vall.
At the end of the varioun periods of plant opcration, cample tubec vill be extracted from tne reactor and tects will be conducted to determine any changen in the phycical properties of the materla] including the nil ductility tempe rature.
As thece tcote art conducted on camples expoce<1 to inercucingly longer periods of neutron bombardment., the progreccive effect of neutron flux on the veccel wall can be monitored.
Ihced upon the nil ductility tempera-tures found from these periodic tento, the oIcrat ind procerfuren vill be written to prevent full precoulzation of the veccel until a veccel val) temperature of NDT + 60*F is attained. Dec puragraph V C.
E.
Comparicon of the Str"ngth of Equivalent. Multi-Inyer and Colid Scetionc A basic advantage in multi-layer vercuc solid vall construction recults from the fact that many layers of thin plate have a cuperior tenci.le ctrength ac compared to a colid plate of equivalent material and thickness. Thic condi-tien 10 due to the size effect, in that t hinner materialo have inherent),y cuperior physical properties.
An analogy to thic condition in the une of a cable rather than equivalent diamet.cr bar of the came material for tencile loadc. Gimilarly, the creater utrength of a cable is due to the curerior physical properties of the many cmall wirec uced in its manufacture.
The curerior ctrendth of many layern of thi n plate ao compared to colid plate of the came material and thickneco have been demonctrated by A. O. Smith in tencile tects. A typical tect of thin type vac conducted on a tencile cpecimen of A. O. Smith 1146 material. Thic cpecimen vac made up of 4-1/2 in, thick forcing butt velded to a atack or platec, the firnt 1/2 In. thick, and sixteen plates 1/4 in, thick each. An expected, the fracture occurred in the forced material at an ultimate tencile strength of 80,300 poi. The thinner layer platec have an ultimate tencile strength of 105,000 poi.
Reference 10 la a photograph shoving the fractured specimens reculting from thic tencile tect.
F.
Ductility of Multi-Inyer Ghelle An decertbed in Reference 1, the final octrecr relief of all precourt veldo required by Section I or Section VIII of the ASME Code 10 purpouely climin-ated in the Saxton vencel, ac it lo in all multi-layer vencela. The ntrecc relief in a multi-layer venoel would tend to remove the dccirab3e precomprecs-ive ntrecces built into the chell luycrc, with the reculting loco of interface precoure between Inyerc, which reduces the overall thermal conduc'tivity of the chell.
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Decause of the inherent ductility and close velding control employed in multi-layer construction, the el.imination of stress relief dnes not limit the reliability or compromise the safety of this type of construction in pressure vessels. In the velding of multi-layer cheD.s to solid vall forgings and plates, any residual stresses in the veld metal and heat affected zone are greatly minimized through lateral movement of the individual layer plates during velding.
In addition mechanical peening has proved to be effective in contnalling distortion in these velds. Also, the use of lov hydrogen velding electrodes enhances the ductility of the ac depocited veld metal.
Further, the veld beads in a multi-layer section joining a coild vall section are multi-pacc, and each cuccessive ve.1d layer progrecc-ively grain refines and heat treate the preceding layers and heat affectc.d zones.
Macro-hardneco surveys conducted by A. O. Smith on non-ctreou relieved cample sections of multi-layer plate to colid forgirica weldo indicate that neither the veld metal nor the heat affected zones are execccively hartrned above the rance which would nomally be expected for carbon cteelc. FMure 2 chovo a macro hardneso curvey of a cpecimen from a forcing of /CTM-A 10';,
Claco 2, material velded to twenty-four layern of VMS W 1%G Spccial Grade A material. This specimen vac made up of the identical material uce<*. in the flange forging and multi-layer chell of the Saxton Veccel and wac not strecn relieved af ter velding. The chemical analysis of the materialc upon which thic macro curvey vac made ic no follova:
Material C
Mn p
S Si VMS W 135G 30 54
.014
.022
.10 Spe cial Grade "A" Inyers SW-45 Weld Material
.13 32
.020
.019
.020
/SIM-A 105 Class 2
.25 1.18
.018
.044 28 Forging Note that the maximum hardnecc found in thic cpecimen in 232 (Vickerc) in the forging heat affected cone.
The cafety and reliability of non-streco relieved vescelc has alco been demonstrated by A. O. Smith in toth shop tects and the following field operating experier.ces.
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INTRODUCTION I
This report summarizes technical information on the Saxton Reactor Veccel re-quested by personnel of the Beacter liazards Ehruluation Branch of the AEC for their use in preparing a fomal recommendation to the Advisory Com:nittee on Beactor Safeguards on the use of multi-layer construction for the Saxton appli-cation. The information presented specifically covers the questions raised by Messrs. N. Grossman, E. C. Miller and E. G. Case of the AEC based upon their
- view of Westinghouse APD report WCAP-1391 " Multi-layer Conctruction For the Saxton Beactor Vessel," (Reference 1) previously submitted to AEC.
The ansvers to the various questions by the AEC have been catagorized under four main headings namely, Multi-layer llistory, Quality Control, Service Stresses, and Operating Limitations. We feel that the information to answer the specific quections raised by the AEC are included under each of these four main heading.
The data presented in this report represents a. joint effort by the A. O. Smith Corporation and Westinghouse Atomic Power Department.
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In 1937, two multi-layer vessels were tested to destrue"on to establich information on the effect of stress relieving temperatu.~a. upon multi-layer construction. Each vessel was built of 12 layers, 1/L 2n. thick vrapped around a 1/2 in. inner shell giving a total vall thicknece of 3-1/2 in.
The vessels were 19 in. I.D. and approximately 7 ft. long. The vesce3s were built of layers which came from identical sheets of steel, split in two at A. O. Smith co that one-half of each sheet was uced on one ver cl, and the other half on the other vessel in identical positions. The princi-pal difference was that the vessel decicnated "A" vac not stress relieved w h reas vessel "B" vas stress relieved at 1250*F. The stress relieving temIerature in vessel "B" was held for 3-1/2 hourc; then cooled at the rate of 80*F per hour. The average phycical propertlec of the steel in the two vessels, determined by test coupons cut from each plate were as shown in tue table below. The test coupons for vencel "B" were given the same r.trenc relief ac the vessel.
Average Phycical Propertien of Steel in Vecceln Couponc O.0% Get 111tima te Etoncation femarka From Yield Str.
Tencile Str.
in 8 in.
VecceQ (pai.)
(pal.)
(Q "A"
39,770 58,8h5 W.5 Not Strecc IWileved "B"
37,670 56,260 32 3 atreca Relieved The precsures and stresses at which the yielding and failure of the tvn vencela are licted below:
Stresses at Yield and Failure of the Vessels vessel O.2% Yleid by Remarks Water Column Burct Point Precnure Streca Precsure Streco (pai)
(pci)
(psi)
(pai)
"A" 13,400 43,100 18,800 60,400 Not Streus Relieved "B"
11,500 37,000 17,375 55,900 Stress Relieved It should be pointed out that the ctrences at yield and failure for the utress relieved vessel "B" are very clone to the yield, cnd ultimate stren6th of the a+ m sc relieved test coupons. On the other hand, the strences at yield and failure for vessel "A" are somewhat higher than the yield and ultimate ctrengths for the non-strent relieved test coupons. It is believed that this may be due to the cold working of the steel in fonning the shel.lc.
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Thin experiment choved that both the otreco relieved and not strccc relieved vennels failed in a ductile manner cloce to the cal ulated preonure icvels. Aloo, the non-ntreou rt lieved vecuel witn the ntroni'er of the two.
Armther test conducted on a multi-layer ve ocl by the U. U. Navy deron-straten the inherent ductility in thin type of constrvetion.
In order to qualify for the non-chatterable requirtmente nf Navy A1 r Flacko, a multi-layer veccel van cubjected to a ballictic te at at hthlp,ren providit groundo. The veccel van 14-3/8 in. 1.D. With a 2-1/8 in. vall thitAneoc and a 48 in atraight icncth.
It van decicncd for h500 poi preocure, using layer material cimilar to ASIM A 225 Grade B, and a factor of cafety of 4 on the ultimte strength of the mttrial. Their vere no nottle penetrationo through the vall.
While charged with sir at 3000 pai prvncure the vesse:1 vac posittorad on the firing range no that the longitudinal axin var perpendicubr to the line of fire at a range of approximately 50 ft.
The veccc1 vac cubjected to the standard test of being hit by a 1.3 in. projectile. This pr @ tfie did not penetrate through the inner chell. It vac tien agreed to fire a larger projectile (40 MM) at the vecccl immediate]y adjacent to the 3.1 in, hit. The 40 mm projectile penetrated the entrant e vall and projcetod through the exit vall without complete rupture or fragmentation of the tank.
Thic report and photographu are contained in Da Shipo letter S49 (Sh8D)
Scr. h548-2506 of June 28,195P to A. O. Smith Corporation vin ITN, l
Milwaukee.
This test demonstratec that under impact multi-laycr fallorc 10 ductile.
In 1953, the A. O. Smith Corporation built two high pre.ccure chemical 4
reaction vencelo for the Demet Golvay Corporation, a divicion of Allied Chemical and Dye Corporation, located near Ibffalo, New York. The reactorn were built of multi-1Ayer construction with a standarv3 Bridgman clonure at the top head. The vencel van 36 in, incide diameter,10 9/16 vall thicknocu, designed for 7500 poi with VMS-W135, Special Grade A layern. The vennel van decigned with a factor of cafety of four (4) on the ultimate strength of the material.
Five small nozzleo penetrated the multi-layer vall; the openincu in the vall were 1-3/4 in. In diameter.
i The two voosels were enclosed by a 12 in. thick reinforced concrete cafety vall on three sideo with the fourth cide open to a Inrge field and the top cf the enclosure oper to the atmocphere. A 12 in, concrete vall alao celet-rated the vertically supported reactors.
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Chortly after the petro chemical plant vao placed into cperation a failure occurred in the cooling syctem resulting in a rice in temperature of the contento of the veccel, and a runavuy reaction occurred. There vac an instantaneous build-up of pressure within the rcactor to excescive propr,r-tions, estimated to be above 25,000 poi. The preocure blev the 2 in, rupture dice, cet for 12,000 poi, installed on the top end flange of the reactor. The gaseoue diocharge from the reactor caught fire. The fire was brought under control with no injurico or loco of life to plant pernonntl.
Incyetion of the reactor a few days later revealed only minor damage to the top head and copper gasket, and no damage to the mujti-layer chell.
The f orce of the explosian left 1/8 in, deep impreccions of the ruin voltin; nuto on the retainer ring. After making the necconary minor reintro ta the top head and installing a new gacket, the reactor was placed back into operation and has been operating catinfactorily cinee 19M.
Thio necident demenctraten the ability of multi-layer corotructiona with penetrationc und veep holes to withstand an explocivt type loadity; of at lenet three timec the design precoure.
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y OPPRATING 1JMITAT10NS In order that the streco IcVe]s in the Saxtori reactor vence] be mi tntrai ned with-in the cafe limits for which it was dec1gned, certain limitationo mact be im,90ced on the vencel operation. These 11mitationo are related te maximum a21ovable veo})
hele leakage, maximum heat up and cooldown : sten, and rninimum hydroctatic tect temperaturen.
A.
Weep Hole Irakage The cix weep holco drilled into the Caxton vc urel multi-loyer che]1 t.i carry off any Itakage. which could renult frvm a crack in t he inner barreJ, v111 l'e piped to a campling cyntem for continuoun monitoring.
Figu re 3 in n rebemtle diagram of the weep hole campling cyclem.
Dich of the oix weep holen, t hree of which are located near the upper girth acam, and thrt t Juct nbeve tDc lower girth ceam are cennected by 1/h in. t ubing and entried t o a cowm header line at the bottom end of the vencel.
Thln htadr r io then onnrect ed t o a pienrure trancmit ter vired to a preenuro recorder ar d ejarm located on the panel J n the main control room.
In cernn with the prvccure trancritter in a precnure relief valve cet to open at 50 prig. The blow-off Jir.c from t hic relief valve runo to the chielded plant container cump. Thio blev-off line alco cor.tains two valvec, a local compling valve Jacated on a tee branch, and an icolation valve which to nomal' y Iceked upen.
both of thece valvea art-hand operat. d.
J If a leak chould develope in the inner barrel during plant operat. ion, t he precoure recorder vill indicate a buildup in the precourc in the line between the inner chcll and the precoure relief valve.
If the preocure cont inuen to build up to 50 poig, the alam vill nound in the control roon and the preabure relief valve vill open, allowing the leakage t,o poco to the plant ennt.niner cump. The 50 poig precoure van chonen for the prencure relier valve ret Linc becaum it reprecento a precoure vc11 below the burnting preocure of the l
outermont layer wrap.
Ar>y leuk of nuffielent magnitude to opcn the relief valve and round the preonure alarm vill be cause for brjnging the plant to the " hot chutdown condition".
During this condition of chutdown, the main coolant loop temperature and preanure are maintained, but the reactor ic ucrammed to l
zero power level. Plant operatinc peroonnel are then permitted iroide the plant container. The local campling valve vill then be opened, an.3 the luolation valve which 10 normally locked open, vill then be cloced.
A cample may then.be taken from the local campling valve to deter'nine the rate and macnttude of the leak.
A leak whleh exhibito an jnerearing rato l
v111 be then connidered cauce for complete cold chutdown and emergency i
repairs to the inner barrel of the vencel chell.
B.
Ileat-up and Cool-down Itateo The desired heat-up and cool-down rate for the Saxton plant has been cotab-liched at 200*F/hr.
k Sinec the reactor vencel is the com}nnent most coverely
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'I effteted by rapid temperature charges in the primary loop, th? Saxte,n vernd has been ar.alyzed by A. O. Omith to detemine the therma) strecces in the vessel reculting from these tem}*.rature changes. The rvcults ef thic ana3ysis indicated that the thermal ctress is highest at the main bolting flange, but are vell within allovable cafe limits. Since the hichest thermal ctresses occur in the bolting flanco and not in the shell, the prob 1cm of heat-up and cool-dovn rate for the Saxton vennel are exactly the came ac for a solid vall vescel in the came application.
C.
liydrociat ic Tent Temierature The hydroctatic tect tem}erature for both chop and in$ tint ficid hydraatatie tooting vill be cet at a value at Jeant 60*F above the nil ductility tempera-ture of the inner barrel and layer wrap mattrial in the multi-layer chcll.
The initial nil ductility temperature of the multi-laycr chell material '13 L be pre-detemined by tectinc campleo taken from the actual pluton used j i the manufnet ure of the Unxton multi-layer chell, an deceribt.d in Paragraph l V b.
The hydrontatic tect tempomture for tente conducted durjnc variouu M.rlodn of plant life v131 be detemined baced upon the nil ducti tity of t he mu'11l-layer chell material at that time. The nil ductility of the chell p] ate material vill be detemined at variouc periods of plant operation by tecting shell plate onmples which hase been irradiated in the operatinc reactor ac deceribed in Paragraph IV D.
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APPL:NDIX D - DI SCH1l'1 ION of compt'TF.8 PROGRAMS liLOWDOWN ANAINSIS - FLASH CODl:
Analysis of the transient flows, coolant inventory, temperature, nn i o c.
of the Reactor Coolant System following a large-a rea rupt ure wa'. per 8 or using the digi tal compute r code rLASll.
This code calculates rate of coolant blowdown and rate of loflux from t he Saf ety inject ion %st em, pressure drop and flow through the core and intact loops, anti account, for energy ent ering and leaving the system, by way of the core an.1 st-o-
generators.
Reactor power is cont rolled by moderator reactivity enterio f unction of time, and a reactor t rip niodel which renresents in e rt1.n as a of RCC'8.
The moderator react ivi t y densi ty cont ribut ion is pre-cal culat ed using a more detailed core model than is now available in FLASil, wi th predicted pressure and flow transients which are checked with FLASH re' ol t,
for conniutency and conservatism.
The l'LASil Code treats the Icactor Coolant System as i f it were comprised g
of three control volumes, and calculates the pressure and inventory of each separately. The selection of these volumen ir a PWR system is made in such a way as to group those portions of the system whose temperatures and pressure are relatively uniform throughout the transient Volume 1 includes the reactor outlet plenum, the hot leg piping and the steam generator tubes.
Volume 2 includes the loop cold leg piping, the reactor coolant pump, and the reactor downcomer and inlet plenum.
Volume 3 includes the pressurir.er and surge line.
I "FLAhll:
A Program for Digital Simulation of the Loss of Coolant Accident,"
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by S. G. Margolis and J. A. Redfield MAPD-T11-534, May 1966.
D-1
ModifIcat ions were made to the original FLASit program to account for it sperifle system configuratlons of the Saxt on system:
the4e im Iwie t i..
single-p.mo rod-type core, anil the locallon of t im react or coolant pow,
and luteralon pumps chararieristirs.
1he result s of FLASil, core cooling inventory, presa,ure quality, llow var.,
through the core, etc., are used for a detailed analysis of the core thermal transient.
CORE l' owl:R TRANSil:NT DURING 11LOWlHlWN The basic tool used for the reactor kinetics calculation is t he till C-K1' Code, which has a point kinetics model and a single channel fuel and ecos.wt de8cription.
in this study the channel was divided axially into five sections, with density in each section a function of pressure and enti. ale.,
plus nucleate boiling void. A nucleate boiling model for highly subcooled conditions was used, even though a large part of the coolant is saturated i
throughout the transient. This was done to minimize apparent void for-mation in order to retard reactor shutdown and yield a conservatively high energy input.
Average core pressure was input as a function of time from the FLASil output.
For small breaks with forward flow, the cure inlet flow as shown by the FLASli code calculations was used as input to CHIC-KIN.
lor large cold leg breaks, with violent flow reversal and then near-stagnation, the core pressure drop as indicated by FLASH was assumed to be a reasonable representation of the forcing action between the two large liquid regioni:.
of the system. This pressure drop was used as input to CHIC-KIN, which calculated flow-response taking into account inertia and losses at inlet and outlet and from grids and friction in the fuel.
The resulting flow transients were very close to those obtained by FLASH.
Each axial fuel rod section was divided into nine radial regions for the heat transfer calculation.
Fuel properties are constant in the code, and best estimate fuel thermal properties were used to determine the core
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Ked fleid V. A., "ClllC-KIN -- A Fortran Program for Intermediate and Fant Transients in a Water Moderated Reactor, W.APD-TM-479 January 1,1965.
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average fuel heat transfer respont.e and the resulting void formation.
Since hot channels have greater than average void fraction, even if DNh occurs in them, neglecting the effect of the hot channels reduces the total apparent void and thus yields a conservatively high energy input.
Trip would be actuated by overpower for the cases with a significantly positive nx3derator coefficient or by low pressurizer pressure with a zero or negative coefficients.
For the small break cases, depressurization after the initial subcooled blowdown is slow, and thus shutdown or voids is also slow. Trip is required in these cases, and-in ClllC-KIN this u.s simulated by a ramp insertion of negative reactivity.
For the large break trip would be similarly actuated, but because void formation is adequate for shutdown, trip was not simulat ed in these studies.
Doppler reactivity feedback was simulated as a function of the average fuel temperature, with a weighting factor of 1.4 used as a lower limit for the initially unrodded core.
Six groups of delayed neutrons were used. A conservative maximum of 0.0072 was used for the delay iraction to slow down power decay.
CORE THERMAL ANALYSIS The LOCTA-R2 tran:.icnt digital computer program was developed for evaluating fuel pellet and cladding temperatures during a loss of coolant accident, it determines the extent of the Zircaloy-steam reaction and the magnitude of the resulting energy release in Zircaloy clad cores.
The transient heat condition equation is solved by means of~ finite dif-ferences, considering only heat ilow in the radial direction.
A lumped parameter method is used - the fuel containing three nodes and the cladding one node, k
3
Internal heat generation can be specified as a function of time, or the decay heat from any initial power level can he calculated by the code.
The decay heat is based on the heat generated from:
a)
Fission Products, o)
Capture Projucts, and c)
Delayed Neutrons It is assumed that the core has been irradiated for an infinite period of time.
In addition to decay heat, the code calculates the heat generated due to the Zircaloy-steam reaction. -The Zr-Il O reaction is governed by the 2
parabolic rate low unless there is an insufficient supply of steam availabic, then a " steam limited" evaluation is made. 1he buildup of the Zircaloy-oxide film is calculated as a function of Iime, and its effect on heat transfer is considered.
An isothermal clad melt is con-sidered haaed on the heat of fusion of Zircaloy.
Once the Zircaloy metal 1
melts, it is retained by the Zirc-oxide, and slumps against the fuel.
The Zirculoy-steam reaction may continue until the oxide melts.
If the oxide melts the remaining Zircaloy is assumed to fall, and 10% of this metal is assumed to react with additional water which is available in the vessel.
The code has been developed to stack axial sections and thereby describe the behavior of a full length region aa a function of time. A mass and energy balaace is used in evaluating the temperature rise in the steam as it flows through the core.
The initial conditions of the fuel rod are specified as a function of power.
The following core conditions are also introduced as a-function of time; as determined by the FLASit Code:
(
4
a)
Masa flow rate through the core b)
Coolant quality c)
Pressure d)
Liquid level Heat transfer coef ficients during the various pha*,ew of the accident
..r-evaluated in the followa'r, mannert ai Nucleate boiling filin coefficients on the order of 20,000 lit u/br ft
'l' are used until DNil.
b)
When DNil occurs, it is aHHumed that the fuel rodH can Immediately develop a condition of stable film boiling. No credit is taken 3
for higher transition boiling coefficients that exist prior to the establishing of a stable film on the fuel rods. The correlation used during this period in k
D Q eu 0.4
" (e" " (r 3 + Qv 0.8 h
.023 !)
))
( F, )
=
n A
y c
y c)
During the time the core is uncovered (period of steam flow through the core), laminar or turbulent farced convective coefficients and radiative coefficients and radiative cou!!.' ients are evaluated.
For laminar forced convection to steamt
(
)
3.66
=
iso
.25 3
(f) h/h iso w
For turbulent forced convection steam:
f=0.020(R b) *
(Pr )
(
)
.5 b
b f
5
_.m.
i i
d) umservat Ive heat t ransf er cuefIlelent u of the order of 2*> htu/hr ft' "r is.ilI thut ja needed to turn bark t he rising clad t eroperat ute wison if m core is recovered.
Information generated by LOCTA-R2 as a function of time f ricludes:
i a) fuel temperature, b)
Clad temperature.
c)
Steam temperuture d)
Amount of metal water reaction, e)
Volume of core melt, and f)
Total heat released to coolant.
Symbols for Equations h
llea t transf er coef ficient on other surf ace of fuel rod (13tu/hr-f t "l) lle Equivalent diameter of flow channel - (ft)
Density (1bs/ft )
n Viscosity (lbs/ft-hr) p Volumetric flow rate (ft /hr) y Ac Area of flow channel (ft )
Cp Specific heat (Btu /lb *F) k Thermal conductivity (Btu /hr-ft 'F)
T Temperature 'F Subscripts Evaluation of the property at the Laturated-vapor condition v
1 Evaluation of the property at the saturated liquid condition b
Evaluation of the-property at-the saturated bulk fluid condition Evaluation of the property at the clad surface temperature-w i
f i
6
,c-,m-,
-,.,:<v--
.~vr,7y
..--,--.I,---,--
..,-.-h,.,,--.,.-~.n.,,y-
,.cym,,e,.--,r.
,,,-,r...-.,-rv,-w,
,--.cr--,-e
CONT AINMl:NT pRl:SStJRL AND TI:MPI:RATURI: TRANS10N1 ANA!.YSIS Calculation of contalnment prc% ure and t emperat ure t ram ient s i s,oeom plished by use of t he digil lai computer code, Coro.
Tio analyti<al model is restrleted to the containment volume and utructure.
Transient phenon na within the reactor coolant system affeel containment condit ton by nivans of convertLee mass and energy transport through the pipe break.
l'or analytical rigor and convenience, the containment air-steam-water mixture is separated into two systems.
The first nystem (onsists of the air-steam phase, while the second is the water phase.
Suffictent relation-ships to describe the transient are provided by the equations of conser-vation of mass and energy as applied to each system, together with appro-priate boundary conditions.
As thermodynamic equations of state and con-d1Llons may vary during the transient, the equations have been derived for all pohsthic cases of superheated or saturated steam, and subcooled 1
or naturated water.
Switching between states is handled automatically by the code. The following are the major nsuumptions made in the analysis:
t
- 1) Discharge mass and energy flow rates through the reactor coolant system break are established from the coolant blowdown and core thermal transient analysis (described in the preceding paragraph,).
1 l
l
- 11) At the break point, the discharge flow separates into st eam and water phases. The saturated water phase is at the total contain-ment pressure, while the steam phase is at the partial pressure of the steam in the containment.
iii) Homogeneous mixing is assumed. The steam-air mixture and the water phase have uniform properties.
More specifically, thermal l
equilibrium between the air and steam is assumed. This does not l
l Imply thermal equilibrium between the steam-air mixture and the water phase.
(
7 l
i m.
_.-p.p.,_gn..yg.m.-m
__g
,.,_y,m,.,,,
_--.p.
y.
q,gg.
9
.p.,.---..-
n q
p g
.,.--ar y
tv) Air is taken an.in i de.n l ga... while compressed wat er and s t eorn tables are employed for wat er and s t eam t her mmlynaml e propertie During the t r ans ient, there is energy t r.rts f e r from the s t i ani-a l t.oW w.it e r systems to the internal struelures and etpsi pment withlsi the shell.
Provision is nade in the containment pressure trannient analysis for heat transfer through, and heat storage in, both interior and exterior walls.
hvery wall is divided into a large nutnber of nndes.
For each node, a
conservatton of energy equation expressed in finite difference form accounts for transient conduction into and out of the node as well as temperature rine of the node.
The film heat t ransfer coef ficients between the cont ain-ment contents and a wall surf ace may be specified either as a f unction of time or ternperature dif f erence.
For this study, tent valves obtained by Kolflate were used to set inner surface film coefficients for condensing steam.
Initial values of 620 and 240 Btu /hr-ft
- F were assumed for ron-densation on steel and concrete, respectively, as recommended in the propo%cd AS A Sect t on N-6. 7, Pr oponeri Des inn and Pressure Decay Reclut ternent *;.
The coefficientw were then linearly decreased with time to a value of 40 blu/hi-ft - F at the end of blowdown.
The latter value is typical of steady state heat transfer coefficients for steam condensing through a stagnant surface air layer where the ratio of air to steam is high.
The former values are reasonabic since the initial blowdown will produce large amounts of steam in the containment together with high turbulence.
Thus the ratio of air to steam will be low and a turbulent boundary layer will exist leading to a high film coefficient of heat transfer. The assumed reduction in film coefficient is conservative since, as predicted by Jacob and demonstrated by Uchida the film coef ficient will be a function of the ratio of air to steam in the containment which will not be reduced significantly until pressure reduction due to steam condensation occurs.
f 8
g
- - - - - - - - - - - - - - - - ~ - - - - - - ~ - - ~ -
provision is s.ade in the computer analysis for the effects of several engineered saf eguardr>, including internal spray, f an coolen t.,
and recir-culation of sump wat er.
The heat removal frem containent ricam-nir phase by internal spray is determined by allowing the spray unter tem-perature to rise to the steaa-air teraporature. This procedure has been experimentally verified.
Finally, hot metal surfaces not cooled by safety injection water are s.imulated as hot walls in contact with the containment stearn-air mixture.
A smn11 film heat transfer coefficient is errployed to ref1cet actual condition since these eurf aces are covered by stagnant otcam inside the reactor j
coolant system.
During the brief period between completion of the initial blowdown and addition o.
safety injection water, the transfer of heat into the contain-ment vessel by convection from the reactor coolant system insulation and by the small residual steam flow is relatively ineffective.
Consequently, cost of the residual heat generation serves to raise the temperature of the core, internals structure, and the reactor vessel. As the Zircaloy cladding temperature in any portion of the core rises above the threshold at which a significant rate of reaction occurs (1800*F), additional energy is produced and stored inside the reactor vessel. Unless the steam evolution rate exceeds the maximum possibic zirconium consumption rate established by the parabolic rate law, all steam evolved is considered to contribute to the core stored heat by reacting with zirconium, and the energy addition to the containment during this portion of the transient is a result of the burning of the hydrogen generated by the zirconium-water reaction.
e The analysis is performed on the basis that the hydrogen generated within the reactor coolant system as a result of a metal-water reaction following a loss of coolant accident will burn as it discharges from the break into the containment atmosphere.
The reaction does not start until-the clad-temperature exceeds the threshold of 1800'F for zirconium, and the hydrogen will be formed at temperatures between 1800'F and 4800'F.
(
9 l
l
Studies conducted by the U. S. liuteau of Mines indicate that hydrogen can ignite without an cxternal energy source when its temperature is above well-defined limits.
1his spontaneous ignition temperature (SIT) of hydrogen varies with the break flow velocity and the steam content.
For low velocit.ics and high vapor content the SIT is conservatively taken as 1250'F.
The hydrogen exit temperature has been evaluated for two possibic cooling rnechanisms for a formation temperature of 1800'F.
1)
The first mechanism evaluated is the heat transferred to the reactor coolant system piping.
The temperature of hydrogen as it enters the containment atmosphere is computed assuming the following 1.
The rate of hydrogen evolution is based upon the steam limited reaction which is consistent with the assuttption of no core quenching and producco the maximum reaction with the cladding and the maximum quantity of evolved hydrogen.
2.
The amount of hydrogen produced is determined by assuming that all the steam formad by boiling of the water remaining in the reactor vessel below the core reacts with the cladding.
Hence no mixing of hydrogen with water vapor has been assumed to occur.
This mechanism produces the maximum quantity of hydrogen.
3.
Hydrogen flows directly from the core, through a 50-foot length of reactor coolant hot-leg pipe before reaching the atmosphere.
4.
A conservatively low value of 300*F is assumed for the pipe wall temperature.
5.
The physical properties of hydrogen are evaluated at 1800'F.
i 10 l
I
~
6.
1he heat transfer coefficient is calculated using the Dittur.-
lioc1ter relationship.
Applying these conservative assumptions the hydrogen it, found to be cooled to 1335'r before discharge.
This temperature is 80'r above the most conservative value of the SIT and 365'r above the expected SIT.
liigher forn.ation temperatures produced on the reaction proceeds without core quenching would result in even higher exit temperatures and spontaneous combustion.
11)
The second cooling mechanism investigated is the effect of mixing with a flow of cooler steam which might by-pass the metal-water reaction.
In this case less hydrogen would be produced.
Assuming again that the hydrogen is formed at 1800*F and that steam is present at 300'r saturation conditions, it is found that a steam bypass flow of 36% will only reduce the hydrogen temperature to 1200'F (a value comparable to the conservative value of the SIT). Volume fractions of steam higher than 36% bypassing the point of reaction i
would limit the reaction rate due to dilution and cooling.
The hydrogen-oxygen reaction is the on)) significant source of energy reaching the containment from the zirconium-water reaction between the time that blowdown is complete and safety injection starts.
If My zirconium-water reaction occurs before cooling by safety injection becomes effective, only a fraction of the total zirconium available for reaction will be consumed, because the safety injection system is acutated in a very short time and the core temperature rise does not progress to a great extent. The energy released from this reaction does not ente' the containment concurrently with the blowdown energy, since cladd.;g temperature does not increase during blowdown.
An influx of safety-injection water is necessary to liberate the stored heat and safety injection cooling does not begin until after blowdown is complete.
(
11
/..
_ _ _ _