ML20085H595
| ML20085H595 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 10/27/1969 |
| From: | Neidig R SAXTON NUCLEAR EXPERIMENTAL CORP. |
| To: | |
| Shared Package | |
| ML20083L048 | List:
|
| References | |
| FOIA-91-17 NUDOCS 9110280318 | |
| Download: ML20085H595 (48) | |
Text
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DOCKLT NO. 50-146 ;
i Amendc.en t No. 7 to Chante Requect No. 32 ,
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- 1. On June 7,1966, - App 11er.nt , in responen to Divie. ion of Reactor L:ironninp.
Actter of December 1966, cubmitted for review and evaluntion a. report 1 entitled: SAXTON - LOSS OF COOLANT ACCIDENT PREVENTION AND PROTECTION.
This report included a description of plans -to -conduct a thorough -
nondestructive inspection of the prcosure-containing co:nponente and pipin'g of the reactor coolant systen and e portion of the main stcon system. Iiy letter dated July 22, 1969, Division of Reactor Licencing requested SNEC !
provite o report covering thin inspection.
- 2. Appliccut hereby submite n report entitled! SAXTON COOLANT SYSTE!! j 2NSPECTJON REPORT. This report provjdes and cycluate,, the resultc of the ~
inr.pcet ions conducted ~- the reactor coolant syrtem for the period No d cr, 1968 May, 3969.
r SAXT0!: HUCLEAR EXPERIDENTAL CORPORATION l
Ly /c/_ R. h. Uc.idfg _ _ . ,
President .
(S E A L) s i f,
Attest i
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/g R. D. Haist .
1 Secretary .
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- _ . Sworn and subccribed to before me this 27th day of Octoter, 1969.
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l (S E A L)
V7 'g e s
/s/ Char 3c3 J. Ausci -a Kotary Fuhlic
. Muh3coberg iv.thship, LcrPs County ;
911028031s 910424 My Ccn;*irsion D71 rec October.--JL,1970 PDR FOIA
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SAXTON COOLANT SYSTDI INSPECTION REPORT 4
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TABLE OF CONTENTS i.
, Section Title Page 1 SU!OiARY AND ll;IRODUCTION 1 .
1.) Summa ry I s[ 1.2 Introduction 1 2 INSPECTIONS 2 2.1 Reactor Vessel Cladding Between Closure 2 Flange and Thermal Shield 2.2 Reac+w Vessel Flange end Flunge-to-Shell Weld -
2.3 Reactor Vessel Outlet ilozzle, ID 2 2.4 neactor Vessel Studn 2 2.5 Re#ctor Vessel Top Head 3 2.6 Reactur Vessel Bottom liead 4 2.7 Steam Generator Iube Place-tc-Shell, lube Place- 4 to-Channel Head, Nozzle-ta-Channel Head Welds 2.8 Steam Generator Secondary Side 4 2.9 Steam Generator Channc' Head ID 5
2.10 Pressurizer OD, Shell-t s-Head, "ure,e Line 5 Fozzle-to Shell, and curge Line-to-Pipe Welds 2.11 Pressurizdr ID $
2.12 Reactor Coolant Pump and Casing Flange 6 2.13 Pm eter Coolant Pump Flange Eolts 6 j.
e_:p 2.14 Rea. tor Coolan Pipe 6 4
2.15 React r Cctlant Pipe Fir.titgs 6 2.16 Rea::te e c lant Pipe Weldments 8
. 2.17 Comronent Structural Supports 8 11 l
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TABLE OF CONTENTS (cont)
Su tion Title Page Appendix n Photoscaphs of Pressurizer Internal Surf aces 10 Appendix B Inspection Program Code References 18 Appendix C Visual Inspection Program 19 Appendix D Lower Core Support Barrel Inspection Report 24 on Discoloration and Discontinuity Observed on the Outlet Nozzle 1
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LIST OF ILLUSTRATIONS 4
Figure Title Page i Typical Reactor Coolant Pipe Fitting 7 2 Schematic Diagram of Reactor Coolant 9
Pipe Weldn.ents A-1 Saxton Pressurizer Interior --
Water Area 11 A-2 Saxton Pressut_zer Interior --
Water Area 11 A-3 Saxton Pressurizer 7nterior --
Water Area 12 A-4 Saxton Pressurizer Interior --
Water At.a 12 A-5 Saxton Pressurizer Interior --
Water Area 13 A-6 Saxton Pressurizer Interior Water Area 13 A-7 Saxton Pressurizer Interict --
Water Area 14 A-8 Saxton Pressurizer Interive Water Area 14 A-9 Saxton Pressurizer Interior --
Water Area 15 A-10 Saxton Pressurizer Steam - Water Interface Area 15 A-ll Saxton Pressurizer Steam - Water Iuterface Area 16 A-12 Saxton Pressurizer Steam - Water Interface Area 16 A-13 Saxton Pressurizer interic~ --
Steam Area 17 A-14 Saxton Pressurizer Intario- --
Steam Area 17 D-1 Saxton Reactor Plant Lower Core Support Barrel 35 D-2 Saxton Irternals Assembly 36 D-3 Enlargement of Vendor's Original Thotograph 37 L- 4 Photograph of Area From TV Tape 38 f D-5 I
- Saxton Lower Core Barrel Stress Distribution 39 D-6 Santon Lover Core Barrel 40 iv 4
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. SECTION 1
SUMMARY
AND INTRODUCTION e
1.1
SUMMARY
The inspections outlined in "Saxton-Loss of Coolant Accident Prevention and Protection" have been performed. Minor modifications in scope were made to ensure a more definitive inspection in some areas. The inspection revealed no defects or problems of the type that would cause concern for enntinued safe operation of the plant.
1.2 INIRODUCTION Westinghouse Nuclear Energy Systems was requested to institute an inspection puogram of the Reactor Coolant System of the Saxton Nuclear Experimental Corporation plant located at Saxton, Pennsylvania. The proposed inspection program war outlined in "Saxton - Loss of Coolant Accident Prevention and Protection," Section 2.2, page 2.2-1, titled " Inspection Program." It is planned that this inspection program will be repeated at five-year intervals.
Construction and other files on record were reviewed and utilized to make this
- inspection more relevant. All nondestructive testing was performed to the requirements of ASME boiler and Pressure Vessel Code,Section III, Appendix IX.
The Magnaflux Corporation, an independent nondestructive testing agency, was 1
emplnyed to perform a portion of these inspections. The results of the i
Magnaflux Corporation report have been incorporated into this inspection report.
i This report will follow the outline established in Table 2.2-1 and Figure 2.2-2 in "Saxton - Loss of Coolant Accident Prevention and Protection." The l
inspections were not performed in the order presented in the outline, but on I
a plant availability basis.
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. SECTION 2 INSPECTl*NS 2.1 REACTOR VESSEL CLADDING BETVEEN CLOSURE FLANGE AND THERMAL SHIELD (REMOTE VISUAL)
Initial inspection was made with binoculars and a television camera. Selected accessible areas were tnen inspected with a 12X glass. No ceficiencies were detected in this area by the above visual methods of inspection.
2.2 REACTOR VESSEL FLANGE AND FLANGE-TO-SHELL WELD (VISUAL)
Visual inspection of the reactor vessel flange revealed no unacceptable cond-itions.
2.3 REACTOR VESSEL OUTLET N0ZZLE, ID (VISUAL)
Since the lower core suppert barrel was not removed, inspection of the outlet nozzle ID was difficult. This inspection was mada with the TV camera. No un-acceptable conditions were detected. During visual inspection-in this area, a surface discaloration was observed on the flange support of the lower core support barrel. A detailed dercription of the discoloration is given in Appendix D.
2.4 REACTOR VESSEL STUDS (VISUAL AND ULTRASONIC OR PENETRANT)
The 36 vessel studs were ulti asonically inspected and-nine studs were fluore-scent-magnetic particle inspected on June 21, 1967 by the Magnaflux Corporation.
On review of the inspection program, the penetrant ,e ultrasonic planned test was changed to fluorescent-magnetic particle inspection since experience at-other p: 7ts has proved it to be a superior method of inspecting reactor vessel studs and nuts.
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Visual inspection of the studs and nuts revealed a rusty surface conditior..
The studs and nuts were cicaned by grit-blasting preparatory to inspection.
After cleaning, fluorescentwnagnetic particle inspection was performed. All 4
, stuos and nuts sincluding spares) were acceptable by fluorescent-magnetic particle inspection.
, After magnetic particle inspection, a further inspection was made of the surface condition of the threaded sections of the studs and nuts as it would affect installation and removal. The surface condition of the threads was adjudged unsatisfactory for continued use without corrective action because of the possibility of galling and seizure. Westinghouse Materials Engineerir...
Westinghouse Research and Development, and Babcock and Wilcox were consulted for recommendations. Babcock and Wilcox recommended a procedure developed by A'len Aircraf t Products for conditioning tne studa and nuts. The procedure calls for phosphate and electrofilm conditioning of the reacto vessel studs and nuts. This preventative maintenance was performed and will minimize'the possibility of thread seizure on subsequent removal and assembly of these studs.
?5 REACTOR VESSEL TOP HEAD (V1SUAL AND PENETRANT)
The reactor vessel head-to-flange weld was cleaned to ;emove paint and rust l
p-ior to inspection. Visual inspections of the entire veld revealed no anomalies. Four six-inch ler3ths of the veld were selected for liquid pene-trant inspection. The three nead-lifting lug welds contact the head-to-flange weld, so the areas selected included both velds. The head-to-flange veld was found acceptable.
Linear liquid penetrant indications (intermittently disposed -- totaling-approximately one-half inch)- were found in the lug weld area. On further .s surface conditioning and penetrant re-inspection, these indications were-found to be 1/8-inch away from the. pressure containing head-to-flange weld ,
and were adjudged irrelevant.
The internal clad surface of the head wcs visually inspected. Particular attention was directed ta the head-to-flange _ eld region.
A manufacturer's repair was found in one trea of the cladding. This area and other selected areas vere examined with e 12X glass and found acceptable.
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2.6 REACTOR VESSEL BOTTOM HEAD (REMOTE VISUAL)
The botton head was examined with a television camera and binoculars. Illumi-
, nation was provided by two 1000-watt and two 500-watt immersible lamps. Video tape recordings were made of the entire bottom head. No foreign material or irregularities were observed.
The thermal shield supports, bottom core support casting and tie rods were inspected with binoculars, No irregularities were observed.
2.7 STEAM GENERATOR TUBE PLATE-TO-SHELL. TUBE DLATE-TO-CHANNEL HEAD, N0ZZLE-T0-CHANSEL HEAD WELDS (V1SUAL AND PENETR/NT)
External visual inspections were performed on the steam generator, tube plate-to-head weld, tube plate-to-shell weld, outlet nozzle-to-channel head weld, outlet nozzle-to-main coolat piping veld. The welds were found acceptable.
1 Four six-inch long areas of the steam generator tube plate-to-head and tube plate-to-shell welds were selected for liquid penetrant inspection. The welds were sound acceptable.
, 2.8 STEAM GENERATOR SECONDARY SIDE (VISUAL)
The manway and the two handhole covers were removed for a visual inspection of the secondary cide internals. The surfaces ofsche shell, tubes, fittings, and the moisture separator were thinly coated with a soft blackish deposit. There was no evidence of pitting of the internal surface of the shell or any of the internal parts. All int.ernal parts, fasteners and locking wires were intact and in good condition. The steam purifier assembly was found to be in good condi 'nn. In the lower section, as observed through the handhole openings, there was no visible evidence of erosic cr pitting of the tubes.- No unusual deposits were observea on th - tube. nlat.e .
l The top head-to-shell weld na visually examined in its entirety. An arc strike was-found on each side of the weld on the east side of the steam generator. These were remove 4 by lient grinding, less than 0.010 inch of
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material-being removed. These areas plus four selected six-inch segments of the weld were liquid-penetrant inspected and found acceptable.
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2.9 STEAM GENERATOR CHANNEL HEAD ID (VIShL)
The entire internal area of the channel head was visually examined. The tube-
, to-tube plate welds, the divider plate, and the cladding (particularly the cladding over the tube plate-to-channel head weld) were visually examined.
No foreign material was found in the channel head. The steam generator channel head internals are acceptable.
2.10 PRESSURIZER OD, SHELL-TO-HEAD, SURGE LINE N0ZZLE-TO-SHELL, AND SURGE -
LINE-TO-PIPE WELDS (VISUAL AND PENETRANT)
Both shell-to-head welds (top and bottom), surge line nozzle-to-shell weld, and surge line nozzle-to-pipe weld were visually inspected. In both of the shell-to-head welds, four selected six-inch lengths were liquid-penetrant ,
inspected. All welds were found accentable.
2,11 PRESSURIZER ID (VISUAL AND PHOTOGRAPHIC)
The heater bundle and spray line were removed to make the internal inspection of the pressurizer. Due to the high radiation levels associated with the pressurizer, a remotely operated camera was used to obtain color photographs of the pressurizer internals. Because of interference with numerous thermo-couple probes, only a few representative photographs were obtained in each of the three areas of the pressurizer (water area, water-steam interface area and steam area). These photographs are shown in Appendix B.
. Limite.d visual inspection was made through the heater bundle opening and the top flange opening (which is slightly less than four inches in diameter). The internal cladding was found to be in excellent condition. The heater bundle war visually inspected-and found to be in acceptable condition.
During the visual inspection, it was ascertained that ,a spray nozzle was not g attached to the spray line. Since a nozzle could not be found inside the pressurizer and since it is physically impossible for a nozzle to exit the pressuricer, it was concluded that a nozzle and never been installed. Further-more, examination of the spray line showed no evidence of thread engagement-or tack welds. A spray nazzle was obtained and installed.
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2.12 REACTOR COOLANT PUMP AND CASING FLANGE (VISUAL)
On March 11, 1968, as part of the reactor coolant pump inspection, inspections were made of the socket welds, the head-to-bend welds for the cooling coils, and the stator cap-to-vent pipe welds. These welds were found acceptable.
. A visual inspection of the pump volute and casing flange revealed no anomalies.
2.13 REACTOR COOLANT PUMP FLANGE BOLTS (VISUAL AND DIMENSIONAL)
On March 11, 1968, the flange bolts and bolt holes were visually inspected and found ecceptable. The flange bolts were measured for elongation and found acceptable.
2.14 REACTOR COOLANT PIPE (VISUAL, PENETRANT, AND RADIOGRAPHIC)
In 1963 and 1965, the 30-degree bend on the 12-inch reactor coolant pipe at the coolant pump outlet was inspected by liquid-penetrant and radiographic techniques. The results or' these inspections were satisf actory.
l A strain gage (one of several used during original plant checkout) welded to the 12-inch pipe, was removed and the veld area surf ace-coelitioned prior to liquid-penetrant inspection. Fusion line indications cere found by the in-spection. The area was further surface-conditioned (Itas than 0.030 inch of I
metal being removed) and reinspected. This area was found acceptable.
l j The 30-degree bend area and the entire pn=p-rn-pipe weld was visually if quid-l penetrant, and radiographically inspected. All conditions were found accep-table.
1.15 REACTOR COOLANT PIPE FITTINGS (VISUAL AND PENETRANT)
The pipe-to-fitting welds, and fitting-to-reactor coolant pipe welds (Figure 1) on the pressurizer surge line, the auxiliary system return and the pressurizer spray line were visually and-liquid-penetrant inspected.
Liquid penetrant indications were found in both welds on the auxiliary system return line. The fitting-to-reactor coolant pipe veld had numerous small indications aligned, with what appeared to be grinding marks,
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perpendicular to the reactor coolant pipe. Surface conditiening of this area (less than 0.005 inch of metal was removed) was followed by a ratisfactory liquid penetrant inspection.
The pipe-to-fitting wcld had indications on the veld and numerous linear indications at both fusion lines intermittently around the weld. Surface conditioning (less than 0.032 inch of metal was removed) was followed by a satis f actory liquid-penetrant inspection.
2.16 Rr. ACTOR COOLANT PIPE k'ELDMENTS (VISUAL, PENETRANT, AND RADIOGRAPHIC)
The steam generator outlet no::le weld, outlet nozzle-to-pipe welds, reactor coolant pump-to-pipe and pipe-to-elbow welds were visually and liquid-penetrant inspected (Fi Fure 2). All welds were found acceptable.
2.17 COMPONENT STRUCTURAL SUPPORTS (VISUAL)
Support brackets welded to the steam generator, pressurizer and reactor cool-ant pump and their attachments overhead were visaally inspected. in addition, the turnbuckles on the steam senerator support rods were visually examined and two turnbuckles were liquid penetrant inspected. All inspection r(sults were acceptable.
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, APPENDIX A PHOTOGRAPHS OF PRESSURIZER INTERNAL SURFACES Representative areas of the steam, steam-water, and water surfaces of the 1 l
pressurizer cladding were photographed and are included in this appendix.
The whitish deposits noted on the instrument probes and on the wall of the I pressurizer are boric acid cyrstals.
Photographs taken of the various areas are as follows:
Area Photograph Number water - 1, 2, 3, 4, 5, 6, 7, 8, 9, steam-water interface - 10, 11, 12 steam - 13, 14 The stained rag shown in Photograph 13 was used during fit-up of piping to the pressurf rer and has been subsequently removed.
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s Figure A-14. Saxton Pressu-izer Interior -
Sten::: Area a
h 17
-# "Ja_. s. _sa & 4, _se'A._ _x 3m--4 a a4 , s A _se.. A,,g,_. ,a ,,,e3 _..y,4 4 a_. 4 :,. g.g .s.,2, ,.
i
. APPENDIX B INSPECTION PROGRAM CODE REFEndNCES
- 1. All inspections were performed by or-under the direction of personnel ,
qualified to the ASME Boiler and Pressure Vessel Code, Appendix IX,Section III.
- 2. Detailed records of all inspections were prepared and retained in accor-dance with ASME Boiler and Pressure Vessel Code, Appendix 1X,Section III, paragraph 225. This includes written procedures for each type of in-spection and for the data developed in each and all inspections. Records are in sufficieric detail that a reinspection can be made' independently at any future date on the basis of the recorded information. Also, the re-i corded information is satisfactory for direct submittal to regulatory agencies or other responsible authorities.
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. . _ . . .. _ . . _ - . . - - - . . . - _ - . . ~..- ._. - - - . _ . . _ - - - - -
h
~
APPENDIX C VISUAL INSPECTION PROGRAM
- 1. VISUAL INSPECTION PROCEDURE - Saxton Plant 1966-69 A visual inspection is employed to gain detailed information as to the genersi condition of the part, materials, velds. components or surface, etc., on such conditions as corrosion, erosion, wear cracks, distortion, alignment, movement or any other type of damage or injury.
Personnel perf orming visual inspections are subjected to a physical examination to assure natural or corrected near distance acuity such that the inspector is capable of reading J-l letters on a stendard Jaeger's test type chart for near vision or equivalent test typr. Color. vision and/or additional medical requirements as applicable are also considered.
The examination is conducted within one year of the time the inspection is made. This examination is equivalent to the requirements of "Recom-mended Practice No. SNT-TC-1A" issued by the American Society for Non-destructive Testing, Evanston, Illinois,
- a. Direct Visual Examinatior.
Direct visual examination may be performed then access is suf ficient -
to place the eye within 24 inches of the surface to be examined-and at an angle no less than 30 degrees with the surface to be examined.
Mirrors may be used to irnre.ve the angle of vision. Lighting, in addition to the general area lighting, shall be provided to illumi-nate the area to be examined at right angles and at oblique angles to -
expose cracks or evidence of corrosson or erosion. Mirrors and lighting must be designed to avoid breakage.
1 19
- - - , - , - - - r w- i m._,-+,,%.r..,,,,,m.yg. ,_ .,,,w- -, , , . . , s,.e,,- ,.w,-->. , c- ,,-r.-,+-p,
i
- b. Remote Visual Examination Remote visual examination may be substituted for direct visual examination when, for any reason, it is desirable. Remote visual examination may include visual aids such as telescopes, periscopes, borescopes, fiber optics or TV cameras and monitoring systems, with i
or without attachments for permanent recording.
Such systems shall demonstrate the ability to provide a resolution at least equivalent to that obtainable by direct visual observation.
Mirrors, movable lights or rotating optics, or any combination thereef, may be employed to display cracks, surface scratches. or evidence of corrosion, erosion, alignment, or movement.
- c. Replication Surface replication methods shall be considered acceptable provided the surface resolution is at least equivalent to that obtainable by the visual observation.
- d. Photography Photoga phy shall be employed to document observed conditions to the maximum extent possible. This can serve as a vivid description of the actual condition as well as a record for future reference, as may be required.
- e. Special or Questionable Areas When questionable areas ara ascertained, additional detailed inspec-tion shall be applied. This may consist of using a 10X glass or other visual aid, if adjudged necessary, applicable nondestructive testing
, may be applied. Some. areas may require surface conditioning to re-veal relevancy of indications. Appropriate engineering review will
. be made in all questionable areas found in the visual inspections.
20
- . . . . . . . . _n -
- 2. VISUAL INSPECTION SCOPE
- a. Components and Piping Welds - All welds to be visually inspected shall be:
- 1) Free from oxide, scale, paint, craters, slag, porosity, cracks,
, incomplete penetration and lack of fusion.
- 2) Free from surface markings resulting from mishandling, punchings, etc., and anf permanent marking not in accordance with the ASME Code.
- 3) Examined for surface irregularities. The weld surface shall merge smoothly and gradually into the base material. ' Butt welded joints may be flush with the base material or may have a re.asonably uniform crown not to exceed the following thicknesses:
Base Material Thickness, Thickness of Reinfercement, inches Inches Up to 1/2 inch 1/16 over 1/2 inch to 1 3/32 Over 1 to 2 1/8 Over 2 5/32
- 4) Free f rom abrupt ridges to valleys.
- 5) Aligned to meet requirements of ASME Boiler and Pressure Vessel Code, Appendix IX,Section III, paragraph N-525.
l 6) Examined _ to ensure that when dif ferent base metal thicknesses i are joined, the finished joint shall have a taper of 1.4 between the thick and thin section.
l
- 7) In suitable condition to.parform any and all required non-t
, destructive testing.
1
. NOTE: a) Other optical aids such as 10X magnification may be employed to verify condition of questionable areas.
b) Areas for other or additionally required nondestructive testing shall be identified by this visual inspection.
l
- 21
- s-
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- c. K Q:.:( q*c n
!Lk 1b. :-Reactor Vessel - Other areas shall be' inspected by TV cameras, y ; binoculars, mirrors or-other optical alds. Areas will includes .;
7c -
.1) 'Claddington the ID of the closure head s
? ,
2);= Cladding from the vessel flange down to the top of the thermal
]
~
- . shield; -i
~
- 3) Bottom head through the holes in the lower core support plate.
bW
' , i
- c. ESteam Generator r
< ljLSecondarySideInspection-HandholeandManway'CoverRemoval
- a) . Examine for corrosion and/or erosion and note unusual i
= conditions in sufficient detail for review and appropriate- ;
- action now and in tbt future.
- b) All internal members shall'be checked for secure attachment.
c): Check' separator drain lines to_make certain that braces are' intact. ,
'e 'd) LCheck centritix septrator_for signs of excessive corrosion
- and erosion.
.e)"/ Check tolmake certain that cl1 fastener locking wires are intact.
f . -
f); Trom handhole inspect-back side of the tube pla'e for mud
, . ' accumulation and_. tube' surface. deposits. .p 12)lPrimart ' Side-
. a)! Examine for corrosion and/or erosion and note unusual
- conditions:in suf ficient- detail. for review and appropriate -
'_"# action'for now.and in the future, b). Visbr.1 examination.-of' tube welds.
c, ; c) LVisual examination of divider plate welds.
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d.. Pressurizer
-Visually examine internals, cladding, penetrctions, weld areas and neater bundle.- Note condition of cladding in steam phase, water
- phase and water-to-steam phase interf ace to determine if there has been significant corrosion or cracking of cladding. Photographs will be taken of representative and significant areas.-. 1hese ob-ser ations and photographs are to serve as a reference for future i
et. amina tio ns .
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.- O e
23 '
4 APPENDIX D LOWER CORE SUPPORT BARREL INSPECTION
. REPORT ON DISCOLORATION AND DISCONTINUITY OBSERVED ON THE OU
- 1. OBSERVATION '
During visual inspection of the lower core support- barrel some surface >
discoloration was observed on the flange support (Figure D-1 Item 7) of of the lower core support barrel in the downstream direction from the weld that joins the flange support (Item 7) to the lower barrel (Item 4).
The discoloration was parallel to the water flow tapering off at the down-stream end.
It initiated at the weld which joins the flange support .to the lower barrel.-
Subsequent inspection by television camera and binoculars confirmed there p
was a surface discontinuity at the origin of the discoloration. The area was television tape recorded (Figure D-4). .The sise of the discontinuity was estimated to be 1/4-inch wide, 4-1/4 inches long, and on the order of 1/16-inch deep.
It was oriented in the d' irection of the veld, i.e. ,
perpendicular to the direction of flow.
The surface within the dis-continuity was not abrupt or indicative of any stress-related structural defect.
- 2. OBJECTIVES OF EVALUATION Since there was no immediate explanation of these observations, an in-
[ vestigation was carried out with the- following objectives:
1, An engineering evaluation and analysis of the conditior as to its nature and possible causes.
2.
An evaluation of the possiblility of- further degradation.
3.
Assuming the nozzle wall could become penetrated, would the additional by-pass flow create a potential hazard?
24
. _ __ _ _ _ _ . . _ _ . . . _ _ _ . _ _ _ _ ..~ _ . _ . . _ _ _ . _ _ __ _ _ _ . - . _ _ _ _ _ . _
- 3. SCOPE OF EVALUATION It is clear that the observed discontinuity was either present when the reactor was first commissioned, or that some removal of metal occurred subsequently.
The investigation was therefore directed at the following areas,
- a. A check on original records to determine if there was evidence of the existence of the discontinuity prior to original commissioning.
- b. A list of possible mechanisms which could account for the appearance of the discontinuity during service.
- c. An evaluation of each mechanism to see if it would account for the- ;
observations, and if so, what prediction could be made as to the i
future growth of the discontinuity,
- d. An evaluation of the stresses existing in the vicinity of the dis-continuity.
- e. An evaluation of the additional by-pass flow resulting from a possible penetration of the nozzle wall at the discontinuity.
4 CHECK OF ORIGINAL RECORDS This check disclosed the existence of a photograph (Figure D-3) showing the -Lower Core -Support Barrel during fit- up of the -internals in the manu-f ac turer 's plant . An enlargement of_the nozzle area is shown in Figure D-3 and may be compared with Figure D-4 The comparison shows that both the-
- discontinuity and some features of the -discolored- area (lines parallel to
- flow) .can be observed in both pictures. The discontinuity extends round the circumference of'the flange support (Item 7), to the interpenetration
- with the conical section of the barrel (Item 3), at which point a mis--
alignment can be observed in both Figure D-3 and Figure D-4; The dis-continuity appears to have been caused by a misalignment of the parts
. (Items 3, 4 and'7) during manufacture. It appears that a fillet weld was deposited in this area to smooth out the profile. No immediate explanation 25
4 can be given of the lines parallel to flow in the discolored area, but they may be grinding marks which did not clean up during machining of
. the nozzle.
, 5. EVALUATION OF MECRANISMS WHICH COULD ACCOUNT FOR THE DISCONTINUITY DURING SERVICE The uechanisms considered were:
- a. Mechanical interaction,
- b. Corrosion or erosion.
Mechanical interaction was ruled out because of the absence of any con-ceivable mechanism whereby any other part could have come in contact with the discontinuity during service.
The possible corrosion-erosion mechanisms are reviewed below:
5.1 Uniform Corrosion (Assuming that the materials compliti with the assembly drawing, i.e.,
. ASTM A240 Type 304 stainless steel was welded with Type 308 stainless steel weld metal .) For Type 304 stainless steel in reactor purity water, Wanklyn and Jones report an average corrosion penetration of 0.06 mils per year and a maximum of 0.12 mils per year. Variations of composition -
among the austenitic grades of stainless steels had little effect on the penetration depth, therefore the same value is applicable to the Type 308 weld, under the assumptioc of uniform corrosion (galvanic ef f ects are considerel below). Variation in water flow rate from nearly stagnant to 30 feet per second (fps), variations in surf ace condition, variations of the water pH between 7 and ll, additions of dissolved hydrogen or oxygen, water temperatures of up to 572*F, and additions of up to 1500 ppm of boric acid did not significantly modify the corrosion rate. For compari-son, the Saxton conditions at the area of interest are: temperature is about 510*F, flow rate is estimated as 20 fps, pH (cold) is 6.0, boron is 0-1600 ppm - boric acid equivalent is 0-9120 ppm.[3'1 '
On the basis of the experimental results for 1500 ppm of boric acid, it is estimated 26
I on a conservative basis that the actual boric acid concentration in the ,
- Saxton coolant would result in a rate of corrosion not exceeding 0.24 mils per year.
Neutron irradiation (in-pile) has been found to enhance the corrosion of
, Type 304 stainless steel clad by a factor of about two, in experiments using nitric acid as a corroding medium.(4) Since the outlet nozzle-is not subjected to the intense irradiation as experienced by fuel rods, the effect of neutron irradiation is negligible in this case.
In summary, the expected normal uniform corrosion of the specified materials is conservatively estimated as 0.24 mils per year under Saxton conditions and'is negligible for. time at temperature experienced thus far by the Saxton Reactor. A time of 18,000 EFP* hours plus a very conserva-tive estimate of at least an equivalent time period at temperature for training and physics' purposes would yield about 1.1 mils corrosion. This mechanism does not therefore account for the observed discontinuity.
5.2 Weld Metal-Contaminants T
The only. veld metal contaminants that could possibly occur would be slag formed-during the welding operation and, less likely, sulfur in the electrode' forming ****1-mulfidea.(5)
In the case of slag, experience [5] has shown that for a situation where-sufficient water flow exists (as in Saxton), corrosion occurs An the area of original slag deposition cnd the corroded material is eroded away.
Corrosion continues until the slag is removed. Such a situation could occur as the result of poor welding practice, but the corrosion rate would be expected to remain relatively constant. Any. future corrosion pene-tration, assuming slag contamination throughout the weld, can be calcu-
, lated - from the present rate of assumed corrosion, as was done in Section 5.1 above. "
- Effective Full Power 4
27 1 4
__. , - . . . _ . _ . . . , . . _ . , . . . ~ _ , . . , _ , , _..m . . . . . _ - -,.-
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In the case of sulfides occurring as a result of sulfur impurities in ;
the weld material, the expected course of corrosion is similar to that described for slag, since the sulfide particles would cause preferential corrosion ultimately leading to removal of the impurity.
In either case, it is normally expected that such contaminants would oc;ur over the entire weld length and corrosion would not be localized as shown in the photograph (Figure D-4).
In summary, weld contaminants could cause the supposed a;osion if it is very conservatively assumed that the corrosion conditiens, particularly water flow rates, are greater in the area shown, result.ing in preferen-tial removal of corrosion products and localized, faster corrosion rates than in the rest of the veld. In this case, it is expected that the rate of corrosion would be constant.
5.3 Gelvanic Corrosion Two aspects of galvanic corrosion are possible in this case. in the first,
. it is assumed that the specified weld metal (Type 308) is dif ferer.t in composition from the joined material (Type 304) and therefore is theore-
, tically subject to galvanic corrosion due to a difference in elcetrochemi-cal potential. However, noted above, this actual e.se has been studied and the galvanic corrosion effects are negligible upon an already neg-ligible overall corrosion rate. It is expected that the same results would be obtained for any other grade of stainless steel which might hate been mistakenly used for the weld metal.
The second aspect considers weld metal other than stainless steel. The worst possible case is considered to be that where a carbon steel veld metal was mistakenly used. An examination of the galvanic series of metals and alloys shows that carban steel is decidedly anodic and hence corrodable when coupled to stainless steel. Therefore, one can theoretically expect preferential corrosion of the carbon steel, es-l pecially since the area of stainless steel is large by comparison. In practice, however, Wanklyn reports that galvanic couples of mild steel l to stainless steel have little effect on accelerating corrosion of the mild steel.
28
4 Assuming, first of all, that the above experimental finding is correct, one can theref ore consider the hypothesired mild steel as corroding alone.
According to Wanklyn,{1l corrosion rates for mild steel are roughly ten times greater than for stainless steels under the conditions described in Section 5.1., Uniform Corrosion. This amounts to about 2.4 mils
. corrosion per year. Conservatively assuming the time at operating temperatures as twice the EFF hours, the expected corrosion is about 11 mils compared with the stated observation of 1/6" or 62 mils.
No information is availabe to justify a future corrosion rate in excess of that already hypothetically established.
Secondly, if the experimental report of galvanic efetts of stainless-to-mild steel couples is assumed to be incorrect, a corrosion depth greater than that for uniform corrosion of rild steel can be assumed. There are no data to expect a rate greater than is already assumed from the depth of the discontinuity. For this situation, it should be kept in mind that two " ifs" are necessary: "if" a carbon steel electrode was mistakenly use3 and "if" Wanklyn's reported experimental findings are not correct.
In summary, galvanic corrosion is a possible cause of the observed dio-continuity only if one assumes two separate errors, and the expectation of continusa corrosion in excess of the established rate is not justifi-able from the literature. In addition, it would be expected that this type of corrosion would occur over the entire length of weld exposed to the water, rather than in the localized manner shown in the photographs.
This is contrary to the observed condition.
5.4 Stress Corrosion Stress corrosion must be considered in any evaluation of corrosion of stainless steels. In reactor waters contaminated with chlorides, oxygen must be present for stress-corrosion cracking to occur.
A check of the Saxton water chemistry I shows that the hot water chloride and oxygen contents are lest than 0.005 ppm, which represents the limits of detectability for each species. Conservatively assuming that at least 0.005 ppm of each species actually exists in the Saxton coolant 29
l l .
water, and comparing with a curve of reactor water oxygen and chloride content as related to stress corrosion cracking under intermittent wetting (a condition whereby chlorides can be concentrated), it is seen that the chloride content is one to two orders of magnitude and the oxygen content is at least two orders of magnitude too low for chloride
, stress corrosion cracking to occur.
a Another possible source of intergranular stress corrosion is from caustic solution such as the LiOH used for pH control in Saxton. Berry lI reviews work which shows that several thousands ppm of caustic solution are re-quired for corrosion of stainless steel. PementIO) found a threshold of 0.1 M lithium hydroxide, or less, at 450-615'T for f ailures of stainless steel. This amounts to about 1470 ppm of LiOH. The Li content in Saxton coolant is 0.1-0.2 ppmE33, and therefore LiOH is highly unlikely to be a source of corrosion.
In summary, stress corrosion is highly unlikely on the basis of the ex-cellent Saxton water chemistry. Recourse to the assumed existing corros-ion rate is necessary if one speculates that the chemistry or literature data are incorrect, but in this case there is no reason to believe that the rate of corrosion would increase w-ith time.
5.5 Corrosion - Summary
- No corrosion-erosion mechanism examined will account for the observed "
discontinuity unless some other unfavorable assumption is made. However, if it is assumed that the discontinuity was in fact caused by some corro-sion-erosion mechanism, there is no reason to suppose that the rate will increase over that implied by the known size of the discontinuity. The thickness of the nozzle is 1/2 inch. If the assumption is made that corro-sion could occur from both sides, then a f urther 3/16" of corrosion (measured from one side) would have to occur before the nozzle wall is
, penetrated. At the assumed rate of corrosion, this would not occur in the expected lifetime of the plant.
30 I
. . ____m .m_ ~ . _ , . . _ . _ -
- 6. STRESS ANALYSIS The structure has been. treated as a continuous, constant thickness shell.
The shell is a cylinder with a conic reduction to a smaller cylinder and a reinforcing ring at each end of the shell.
The computer program " Seal Shell 2" has been used to determine the basic shell stresses.I1'b in the region of penetration the basic stress is essentially a uniaxial hoop stress of 300 psi. A resolution of all the 4 component stresses on a Von Mises criteria or on an Ltahedral stress basis gives a resolved stress of 350 psi.
Because of the complex nature of the nozzle cotstruction, no accurate stress concentration factor is.available. ..A factor of 5 has, there-fore, been used.
The basic stress is of the order of 2% of the material yield stress and is compressive. Allowing a stress concentration of 5, the stress in the nozzle area is still less than 10% of the material yield-
. strength at temperature.
On the basis _of such low steady state stresses, it has been concluded that there is. adequate margin for safe operation of the reactor.
6.1 Input for Seal Shell 2 Calculation Loading Conditions Temperature 510'F.
Vessel Differential Pressure 11.3 psi Core Differential Pressure 4.1 psi Self Weights
=-
-Lower Core-Barrel - 2750-n -Baffle Assembly = 1230
- Core. Support Assembly = 1750
, 21-Fuel Assembly- n. 2310 '
- 21. Dummy Assembly .
=. 1210 TOTAL 9.250 lbs.
l I
31 l
Pressure Drop over Core =
4.1 psi 41" 1/D core area = 1320 in2 Lift force = 5,420 lbs.
. Load on Core Barrel a 3,830 lbs, axial Materials Barrel ASTM 240 T 304 Grade 5 Electrode ASTM 298 E 308 Specified minimum yield 30,000 psi Specified ultimate strength 75,000 psi _
Yield strength at temperature 18,200 psi 6.2 Basis for Estimation of Stress Concentration Factor From " Seal Shell 2" Printout by inspection, the stress field across the penetration at mid-plane is effectively uniaxial and constant at 350 psi (Figure D-5).
Points of interest are at 90' and 135' as noted (discontinuity lies
- in this region) 1.e., a nozzle type reinforcement in uniaxial stress.
(Figure D-6)
Q%r m.
-w/+-t
,' _d
, [ ['; d/D =
.259 1 = .67 w/wh (,..u A T a a_' t = .045 _T_ = .017 d D D
Y Reinforcement Factor = .203 Penetration Angle = 42.4* Lateral
- 1. From the reference by R. T. Rose the stress concentration factor for a completely unreinforced nozzle ( p = -f) f or -h- = 0. 017 and and d/D = 0.259 is given as 5.0.
32
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The nozzle actually has 20% reinforcement therefore the i factor 5.0 is high. l
- 2. From the ASME Code 68 the recommended factor is 2.6.
For the lateral connection of a cylinder at an angic the stress concentration index = K {1 + (tan () 4/3) where K = inside stress index.
In this case K = 1.0 tan c s 1.0 and therefore stress index a 2.0.
The stress concentration factor is difficult to estimate for this case since there are no directly applicable references.
A minimum value of 2.5 and a maximum value of 5.0 for stress con-centration appears to be a valid range. This range is based on judgement from the list of references given.
The resultant maximum stresses in this region of the nozzi, are in the range 825 psi to 1750 psi.
- 7. EVALL'ATION OF BY-PASS FLOW 9
The only potentially serious consequence of a penetration of the nozzle wall is the increase in flow of primary coolant bypassing the core. Cal-culations have been made which show that a bypass hole in the nozzle would leak flow at the- rate of 1 percent of the total system flow per square inch of leakage area. The Saxton Core III design values of 7880 GPM total system flov and 15% core bypass flow contain at least 5% con-servatism, i.e., the-design value of heat transfer flow in the core at 0.85 x 7880 GPM has at-least 5% design margin. Therefore, the 2
core design is not jeopardized by a 5 in hole in the outlet nozzle.
, 8. CONCLUSIONS This evaluation indicates that the discontinuity occurred in manufacture.
. No mechanism has been identified to account for the appearance of the 33 s
s
- _ . . . _ _ ., - - . - ~ _
discontinuity in service. Ilowever, even if these conclusions are in-correct, the evaluation demonstrates that no hazardous condition could develop during the plant lifetime.
S 9
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REFERENCES References 9 through 20 if not specifically referenced in the text are general background references for the Mechanical Analysis section this report.
1 Wanklyn, J. N., and Jones , P. J. , "The Aqueous Corrosion of Reactor ,
Metals", J. of Nucl. Materials 6, No. 3 (1962) 291-329.
- 2. R. Stansfield, personal communication, 6/10/69.
- 3. R. Swift, SNEC, personal communication, 6/16/69.
4 Duncan, R. N., " Stainless Steel Failuta investigation Program" Final Summary Report , GEAP-5530, February -1968.
- 5. Leo-Marti-Balaguer,.L., personnal communications, 6/11/69 and 6/16/69.
- 6. Metals Handbook, 1948 edition, p. 559.
7 Berry, W. E. "Some Facts About Stress Corrosion of Austenitic Stainless Steels in Reactor Systems", Reactor Materials 7, (1964-1965), p. 4, 8.- Pement, D. C., Reactor Chemistry and Plant Materials. WAPD-BT-16, '
December 1959.-
- 9. Taylor-and Lind, ' TAM 270 University lilinois'.
10 - Hardenbergh and Zamrik, " Experimental Investigation of Stresses in Nozzles in Cylindrical Pressure Vessels".
- 11. 'Stepanek, " Stress Concentrations in Nozzle Ring of a Pressure Vessel".
Nuclear Structural Engineer 2 (1965).
Rose, " Stress Analysis on Nozzles in Thin Walled Cylindrical Pressure
~
12.
. -Vessels".
4
- 13. Waters, " Stress Near a Cylindrical Outlet in a Spherical Vessel".
14 . Lind, " Estimation of Elastic Stress Concentration of a Nozzle in Spherical Pressure Vessel".
41 l
1
i
- 15. Clari and Gill, "Effect of Diameter / Thickness Ratio of Flush Nozzles in Cylindrical Pressure Vessels".
- 16. Lechie and Penny, " Stress Concentration Factors for Stresses at Nozzle
. Intersections in Pressure Vessel".
- 17. Rose, Design Method for Pressure Vessel Nozzles, Eng. , June 20, 1962.
- 18. Kitching and Peckins, " Stress Analysis of Rim Reinforced Openings in Pressure Vessels".
- 19. Friedrich, C. M. " Seal Shell 2, A Computer Program f or the Stress Analysis of a Thick Shell of Revolution with Axisymmetric Pressures.
Temperatures and Distributed Loads", WAPD *Di-398, Westinghouse Bettis Atomic Power Laboratory, Pittsburgh, Penna. (August 1963).
- 20. Kraus, H., "A Review and Evaluation of Computer Programs for the Analysis of Stresses in Pressure Vessels", Welding Research Council Bulletin No. 108 t
- 9 t
42 l
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