ML19340A529
| ML19340A529 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 07/03/1961 |
| From: | Alexander A, Brandt F, Craig W GENERAL ELECTRIC CO. |
| To: | |
| References | |
| GEAP-3753, R61APE61, NUDOCS 8008130154 | |
| Download: ML19340A529 (38) | |
Text
GEAP-3753 i
,f R61 APE 61 k
5a.)g' +A,5,
n
{p ik. /*/f',
Ab l d'#
g,,-
6 se INSPECTION OF DRESDEN REACTOR VESSEL AND INTERNAIS I
By A. J. Alexander F. A. Brandt f,
W. D. Craig I
July 3, 1961 1
l cc c
J i
GENERAL ELECTRIC TOMIC POWER EQUIPMENT DEPARTMENT l / (O 4s\\'
'/
\\~
Occun p
OCTso;**['o SAN JOSE LIFORNIA
)
j zT A
REGl,4 TOR CEl.'3L flfs.
i
-,,,,, # T o @ M N 00
~
lisyl@&Y 00Em VECOPY
FOR ust QF O E EupLOYEEs ONLY GENER AL $ ELECTRIC ATOMIC POWER EQUIPMENT DEPARTMENT TECHNICAL INFORMATION SERIES TITLE PAaC AuTNon susaEc7 cLAssarICATION N O.
"Ol)O A.J. Alexander Reactor Pressure Vessel F.A.Brandt Inspection
^Iuly3,1961 W.D.Craiz TITLE INSPECTION OF DRESDEN REACTOR VESSEL l.ND Ih7ERNALS t
The inspection of the Dr-sder reactor pressure assTeac7 vessel by APED engineers during the Dec. 1960-May 1961 shut-down is described. Direct visual and periscope viewing of the reactor top head, outlet and inlet cozzles, top grid, core support structure and control drive thimbles was per-formed. All structures were in excellent condition except for several welds in the grid support structure where small uoro derocrod.
Arackn paralln1 en rhe veld-bend edge o.E. class.
nEracoucleLE copy r leo AT NO. PAGEs Class II APED Library a ov c'^**-
2151 South First Street 35 nnne San Jose, California cQ N CLuslONs The reactor vessel is in excellent condition. The cracks noted in the grid support struct2re do not affect its load carrying ability.
By cutting out this rectangle and folding on the center line, the above informJ;en can Le fitted into a stenderd card file.
For list of contents--drawings, photos etc. and for distribution see next page LFN-610-2).
Atomic Power Equipment De pa rAment _
iNronNATioM pe panso r.
A. f)- I & o S [ 0 $ & O a k ). 0$ & G>
_. Alexander. F.A. Brandt. W. D. Craie. A.11 Hubbard tests unoE sy lt
/ -
M sEcTION Deslen PrMnporing couMTERaleNEo V t W o w'f suitoiNa ANo noou No. Rm 74 Bldg A San Josa California go c,,,,,
atPt-ils Ci s73
11 DISTRIBUTION A. J. Alexander J. L. Murray R. J. Ascherl J. F. O'Mara F. A. Brandt E. W. O'Rorke C. D. Carroll S. Naymark K. P. Cohen E. R. Owen W. D. Craig K. T. Perkins V. A. Elliott F. C. Rally W. H. Ellis R. C. Reid S. C. Furman R. B. Richards F. A. Hollenbach T. F. Robinson D. H. Imhoff G. M. Roy J. Jacobson G. Sege C. R. Jones J. N. Siltanen I. R. Kobsa W. R. Smith A. J. McCrocklin T. Trocki D. McDaniel J. B. Violette C. E. Morris R. M. Curran, GE - Lstg, Schenectady Commonwealth Edison H. K. Hoyt N. A. Kershaw G. L. Redman (2)
C. B. Zitek Nuclear Power Group (15) c/o K. T. Perkins APED Library - 7 VAL Library - 3
111 TABLE OF CONTENTS Page No.
TIS Title Page, Abstract, Conclusions i
Distribution 11 Table of Contents 111 List of Figures iv I
Summary II Introduction III Inspection of Reactor Top Head and Flanges IV Inspection of Nozzles A.
Outlet Nozzles - Internal examination B.
Outlet Nozzles - External examination C.
Inlet Nozzles - External examination V
Inspection of Reactor Internals A.
Inspection Equipment B.
Inspection Procedure C.
Examination of Top Grid D.
Examination of Core Support Plate E.
Examination of Core Support Structure F.
Examination of Control Drive Thimble Penetrations VI Comments VII Appendices 1.
Letter Report - Primary System Weld and Vessel Inspection, Dresden Nuclear Power Station, Dec. 30, 1960.
G. L. Redman, 2.
Letter Report - Reactor Vessel Inspection, Dresden Nuclear Power Station, Jan. 20 and 23, 1961.
G. L. Redman 3.
Letter, R. M. Curran to T. Trocki, April 8, 1961.
4.
Photographic details, core support structure cracks.
5.
Comments on equipment for future inspections.
iv LIST OF FIGURES 1.
Inspection of Vessel Outlet Nozzles 2.
Inpsection of Vessel Outlet Nozzles 3.
Outlet Nozzle Orientation 4.
Interior view of Outlet Nozzle 5.
Inspection of External Welds, Outlet Nozzle
+
5.
Borescope Inspection 7.
Borescope Prism Objective 8.
Lower Grid Plate Support, Crack Location Diagram 9.
Beam Weld Crack 54 05 10.
Beam Weld Crack 54 05 11.
2eam Weld Crack 58 16 12.
Ring Weld rack 67 02 i
I.
SUMMARY
The Dresden reactor yessel a nd internal structure were inspected in December, 1960 and March-April,1961 while the plant was shut down. The examinations were made sy engineers of the General Electric Company and the Commonwealth Edison Company.
Direct visual examination of the reactor top-head closure and mating top flange surfaces showed them to be in excellent c ondi t ion.
Direct visual examination of interior c.nd exterior reld surfaces of the 12 reactor coolant outlet nozzles was pe': formed, including dye penetrant checks of the external surfaces and welds of five nozzles.
These nozzles were in excellent I
condition. The external surfaces of one of the four coolant inlet nozzles were examined and dye checked. The condition of this nozzle was excellent.
A borescope inspection was made of the top grid, core support plate, and core support structure. A total of 48 representative welds in the top grid were examined and found to be in excellent condition. The core support place was examined extensively; 207 of the 483 fuel positions were viewed. No irregular-itie s were found and the condition of the plate was excellent. The core support structure, or lower grid support, was found to contain some cracks along the fusion line adjacent to vertical welds in the beam array. Of a total of 572 selds in the core support structure, 263 were examined. Ho defects were noted in the weld beads; however, hairline cracks we ce found in the heat af fected zone, or fusion line, paralleling 8 of the examined welds.
An inspection of the control drive thimble penetrations in the bottom head of the vessel was performed, also with the use of the borescope. Thirty-four of the 80 thimbles were examined and conditions of the bottom head, thimbles, and seal weids were found to be excellent.
II.
INTRODUCTION The Dresden reactor shutdown and plant outage from December, 1960 to May, 1961.
af ter 1450 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.51725e-4 months <br /> of equivalent full power operation, for rework of control drives end replacement of control blades afforded an opportunity to inspect many parts of the reactor and system. The reactor vessel and internal hardware were inspected with emphasis upon their structural condit ion.
The fuel was removed from the core facilitating this inspection, and made possible direct visual examination of many nozzles and welds in the vessel. Residual radiation from the core structure and the reactor vessel at the core elevation prevented direct personnel i-I l
j access into the vessel below the level of the coolant outlet nozzles and a borescope was used to determine the condition of core structure and reactor l
bottom head.
It was possible by direct visual observation and borescope visual observation to make a comprehensive inspection and evaluation of the reactor vessel and internal hardware.
This in:;pection was initiated in December, 1960, and continued to April 10, 1961. The methods employed and the information obtained in the inspections are given belos.
III.
INSPECTION OF REACTOR TOP HEAD AND FLANGES On January 10, 1951, the following persons visually inspected the Dresden top head and flanges:
E. C. Bailey Commonwealth Edison W. S. Gibbons GE APED G. Redman Commonwealth Edison J. Spero State of Illinois J. W. Sutton Traveler's Insurance Co.
I The visual examination sas made with flashlights and approximately 5X n.agni-fying glasses af ter the head had been scrubbed to remove surface crud and loose contamination. Particular attention was given to the flange bolt circle and the welds at the head penetrations. One questionable area was discovered and appeared to be a possible three-inch long crack on the flange adjacent to one of the stud holes. Since it was impossible to decide, on the basis of visual examination
]
alone, whether this was a crack, a handling scratch, or a mpchine tool mark, a dye I
penetrant examination and light polishing were performed.
No crack was indicated.
IV.
INSPECTION OF N0ZZLES A.
Outlet Nozzles - Internal Examination The radiation level inside the vessel at the coolant outlet nozzle elevation was about 250 mr/hr, S-y with fuel removed from the core and with the vessel water level just below the outlet nozzles This radiation level allowed a j
reasonable stay time for direct visual inspection of the outlet nozzles. The high water level made it necessary for the inspector to be in a prone position to view the nozzles and a portable viewing chamber capable of carry-ing an inspector in this position was constructed from a length of corrugated-culvert pipe. This device is shown on Figures 1 and 2.
The inside of the i
chamber was fitted with a rolling dolly for the inspector to lie on.
Breathing air lines for the inspector's plastic hood and telephone lines for communication
.g.
i
.,_,s..._,.
between the inspector and his crane handlers above were provided. The chamber was maneuvered with the reactor service crane using a long double sling to provide good fore and aft stability. The reactor service platform was used as a staging area. Auxiliary safety lines sere strung from the chamber to the platform to prevent lowering the diamber too far below the nozzle level.
The inspector entered the chamber at the reactor service platform and it was lowered with the crane to the nozzle level. The service table was rotated to index the chamber toward each nozzle in turn.
Retractable struts bearing against the vessel wall were used to steady the chamber. The inspector was able to position his eyes within 6-12 inches of the nozzle to view vessel we ld surface s. The nozzle transition and field welds could be viewed directly only from a distance of 3 feet and a device was made to inspect these welds more closely. This consisted of a light source, mirror aid binoculars held in rigid alignment to se rve as an open periscope which could be inserted through the nozzle up to the inner weld surfaces. Unfortunately, the unit was too large to fit the pipe 1.D. and there was insufficient time to modify it.
The inspection procedure began with the lowering of the vessel water level to an elevation just below the outlet nozzle. The vessel and nozzle surfaces were allowed to drain and become almost dry.
The inspector was lowered in the chamber to the nozzle elevation and was able to view the vessel surface above the nozzles as he descended. The inside surface of the vessel and structure was covered with a thin, porous, uniform film of red crud which acted like dye check developer when wet by water seepage from a crack or open joint in the reactor inner surface. As an example, the non-welded joints between segments in the upper nupport ring for the diffuser-turning vane were cbserved to stand out in bold relief, as maroon lines on the uniform field of orange-red crud. The se remained for more than eight hours af ter the surface of the vessel had dried.
The inspection of the internal surfaces of each of the 12 outlet nczzles and the vessel internal surfaces above these nozzles was performed March 18-20, 1961 by F. A. Brandt. W. D. Craig and A. M. Hubbard.
No cracks were visible ou che vessel interior or nozzle interiors as far back as the transition welds and field weld connections to the piping. These welds are approximately three feet from the point of closest observation. An
~ ~
- - - ~ _ -
i c
i accumulation of " salt" deposits was observed along the upper surface of l
nozzle s 0118, 0119, and 0120 (iigure 3), and on connecting piping to two of
)
i these nozzles, (Figure 4).
The deposits give the appearance of residue le f t i
l by evaporation of water and occured as lenticular areas, 1/2" 1/2" wide a
j and 2" - 12" long. One nozzle was not completely dry at this location, when inspected, and water had collected in the same area in the same pattern. Wire i
brushing and inspection tkoc. a distance of about 18" revealed no evidence I
of cracking. The accumulation of water leads to the conclusion that the i
i deposits are the result of flat-pitched nozzles and surface tension holding a
t he water in these areas. Subsequent drying leaves a residue.
It is also l
possible that this could occur on other nozzles if bubbles were trapped 3
during stagnant or shutdown conditions. No scrapings were obtained.
[
The surface crud was extremely tenacious and difficult to remove with a stainless steel welder's brush, and a power wire brush or grinder would be required to clean any substantial area.
It appeared quite similar to a light coat of spray-painted, red-lead-primer paint, deposited as discrete particles, rather than heavy scale or dried slime. The red color was interspersed with blue, or gun metal, colored patterns of random lines.
It appears that this could be the background color of the stainless steel which has not yet accumulated crud.
4 1
Dye-penetrant testF of th3 nozzle s were not possible because they had not dried sufficiently before inspection time ran out.
We do not believe the dye penetrant test was necessary, because water leakage from cracks is an 4
1 l
equally sensitive test according to our observations at VBWR and from the
[
j indications at the open joints in the Dresden turning vane support ring.
l B.
Outlet Nozzle - External Examination Concrete blocks around a riser pipe in the primary shield were removed to gain access to the outlet nozzles on the outside -of the reactor vessel. The I
aluminum reflective insulation sleeves were removed from five of the twelve nozzles to permit visual inspection. The nezeles inspected were Nos. 0111, I
0112, 0113, 0114, and 0122.
~
It was possible to observe the condition of the external surfaces of the vessel-to-nozzle weld, the nozzle-to-transition-section weld (plain carbon i
steel to austenitic stainless steel, performed at vessel fabricator's shop),
-h-i l
and the transition section-to-riser-pipe weld (austenitic, performed in construction at the site). Superficial rust was general on the carbon steel nozzles; wire brushing disclosed a uniform, smooth, forging finish. Other than this rust, no corrosion was noted on the inspection areas. Dye pene-trant checks were made of the nozzle-transition section and transition section-riser pipe welds.
Figure 5 shows the general external surface condition of the nozzle areas.
This inspection was performed on December 30, 1960 by Commonwealth and on March 21-24, 1961 by W. D. Craig.
C.
Inlet Nozzles - External Examination One (No. 0130) of the four 22-inch diameter coolant inlet nozzles was inspected on December 30, 1960 by Commonssalth and on March 23, 1961 by W. D. Craig. A visual examination and dye penetrant check were made, covering the exterior surfaces of vessel-nozzle weld, the nozzle-transition section weld, and the transition section-inlet pipe weld. No cracks or surface imperfectior.s were noted.
V.
INSPEC1 ION OF REACTOR INTERNALS A.
Inspection Equipment A 1-3/4" diameter Lenas borescope sith modular 10 foot extension sections was used in the examination of reactor internal surfaces below the outlet nozz le s. The borescope was mounted on the reactor service turntable above the vessel flange (Figure 6).
Both direct (vertically down) and right-angle-prism (horizontal) viewing objectives were used en the end of the bore scope.
The reactor service crane was used to raise and lower the borescope into viewing position and to assist in the addition and removal of extension sections. To provide stability and comparative ease of location, a movable guide system for the bores cope was installed on the turntable. This consisted of a steadying fixture mounted on a double set of horizontal ways, to provide X and Y motion. The steadying fixture was provided with a vertical screw jack capable of an 18-inch vertical motion. When in approximate position, the weight of the borescope was shif ted from the crane to the steadying fixture and the screw jack was used to precisely position the borescope. Horizontal translation was accomplished through use of the horizontal ways of the guide system. The reactor service turntable was used for horizontal rotation. The
-5
0 borescope objective could be rotated 360 horizontally to scan all areas at any vertical position. The image formed at the eyepiece of the borescope was approximately IX magnification.
In e f fect, the object viewed appeared to be within 2 to 3 inches of the inspector's eye.
Field of view was approximately 1 inch in diameter. The prism head (Figure 7) was fitted with two 500 watt GE quartzline lamps with cylindrical reflectors on each side of the objective.
These were used alternately to provide two obliquely-lit views of the surface inspected. With this lighting, objects and surface irregularities could be brought out in sharp relief. Additional lighting was provided by a similar lamp on a pole.
It is estimated that with this equipment it was possible to detect and resolve an irregularity as small as 0.002" wide, or less, if the edges of the irregularity were sharp.
B.
Inspection Procedure The reactor vessel was flushed and filled with demineralized water to assure minimum turbidity. The borescope was made up and lowered through the guide system on the reactor turntable. The turntable was rotated to expose the area of the vessel to be examined, and, with the use of the service crane and the movable ways of the guide system, the borescope objective was placed directly above the inspection area.
For examination of horizontal surfaces such as the top of the top grid, core support plate, and the bottom in the thi=ble area, the direct-down objective head was used on the borescope.
Lighting was provided by the pole-mounted lamp, which was shifted to provide many angles of oblique lighting.
For examination of vertical surfaces, such as the fuel holes in the core support plate and the beam welds in the core support structure, the right-angle prism objective head with attached lights was used.
Viewing cond4. ions were excellent, undisturbed by water turbulence since the object, examined were many feet beneath the water surface. The objective head was, in all cases, no more than five inches (in water) away from the e::amined surface. Stability provided by the guide system greatly reduced the fatigue of the inspector.
It was possible to see the " complexion" of the viewed surface, including file marks, weld ripple and individual weld passes. The vertical marks lef t by ae borescope in its vertical passage through the fuel holes in the core support plate were plainly visible.
Two inspectors were assigned to each of the three, eight-hour shif ts during the inspection. The inspection proceeded, around the clock, from April 1, 1961 to April 10, 1961.
C.
Examination of Top Grid The top grid, or upper guide array, was viewed from the top and horizontally from inside the matrices to a scertain the condition of the top and bottom welds at the beam intersections. These welds were seen to be in exactly the same condition as at installation. A thin, uniform film of crud deposited over the surface did not obscure the individual file-tooth marks around the welds. No dif ference in surface texture or coloring was found. As noted during fabrication, undercut was present along many of the weld edges. A total of 48 welds were viewed in a broad sweep fron. the center of the core to the core periphery.
D.
Examination of Core Support Plate The 2-1/2" thick core support plate (Figure 8) was also viewed from the top and horizontally from inside the matrices.
A 360 scan was employed in viewing 207 of the 488 fuel holes.
It was possible to see the difference in surface condition between holes which had been loaded with fuel bundles and those in the outer periphery which had not.
In the latter case, crud build-up was similar to that found on other non-bearing surfaces in the reactor vessel.
In t he fo rme r ca se, crud build-up was much thinner where fuel-bundle nose pieces rested on the bearing surface. No wearing of the plate was seen, aside from the bright new scratch lines le f t by the bore sco pe object-ive in its passage. No dif ference in surface texture or coloring was found.
No peeling or breaking was seen on the fuel bundle bearing surfaces and lead-in edges. The outer periphery of six cor. trol rod guide penetrations through the core support plate were examined. No irregula rities were noted.
E.
Examination of Core Support Structure By lowering it below the co re support plate to the core support structure, the borescope right-angle-prism-objective head was used to view the welds los
-' east-west beams to the north-south beams; and to examine the beam-s,,....ing strength welds near the vessel wall.
See Figure 8, which shows the core support structure, the location of all welds, those examined and the location of defects. The 1/2" fillet welds between the beams were seen to have some undercutting, as was noted during fabrication.
Figures 9 and 10 depict defects adjacent to the fillet weld at fuel position 54-05, under two lighting conditions. A crack starts at the top of the E-W beam and parallels the weld to the center of the 12" depth of the beam where it stops. On the other edge of the weld at t his level on the N-S beam a crack starts and extends to within 1-1/2 inches of the bottom of the beam.
No irregularities are noted in the weld metal between the cracks. The crack runs in the heat af fected zone of the base metal parallel to the edge of the weld.
It does not enter the weld bead. Some low-temperature rust, presumably formed by water oozing from the crack when the vessel was drained, outlines the crack. The cracks a re very tight hairlines and appeac to be less than 0.005 inch in width, and are uniform in width along their length. The contour of the hairline is smooth and undulates along the edge of the iTdividual weld puddles.
No pro-pagation into the surrounding base metal 1, seen.
No cracks of any sort were detected in the other three corners of the i ntersection on opposite sides of the plates.
Examination of 207 of the 488 beam-to-beam welds was made and a total of 7 cracks was found, with lengths ranging from 3 inches to 10 inches. These cracks invariably followed the fusion line along the weld in the beam, and could be described as similar to the crack at position 54-05 except that they were less well defined (See Figure 11). None extended to the edges of the beam.
In each case none of the three other corners of the inter sect ion showed any defects.
In addition, the 56 accessible of 80 beam-to-ring welds were examined.
(See Figure 8). One defect was detected on the beam side of the weld at fuel hole 67-02.
A hairline crack 1/4" long is located 1-1/2" from the top of the beam (Figure 12).
The above examinations were performed by A. J. Alexander, F. A. Brandt and W. D. Craig of APED and D. Bridenbaugh, M. Black and F. DePratt of the Chicago Installation & Service Engineering office of the General Electric Company. Some of the defects ere also viewed by Commonwealth personnel, by W. R. Smith, Sr. (APED) an) R. M. Carran (GE Large Steam Turbine Depart-ment, Schenectady).
F.
Examination of Control Drive Thimble Pene t rat ion s Using the vertical, direct-down objective head of the borescope, 34 of the control drive thimble penetrations of the vessel bottom head were inspected, (Table 1).
It was necessary to observe a thimble through four dif ferent
_ 8.
fuel holes in order to view the entire circumference of the seal weld.
It was also possible to eaamine the condition of the cladding overlay in the bottom head in the area of the thimbles viewed. All seal welds were in per-fe et coadition, as were the adjacent areas of the bottom head. Crud deposi-tion was thicker here than on the higher vessel surfaces previously vieaed.
The crud appeared less tenacious, as capected.
Table I Control Rod Thimble Seal Welds Examined Full
- 3/4*
1/2
- D-9 E-7 C-9 E-5 F-5 D-5 F-5 G-6 D-8 F-7 J-8 D-10 G-3 E-5 G-4 E-9 G-5 F-3 H-8 F-8 J-7 G-2 K-5 G-7 G-8 H-3 H-5 H-7 H-9 J-5 J-5
- Extent of seal weld inspected K-4 K-3 K-7 VI.
INSPECTOR'S COMMENTS BASED ON OBSERVATIONS A.
The reactor top head and flanges, inlet and outlet nozzles, are free of visible defects and are in excellent co.:dition.
B.
The top grid, core support plate, control rod drive thimble penetrations, and lower head are free of visible defects ia the examined areas. They are in excellent condition.
C.
The defects noted in the lower core support structure have the appearance of typical weld toe cracks, and in all probability these cracks follow the weld fusica zone and do not propagate into the adjacent base material. Weld Nos. 2, 3, 4, (see sketch) and beam plates on the reverse side of all observed defective welds are in excellent condition.
If the se de fects were other than toe cracks, they would have propagated to the reverse side of the weld.
_9.
.f 3
4 a
/1 5i
((f l
Tjpical Defect h
SKEI'CH OF TYPICAL LOWER SUPPORT STRUCTURE WELD APPENDIX I PRIMARY SYSTEM WELD AND VESSEL INSPECTION DRESDEN NUCLEAR POWER STATION December 30, 1950 Inspection by: James W. Sutton - Travelers Insurance Company John Spero - State of Illinois E. C. Bailey - Commonwealth Edison Company G. L. Redman - Commonwealth Edison Company Reactor Status This inspection was conducted during reactor shutdown. The reactor had been drained of all water subsequent to fuel removal and control rod removal.
Forty of the eighty control rod drives had been removed at the time of the inspection to permit replacement of certain drive parts wi;ich had been determined as un-satisfactory for their service requirements.
It is expected that fuel will be in the reactor during all inspections following resumption of plant operation in 1961. Total reactor operating (critical) time to date is 2695 hours0.0312 days <br />0.749 hours <br />0.00446 weeks <br />0.00103 months <br />, and equivalent full load (626 MW) operating time to date is 50.5 days.
Radiation Conditions It is felt that lo ier radiation levels existed during this inspection than will exist following future plant operation, particularly within the reactor primary shielding, i.e., at the reactor inlet and outlet nozzles. Contributing to higher radiation levels during future inspections will be the normal loadi ng of fuel in the reactor, increased activation of the reactor vessel materials, and possibly some increased corrosion product activity which is deposited throughott the primary system.
Increase of radioactivity will be particularly noticeable in t he area o f t he reactor outlet nozzles, which is already the area of highest radioactivity.
Future inspection experience will indicate the actual radiation levels to be expected under normal reactor shutdown conditions.
Radiation Exposure Limits The AEC limits are revised in the Federal Register 10CFR20 ef fective January 1, 1961, to reduce annual exposure limits to 5 REM, and 1.25 REM (1,250 MREM) quarterly. Dresden Station Radiation Control Procedures specify in addition to these AEC limits an exposure limit of 300 MREM per week.
Such ihnits are necessary to control exposures of personnel who are required to work periodically in radiation fields.
-11
Preparations for Inspection The quick removal aluminum foil thermal insulation sections covering the welds to be inspected were removed prior to the inspection date. Reactor primary shielding blocks had to be removed at the reactor nozzles to permit access to these areas. Blocks were removed at the N.E. reactor inlet nozzle (e lev. 517 ',
nozzle at 3300 azimuth) and the north side at the No. I north east reactor out-let nozzle (elev. 550", 11-1/2" nozzle at 285 azimuth). The south primary steam drum bolted manhead and welded stainless steel diaphragm were removed. The "D" secondary steam generator (Bechtel equipment piece No. E2-D) 15" shell manhead, east 4" shell handhole and west 4" channel handhole and stainless diaphragm were removed.
Inspections sere conlucted under Special Work Procedure No. 573 requiring pro-tective clothing and respiratory protection where beta contamination was possible.
Paired pocket dosimeters, film badge and self-reading pencils served to record radiation exposuresand timekeeping served to control total radiation exposures.
Inspection Results Visual inspection was made of 10 primary system welds and no defects were noted.
Visual inspection was also made of the primary steam drum internals from the south manhole, and of the "D" secondary steam generator shell internals from the shell manhole and handhole openings.
The steam drum internals appeared to be intact. A thin coating of reddish-brown Jeposit (assumed to be primarily iron oxide) was noted on all surfaces. The secondary steam generator shell appeared fairly clean with a brown costing on the moisture separator vanes and screens. The secondary steam generator tubes were clean sith a dull gray surface coating, and some reddish-brown deposit at the tube rolls. No corrosion was noted, and all internals appeared to be intact.
Locations of inspected welds and radiation eaposures experienced during inspec-tion are detailed in the attached table.
Discussion Notes Mr. Sutton requested notification of design changes of equipment such as con-densate pumps and control rods as well as brief inspection results (he has pic-tures of cr.?.cked drive inJex tubes F-9 and F-5, and 11-9 cracked guide roller coupling).
.le also requested two secondary steam generators have one 4" shell handhole and one 4" channel handhole opened for the next steam generator in-spection.
Mr. Spero is only interested in pressure parts.
It was noted that the drain lines from the emergency condenser return lines were supported as requested at the last inspection. Five of the ten primary drum re-lief valves were replaced with recently tested spares.
Weld reference drawings are included herein as requested.
Mr. Bailey will ar-range a schedule of inspections and discuss with State of Illinois and Travelers Insurance Company.
/s/
G. L. Redman Supervising Engineer Dresden Nuclear Power Station cc:
H. K. Hoyt E. C. Bailey John Spero J. W. Sutt on Tech. File
. 1
~
..~... _ - -....
. _ - ~ _ - ~ ~.
_. ~ ~ -
c TABLE OF INSPECTION IDCATIONS AND RESULTS RADIATION i
LOCATION AND VESSEL PIECE WELD MAX.
AVC.
TIME EXPOSURE ITEM OR PIPING DESIGNATION STAMP NO.
TYPE STAMP NO.
MR/HR MR /ilR MIN.-
MR REMARKS 1
"A" Sec. Stm. Gen.
C98669-2 Iland SS330A3 30 30 6
3 Welds OK Comp't 22" Recire.
Ser. 14 Mach.
Water to Reactor 2
"B" Sec. Stm. Gen.
Iland SS300LA3 18 18 7
2 Welds OK - Brown
, Comp't 22" Recire.
Mach.
Splatter Spots Water to Reactor Noted on E11 and Machine Weld -
j Analyze for E. C.
4 Bailey a$
e 3
"C"(DITTO)
C98669-2 Iland SS291B6 15 15 5
2 Welds OK Ser. 8 Mach.
4 "D"(DITTO) 1-3074 Iland. 36COSS35 25 25 15 6
Welds OK Mach.
f 5
Reactor Pri. Shield-Iland SS39702-A3 100 100 15 25 Welds OK ing Elev. 517; 22" Reactor inlet nozzle (3300 azimuth) 4 6
Reactor Pri. Shield-11and SS280G5 600 350 6
35 Welds OK i
ing Elev. Reactor Outlet nozzle (2850 asimuth) i 1
7 Primary Steam Drum -
300 3
15 Internals intact l
South Manhole Reddish-brown coat-ing on all surfaces (assumed to be iron.
oxide) viewed from-south manhole.
~.
[
TABLE OF INSPECTION LOCATIONS AND RESULTS (cont'd)
RADIATION LOCATION AND VESSEL PIECE WELD MAX.
AVG.
TIME EXPOSURE ITEM OR PIPING DESIGNATION STAMP NO.
TYPE STAMP NO.
MR/HR MR/HR MIN.
MR REMARKS 8
"D" Sec. Stm. Gen.
HSB-5046 10 5
6 1
Shell internals intact - brown (Bechtal Piece No.
E2-D).
coating on all surfaces. Tubes clean. Shell viewei from 16" manhole ar';
4" east handhole.
Channel viewed fron-west 4" handhole.
No corrosion.
l l
i i
APPENDIX II REACTOR VESSEL INSPECTION DRESDEN NUCLEAR POWER STATION January 20 anl 23, 1961 Inspection by: Jame s W. Sutton - Travelers Insurance Company John Spero - State of Illinois E. C. Bailey - Commonwealth Edison Company G. L. Redman - Commonwealth Edison Company W.
Gibbons - General Electric Company Reactor Status This visual inspection was conducted during the shutdoan for control rod drive repair.
Conditions remained the same as during the December 30, 1960 inspec-tion except the reactor vessel was filled with water, thereby permitting access to the canal for insyection of the reactor head and vessel flange.
Preparations for Inspection The inner surface at the head nozzle penetrations were brushed clean and the head flange was wiped clean to facilitate inspection of these areas.
The bore-scope sas checked prior to its use to assure its operability. Radiation surveys were made in both areas of the inspection to establish dose rates for the clean-ing anJ inspections.
RaJiation ConJitions Dosa rates were highest at the center of the head, due to the large area of radioactive corrosion products on the inner surface of the head. Gamma radiation levels sere noted as follows:
20 mr at the center ana 38 mr at the flange of the head.
The gamma radiation level at the reactor flange was 50 mr, resulting pri-marily from gamma leakage up through the vessel-canal expansion seal.
Inspections of the head and vessel internals were conducted under Special Work Permits #509 and 610 respectively.
In addition to the more common protective clothing requirements, face shields were worn for protection from beta radiation during close inspection of the head and vessel flange.
Half masks were worn during the head inspection due to the dust on the inner surface of the head.
Paired pocket dosimeters, film badge and self-reading pencils served to record radiation exposures which were lir.ited to less than 300 mrem.
Following inspection of the vessel flange on 1/23/61, nasal contamination was noted in all of the inspection personnel. This was subsequently removed by several thorough cleanings of the nostrils using swabs moistened with water.
__m.
i The source of this contamination was attributed to the dust on canal walls which was loosened by movement of personnel close to the walls.
4 Inspection Results i
Visual inspection wAs hade of all the head nozzles at the stainless steel weld inlay, of the areas around the head bolt holes and the head flange seating surface.
Appro..imately 20% of the head flange bolt holes could not be inspected due to the obstruction of the head holding fixture. All areas were found to be in good con-
-dition. Two circumferential masks were noted about 1/4 inch from holes #12 and #18.
These were subsequently re-e.camined and dye checked without indication of any cracks.
An attempt was made to use the borescope to inspect reactor vessel nozzle penetra-tion internal surfaces. The borescope was entirely unsatisfactory for this use because of awkwardness in locating the inspection head opposite the area to be viewed, and in attaining a viewing position. The surfaces had a deposit shich could not be satisfactorily removed to pa rmit definition of the surface conditions.
The vessel flange was visually inspection and found free of any defects. The two 0-rings were not removed and, therefore, the recessed sur faces could not be viewed.
Inspection of the flange was conducted from a 4-foot square platform supported from the crane hoist by 4 attached cables. Reactor water level was one foot beloa the flange.
The attached pictures of the inspected areas provide a reference for location and methods used. Pictures of the inspection areas taken subsequent to the December 30, 1960 are included.
/s/
l l
G. L. Redman j
Supervising Engineer Dresden Nuclear Power Station ec:
- 3. K. Hoyt E. C. Bailey John Spero J. W. Sutton Technical File 1
!,~
1 APPENDIX III Dresden Nuclear Power Station R.R. #1 Morris, Illinois April 8, 1961 Mr. T. Trocki-Atomic Power Equipment Department General Electric Company 2151 South First Street I
San Jose, California
Dear Mr. Trocki:
I have examined the construction of the core support grid in the Dresden Reactor vessel and a number of the welds in this assembly using the special borescope arrangement.
The conditions for this examination were excellent.
It was possible to detect minute details such as weld ripples, scratches and machine marks. The weld
+
joints are essentially-free of surface scale. The cracks adjacent to welds are very obvious and I am convinced that there is no significant surface cracking adjacent to any of the welds examined which has not been detected in this examina-tion.
The two cracks which I~ observed closely, follow the fusion line of the weld and in my opinion, they are associated with the stresses set up during fabrication of
-the assembly. Due to the fact that the weld metal has a higher coefficient of expansior than the grid material, the stresses in the cracked location will be I
compressive when the grid is at operating temperature and I consider it unlikely I
that the cracks occurred while the unit was in operation. I see no pattern which t
would indicate continued cracking would be expected and I believe that continued
)
operation of the assembly with a scheduled later examination to be completely satisfactory.
Since ely, R. M. Curran Supervisor - Turbine Materials Engineering Large Steam Turbine Generator Department General Electric Company i
4 '
_.. ~
-_r.
APPENDIX IV Photographic Details. Core Support Structure Cracks The core support structure crack photographs were made with an Exacta 35 mm camera adapted to the Lenoa borescope eyepiece.
It was necessary to remove
{
the borascope outer eyepiece lens and shorten the viewing tube. The camera without lens was held securely at the focus poi-af the image formed by the inner eyepiece lens. The Exacta waist level view-finder was used for focusing since it formed the best image under the dim lighting conditions existing. The prism head and two 500 watt quartzline lamps were used at the lower end of the borescope.
Exposures were made at 3/4 inch intervals to form a composite series of photo-graphs along the length of the weld. The exposure time was determined by trial.
It is difficult to judge a general exposure time since the angle between borescope and image is varied to obtain the clearest picture and the amount of light is highly dependent on this angle. The negatives used to make Figure 10 i
were slightly overeuposed. The negatives used to make Figure 11 were greatly overexposed with the time reduced from 40 to 25 seconds.
Table 2 Figure Film Lights Exposure Seconds 9
Plus X 2
20 10 Plus X 1
40 11 Plus X 1
25 12 Plus X 1
25 i
l l
l t. _ _,
APPENDIX V i
Comments on Equipment for Future Inspection
-1.
Direct visual observation of nozzles from a chamber - It is highly im-probable that this type of inspection will be possible again in the Dresden reactor. At this shutdown the fuel was removed and three months cooling time allowed the most active remaining isotopes to decay. If the radiation field within a reactor vessel will allow a similar type examination, numerous equipment improvements are required. The fully enclosed culvert was intended to give some radiation protection, but it did this at the expense of enclosing the inspector in a potential cof fin.
It is recommended that an open top device be used. This will allow the assistants to observe the inspector at all times and will also allow them to aim the device more easily toward a given nozzle. The observer will be able to move about more freely and escape if he must.
A telephone microphone must be mounted inside the breathing helmet. A fresh 4
4 air helmet is desirable for operator comfort and the airline must be rigged w ith care to prevent fouling. High pressure air should be piped to the viewing platform and expanded to low pressure at a valve controlled by the inspector. In this fashion he will get benefit of the cooling due to the expansion and he can more easily adjust the flow to his own requirement.
2.
Borescope operation - The service crane is an integral part of the borescope operation equipment. An essential requirement is positive dead-slow operation of the hoisting speed. The Dresden reactor service crane is not capable of consistent dead-slow operation and its controls should be modified if additional inspections are to be made. A reliable crane would obviate i
the need for a jack and would speed the inspection operation m7ny fold.
The borescope in this operation sas guided by passing it through an 18" length of pipe attached to the jacking frame. The borescope was supported by a sliding collar which rested on the top of the guide pipe. The collar and top of the pipe were not particularly true and when the borescope was rotated the-error was magnif.ied considerably at the objective end 20 to 30 feet away. A rpherical joint, precisely made, would be preferred to support t he borescope. The guide pipe length should be reduced to about 4 inches and its diameter increased so the borescope hangs freely from the spherical joint.
Guiding should be done from a point near the objective end..
p
?:
D
/
/
f,}," ~'
% S
\\\\
I j
//i- - 'f_.]
4
(
l,jff y
is \\
\\
g!
I
' './
_f f d %%I_
g' e
p
-b-
+
\\'
C',. ;
w NT' N
Y" s
End View of Viewing Chamber l
( ^.3' 7 f y
J I
s.
%==..
T'b
$)
Q 9
t
'T
.l
-]
f
+
i-
'4-l Inspector Enterirg Chamber on Reactor Service Platform Figure 1. INSPECTION OF VESSEL OUTLET NOZZLES VA L 13 9 7
,7;.-.
+
~
. /
f k t. 4 0
\\s
.s
.k i
f.
l s'
/
- li r'
/
1:
iI h
Chamber Clearing Hole Through Reactor Service Platform
'3'}i om W
j 46 w
O i
- :P A
Chamber in Inspection Position Figure 2. INSPECTION OF VESSEL OUTLET NOZZLES VAL 13 90
l Y.
i i
y 0117 i
6 3
7 0111 e.
f*'
.ff r
0118 d~
q q
bhl,
7 Ull2 g
- n;.,.
y[
......,s 0119
' ^
e.. -
3 Q,
N*
k 4
5TE4j'%.-
~s%
_.f
-, A,
%~,p;.,
w A;_ ;;g..
<3 m%v:. :,
4,,
_ p 3.,.
~ "?
0113
~
Y~
W
_- }^{s
. i}
e.
db p-kj$1,<
. % T. Q~ m,. E. _
012 0
-x_.,
.~
l e.
X_
%{ ~b, ~
~
- .hs
. 231..,
i
['
4.[
m.:
- ;g @"
0114 c.
-Y.
,. D ; -:2 jy ij O Q,
~
N,'$
SkYI),v.
Q~%,. ?
0121 F
d
/
NU h'M'1 i
4 H
u.
4 0115
\\
x i!-
Ap 1
0122 is m
0116 i i!
Figure 3.
OUTLET NOZZLE ORIENTATION 44 'I99 6J of Pi l
=
a.
i A
rg:pfrpyM
~
Y
{'.j.p.+ -
3 m
- zy p,
s W:i&.,.
6
..p;;,3 - ap ;,$O
,.1 Sips
.,,. (
.~ a -y, ;. :,:;, 9;g,'_'*;_
)
.'f,jyN O '
1 4
l l
S: _
)
. 'f n
i i
\\m
. :.~
5@
D I M_
q u.
( g-1 3
Q p
gf -
/
b Nozzle 0117. Note Salt Deposits.
Nozzle 0118. Note Salt Deposits.
i Figure 4. INSPECTION OF VESSEL OUTLET NOZZLES 1
l l
vAti4oo l
' //
l
~..
- ?.fC.
2.~. X
'$Y,
. -7,s '
~$s ;
,.,llQ 1
hl,. f. h.h'$ 'll W- {\\fyg.,
A.
t c
-;Y, y '.
..} < 2 4)
'*t g
,-- {-
.g
- y Q.;
i
-1
..1.[' < c, -
fe 1
e,s.
I
.,f s
If :5 I,JlE j' p ',
G j
' ^ );fy., e, c
~!*.
. s?" ;
. s i Ao,mr i
Figure 5. INSPECTION OF EXTERNAL WELDS, OUTLET NOZZLES l
1 I
VAL I4 01
y l
N H
i 2
. ~.
'r
\\
-i
. g l
n
. F
-l$
N-
- $ gnomy j
g;* -
k
]--
4
=
[
.w
.., 2 r Y T
' fj.j i
I i
/
.J J
Figure 6. BORESCOPE INSPECTION d
o I
VAL I4 0 2
m i
a Figure 7. BORESCOPE PRISM OBJECTIVE li I \\l n
VAL 1403
3 E
i e
[ l E
I i
r, 3
a,.
i 5
j n
~.. ',,
j i
i i
4 3
i
=
~
w ::,p s
a :
I
-[
}
[# '
=
=
i e -
j
- rw9yi,
'r
,s..!
g m.
c 1
I s
Ws' 3 s
s t -' =
o 7
s o
s 5
E
- L l
- l z
s o
r -
1 1
4 8
c c
g I
%J<
m 44l. =!
_=N g
/ = i-
- h 2
l. -(
.L h
h
.l'. Y. !.y
/-
, -l
/.
=
=
. i.
[.
A
~ * -
3
.l o
- i..
i
..g
~
8 o
l i
?
?
i i
-.h"
. d ___.
j
- h. 4.L. *.
a
=
~
.r.
I J
8 i
i 1l t
i
- l
?
e
'j 3
.I.*..*..*. _ JM Q
1*
F J
- 2o i
?
i i,
as 8
i-
.M c,
is y w. - - + ~ n .,% 3 d s ~T jj 1 L 1. g II i e l... .d..". 4
- j.,.
2 ;; 2 2 k I/. 'i 2 2 W. 2 2 ...... _.... d %j.j;
- j.,,.
-...I ! s N. . - l. 4 i =
- l...
'~ t .i N; ../:li>'; 1, x. li ji l wl w i t l l 3: I ~ ": *;:
- 1:
El2 tj* 2l2
- l2
- l2
- jf g[g gjg gjg 3
3 a D**D 0 3' K @ oo oJ AkL e 5 r
Top of Crack f on E-W Beam e intersect,on of Core i Support Plate and Core Support %, + '1 Bottom of Crack Top of Crack on E-W Beam on N-S Beam -2 k -l .E -0 ~ Bottom of Crack on N - S Beam i 9 D T .S Bottom of Core g v Support Structure Crack 1. Grid Position 54 05, NE Corner Figure 9. BEAM WELD CR ACK V A L 14'0 5 '
Intersection of Core Support Plate and Core Support l Structure l f Bottom of Crack on E-W Beam Top of Crack on N-S Beam l 2 E 5 -1 f c i - O e i ^ Bottom of Crack on N - S Beam qw m g- . 2. J w ow w f Bottom of Core Support Structure Crack 1. Grid Position 54 05, NE Corner Figure 10. BEAM WELD CR ACK VAL I406
t f Intersection of Core Support Plate and Core Support Structure i 4 i i Top of Crack on E-W Beam -s 2 - o 1-1 5 O - Bottom of Crack on E-W Beam 11 1l 11 QQ' ,Q' J w w fB o Ju 11!nla Crack 6. Grid Position 6816, SE Corner Figure 11. BEAM WELD CR ACK VAL I407
i 1 l l l 6 l 1 -2 Intersection of Core Support Plate and Core Support Structure j I 5 / -0 Crack ( On Beam Side of Beam-Ring Weld ) %" Length Crack 8. Crid Position 67 02, SW Corner Figure 12. RING WELD CRACK VAL I4 0 8 -}}