ML19345C083

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Radiation Level Assessment of Dresden 1 Decontamination Pilot Loop-Final Rept
ML19345C083
Person / Time
Site: Dresden Constellation icon.png
Issue date: 05/31/1977
From: Anstine L, Kenitzer H
GENERAL ELECTRIC CO.
To:
Shared Package
ML19345C060 List:
References
NEDC-12620-1, NUDOCS 8012030854
Download: ML19345C083 (42)
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NEDC-126201 PRIVILEGED INFORMATION M AY 1977 f RADIATION LEVEL ASSESSMENT OF THE DRESDEN 1 DECONTAMtINATION ~ l PILOT LOOP - FINAL REPORT le f t L. D. ANSTINE H. L KENITZER j'. PREPARED FOR THE COMMONWEALTH EDISON COMPANY BY THE GENERAL ELECTRIC COMPANY UNDER P.O. 802551 rI-VALLECITOS NUCLEAR CENTER PLEAS ANTON, CALIFORNIA GENER AL$ ELECTRIC 8012030BW

,i. NEDC-12620-1 Privlieged !nformation May 1977 I' i: t,-

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RADIATION LEVEL ASSESSMENT OF THE DRESDEN 1 DECONTAMINATION PILOT LOOP - FINAL REPORT 9 L. D. Anstine i H. L. Kenitzer f f' 1-I Approved: /d We W. H. Reas, Mdnager Nuclear Processes Development t l Prepared for the Commonwealth Edison Company by the General Electric Company under P.O. 802531 =! I 1 i BOILING WATER REACTOR SYSTEMS DEPARTMENT

  • GENERAL ELT. TRIC COMPANY VALLECITOS NUCLEAR CENTER, PLEASANTON, CALIFORNIA 94566 E5fc"f72n?'

GER ERfiL h ELECTRIC ~ Y

i l = NEDC-12620-1 l .l l DISCLAIMER OF RESPONSIBILITY This' document was prepared by or for the General Electric Company. [ Neither the Gen al Electric Company nor any of the contributors to this document: A. Make any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, sr that the use of any information disclosed in this document may not infringe privately owned rights; or B. Assumes any responsibility for liability or damage of any kind which may result from the use of any information disclosed in h this document. I ,)

..g NEDC-12620-1 .l g t I f CONTENTS Pg. ABSTRACT 1 ' l. INTRODUCTION......................... I 2. SUM M AR Y............ 2 3. CONC LUS!ONS......................... 2 + 4. EXPERIMENTAL TECHNIQUE.................. 4 4.1 Pilot Demonstration Loop 4 1 4.2 General Electric Ge(Li) Pipe Gamma Scanning System....... 4 ' 4.3 Experimental Procedure................... 7 5. RESULTS............................ 12 5.1 Spool Piece 5.2 Pilot Loop 12 20 5.3 Error Analysis....................... 26 ACKNOWLEDC:'ENT...................... 27 REFEAENCES....................... 27 i APPENDICES A. Spool Piece Gamma Spectra - Section A 23 B. Pilot Loop Gamma Spectra - Location 3........... 33 i A t i.f ,...(

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j y .= NEDC-12620-1 RADIATION LEVEL ASSESSMENT OF j TifE DRESDEN I DECONTAMINATION PILOT LOOP - FINAL REPORT i L. D. Anstine H. L. Kenitzer ABSTRACT i tl The radionuclide concentrations of the Dresden-1 decontamination pilot loop were determined by gamma spectroscopy. The General Electric Ge(LI) pipe gamma scanning system was utilized to take measurements at eight locations both before and after the pilot demonstration of decontamination y ~ procers. Dose rate measurements were taken with a Cutie Pie at 30 additional locations. The percentage of Co-G0 removed was calculated and the results were interpreted. i

1. INTRODUCTION As a result of the increasing radiation levels of the primary components of Dresden Unit 1, the Commonwealth Edison Company (CECO) Initiated a program to reduce the occupational radiation exposures by total plant decontamination. The Dow industrial Service Division of Dow Chemical Company was selected as the prime contractor. After evaluating k'nown decontamination technology, Dow recommended a chemical cleaning j

process utilizing Dow Solvent NS-1. An extensive materials testing program has been undertaken with reported favorable results to date.- A pilot demonstration of the solvent and cleaning process using the old General Electric Task K -Corrosion Fatigue Loop at Dresden--I was conducted from June 3 to June 16, 1976. Numerous corrosion specimens were subjected to the solvent and an examination of the loop piping and welds was performed-before and after the test. A special spool piece containing sections of the 4-inch bypass line from Dresder 2 and -3, Monticello, and, Quad Cities 2, and a new 4 i Dresden-1 safe end was fabricaud and insta!!ed in the loop for the pilot demonstration. The Nuclear Energy Systems Division of the General Electric Company is serving as a i consultant to CECO for the decontamination project. One of the GE functions as a consultant was to measure the quantity and distribution of radionuclides present on the piping in 'the pilot loop and the spool piece both before an ' af ter the pilot demonstration. The experimental procedures and results of the baseline and af ter--cleaning mesurements are described. - . V 1 ,w-c,,. y +-w-- w > - m rtr+- T we- -er-e vet e 4v=--'v-- +v w --T

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SUMMARY

Gamma scan measurements were taken at four locations on the pilot loop piping and four ,} locations on the special spool piece both before and after the decontamination of the 'Dresde -! pilot loop. The gamma scan menerements indicated greater than 90% of the .Co-60 r.ctivity had been removed from the loop and spool piece. However, the dose rates determined with a Cutie Pie at 30 additional locations indicated a considerably lower percentage of the activity had been removed. Subsequent measurements with a

directional probe demonstrated this inconsistency was due ta external contamination in the loop area.

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3. CONCLUSIONS d

The major conclusions from the General Electric radiation level reduction measurements - j' made before and after the pilot demonstratio,. of the Dresder>-1 decontamination process are: I 'l 1. The proct.;s appears capab!c of removing greater than 90% of the activity. This f percentage.was determined from the observed reduction ir Co-60, which is the major radionuclide present on the Dresden-1 piping. 2. It appears the' decontamination effectiveness will be partially determined by how well the system is vented and filled with solvent. If large portions of the reactor sy,cem are not fully covered with solvent the decontamination results may be considerably less than the 90+% observed for the pilot demonstration. 3. The residual Co-60 concentrations in the. penetrations on the inside diameter surfaces of the four sections of the spool piece indicate the rinsing procedure may be. inadequate f'r crevices (such as, socket ' welds) which cannot be thoroughly flushed or ' 4.. The relatively uniform, residual Co-60 concentration measured on the loop, on the rop' portion of the spoo! piece, and on the previously uncontaminated Dresden-1 safe-end ~ indicates the decontamination process leaves a reasonably low level of residual activity on straight runs of pipes. 2

NEDC-12620-l ~ i i 5.. ' Based on the post-decontamination' dose rate measurements, it appears external contamination may well be the limiting factor in the reduction of radiation levels at locations where there is excessive external conta, nation. The externa! j contamination varies widely throughout the Dresden-1 reactor, and so will the impact of this contamination en the over-all success of the decontamination program. The pilot loop area probably represented one of the worst cases. Thus, the dose rate reduction (factor of 2 even though greater than 90% of the activity was removed from inside the loop) found in the external measurements of the pilot loop area (after a major portion of the lead had been removed) will most likely be at the lower end of the expected range of dose rate reductions. Cleaning of certain areas exhibiting high levels of external contamination may be required. Il 6. _ Because of the excessive external contamination present in the pilot loop area, extrapolation of the results of the pilot demonstration to the full plant decontamination is extremely difficult, since base line dose rate data for the clean piping without external contamination could not be obtained. While the gamma scan results give an indication of the ' cleaning effectiveness of the process, the true measure of the success of the decontamination program will be L the residual dose rates throughout the Dresden-1 reactor. As discussed above, thesp dose rates will result from a combination of the residual internal activity and the external contamination. Prediction of the relative contributions will depend not only upon the uniformity of the cleaning process, that is, the vent and fill effectiveness, the flushing procedure and the number of crevices (some of which might be inadequately flushed), but also upon the distribution of residual external activity. 7. The quantitative evaluation of the radiation level reductions obtained in a nuclear power plant decontamination is extremely complex and requires a variety of different instruments for measuring radiation levels. Sophisticated directional GE(LI) systems with 2 to 4 inches of shielding are equired to provide accurate t measurements of the activity on the inside surfac.es of the piping and other components. These measurements allow the activity levels both before and after the decontamination to be evaluated 'without interference from -external j contamination. Also, measurement of the ' radioisotopic composition - will frequently provide an insight into the source of the residual activity. Hand-held 3'O direct!.inal probes and Cutie Pies are also required to, locate hot spots (both internal and esternal to the reactor system) and to establish the environrr. ental w dose rates. 3. u-.m

NEDC-12620-1 Measurements taken with these instruments mus: be integrated with the Ge(Li) gamma scan data to provide realistic interpretation of radiation level a reduct!ans obtained in the decontamination process.

4. EXPERIMENTAL TECI'MlQUE 4.1 PILOT DEMONSTRATION LOOP The old General E!cctric Task K - Corrosion Fatigue Loop was constructed on Level 517 in the Dresderr! sphere under an AEC contract. A schematic dia ram of the loop,2 is f

shown in Fi ure 1. During operation, coolant at 500 F and 1000 psig was withdrawn 6 downstream from the outlet of the secondary heat exchanger, circulated through the loop, and then returned upstream of recirculation pump. The loop consists of three test vessels connected primarily with 1-1/2-inch-<!iameter pipe. The dynamic and static vessels were installed in 1969, while the crack propagation vessel was installed in 1974. A summary of the loop operating history!,2 is given in Table 1. Generally, when the loop was not in service it was drained and backfilled with either nitrogen or argon. The loop was decommissioned in June 1975 at which time it was turned over to CECO for the decontamination pilot test. 4.2 GENERAL ELECTRIC Ge(Li) PT '. GAMMA SCANNING SYSTEM The General Electric Ge(Li) pipe gamma scanning system is shown in Figure 2. The specially designed detector and cryostat are shielded by 2.5 inches of Hevimet which provides approximately a 250-fold reduction of the Ce60 photoeak intensity. The detector is collimated by utilizing a series of Hevimet plugs with apertures ranging in diameter from 0.7 to 3.7 cm. The detector output is processed by a multichannel analyzer, and the resultant gamma ray spectrum is analyzed for significant gamma-ray photopeaks by a programmable calculator. The calculator then prints out the gamma photopeak energy, the background counts, the total counts, and the counts per minute. The spectrum is also plotted with an X-Y recorder. The pipe scanning system has been calibrated at General Electric's Vallecitos Nuclear Center for each aperture for a variety of pipe siws and distances.3 These calibration data allow the net counts per minute data to be converted to microcuries per square centimeter for each photopeak identified. The net count rate is determined by subtracting the background enunt rate measured with a solid plug (that is, with the detector totally shielded) from the count rate measured v. th the test aperture. 4

i i NEDC-12620-1 i TABLE I. Dresden Test Loop Operating Summary Loop Cumulative O erating Operating s'eriod Startup Shutdown Time (days) Remarks 9/16/70 10/2/70 17 Loop shutdown in anticipation of scheduled plant maintenance outage. 11 10/72/70 11/5/70 32 Loop shutdown in anticipation of ( schedulec plant maintenance outage. III 11/25/70 12/31/70 69 Loop shutdown to repair leaking fitting. IV 1/7/71 1/14/71 76 Loop secu ed af ter reactor scrc.m. V 1/27/71 2/3/71 84 Loop shutdown for scheduled piant outage. VI 2/17/71 2/IS/71 85 Loop secured to repair steam leaks and install cubicle' air monitor. Vli 6/3/71 6/23/71 105 Loop automatically isolated due to low pressure trip. Vill 6/29/71 8/12/71 150 Loop secured af ter reactor scram. l IX 3/24/72 4/h/72 171 Loop secured af ter reactor scram. X 5/10/72 5/12/72 173 Loop secured for plant shutdown. I XI 3/25/72 5/26//2 173 Loop shutdown to repair leaking fitting. M1 6/16/72 7/10/72 198 Loop shutdown to sealweld leaking fittings. Xill 12/S/72 i XIII 2/4/73 256 Loop secured for plant shutdown. XIV 2/8/73 3/16/73 292 Loop secured for plant shutdown. XV 5/9/73 6/13/73 327 Loop secured for plant shutdown. i XVI S/7/74 8/13/74 333 Dynamic Test Vessel remained ou of service. r f XVII 1/29/75 3/18/75 331 Final Operating Period. y .i LOOP DECOMMISSIONED, JUNE 1975 m 5

'D 6@ $ Pr) rig r 1 NEDC-12620-1 i -@dAi UJ uJ_ t t m ,.1 = -n F _ _ _ _ _ _ _ _ _ _ _ _ - - _ _= _ _ _ _,i l l 1i 1 l l 1 [ DRESDEN1 I ~ PREINRE I CRACK l vggeg t l PROP A G ATim l g VE SS E L g I 1 1 1 (,~,~',~~',' l TEST LOOP CUBICLE l SECONDARY l / SPOOL PIECE l EXCHANGER l eLoor l i [ l I e STATIC l g VESSE L l REclRCULATOR POMP I I' l v l l DYh AMIC l l VE33EL 3 l l l g -- e t____________________; FIGURE 1. GENERAL ELECTRIC TA$K K - CORROSf 0N FATIGUE LOOP TetulDETECTOR G BIAS SUPPLY COLLneA704 [ [ PLUG 200 F OOT Eg'a(.,m, s n y CASLE .n O APERTUR E g HEvrMET SHIELDING ll TABLE l MULTICH A>*ME L ll LlOUID ANALYZER NITHOGEN Il DEWAR !! / lI I LJ I

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CALCULATOR X.Y RECORDER ,1GURE 2. GENE RAL ELECTRIC Ge(U) PIPE GAMMA SCANNING SYSTEM \\_ 6 i-

NEDC-12620-1 4.3 EXPERLMENTAl. PROCEDURE Eight target locations _were gamma scanned both before and af ter the pilot demonstration. The experimentai 'vocedure, including the geometrical configuration of each scan, was identical for the two sets of measurements, with the exception that a 2-inch-thick lead brick was attached to the piping behind the target area for the second set of measurements on the pilot loop piping. All eight scans were taken using aperture No. 2 e (1.23 cr-diameter) and were for 20 minutes. All eigh't background scans taken with the solid pit.J were for 10 minutes. A summary of the experimental conditions is given in Tables 2 and 3 for each of the eight gamma scans. TABLE 2. Experimental Conditions Pilot Loop Gamma-Scans Shielding Height to Pipe above. l Distance" Floor Year Number Location (inches) (inches) Installed i 1-Static Vessel, Outlet 5-1/2 30 1969 2 Dynamic Vessel, Inlet 4-1/2 30 1969 3 Crack Propagation 3-3/4 53 1974 Vessel, Outlet 4' . Crack Propagation 26-1/2 53 1974 Vessel, Inlet ]f TABLE 3. Experimental Conditions Spool Pieces Gamma Scans Shielding Scan Sample-to Pipe Date of Date of l Number - Identification Reactor Distance" Startuo Pipe Removal ~5-D Dresden-3 Contact 7/71 5/75 6 C Dresden Contact 4/70 1/75 7 7 B Quad Cities.2 Contact 3/72 1/75 l 8. A-Monticello-Contact 3/71-1/75 "For_ detector face-to-pipe distance add 2-5/8 inches. L Four of.the gamma scans were of the loop piping, of which two, the outlet line from the h static vessel (Location 1) and the inlet line to the dynamic vessel (Location 2), were of the 3 L_' piping installed in 1969. The other two locations, the outlet-(Location 3) and the ' inlet I (Location 4) lines -f the crack propagation vessel were of piping installed in 1974. ' The f 7

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t ~ g r NEDC-12620-1 exact locations are shown in Figure 3. All four scans were of vertical runs of pipe. The U detector spcrture centerline was 45 and 90 from the inlet wall for Locations I and 2, respectively, and was identical to the centerline of the crack propagation vessel for i Locations 3 and 4. The extremely crowded condition of the loop area prevented the movement of the' assembled detector, therefore, it was necessary to assemble and disassemble the Hevimet snielding at each of the four locations. At Locations 1 and 2, the detector was set on a single layer of lead bricks; while at Locations 3 and 4 the detector was positioned on top of the lead brick shielding that surrounded the crack propagation vessel. The second gamma scan of the 1974 piping (Loc.mion 4) had to be taken at 26-1/2 inches from the shielding to pipe because of the difficulty of positioning the 350-pound shielding and detector. t-The remaining four gamma scans were contact centerline scans of each of the four sections of recirculation bypass piping that made up the special spool piece. The target locations are shown in Figure 4. Before the post-decontamination gamma scanning of the spoo! piece,' Dow personnel expressed concern about the spool piece being cleaned uniformly. They suspected it had not been completely vented and consequently only the bottom portion of the spool piece had bp:n exposed to the solvent. To evaluate this possibility 1-minute scans with aperture No. 2 were taken at the top and bottom of each of the four sppol piece sections. The detector was positioned so that the centerline of the aperture corresponded to a cord 1-1/2 inches from the pipe centerfir.- as shown in Figure 5. The distribution profile was also evaluated for the Monticello and D esden-3 sections by taking 1--minute scans with aperture No. 2 at 1/4-inch intervals from the top position (1-1/2 inches above the centerline) to the bottom position (1-1/2 inches below the centerline). Background scans with the blank plug were not taken at any of these ' locations. The background count rate did not vary significantly (less than 10%) across the profile, and thus, a more accurate value could be obtained by utilizing the center;me background count rate which was taken for a longer time period. Gamma scans with aperture No. 2 and the blank plug were also taken at the top (2-1/2 inches from the centerline) and bottom (2-1/2 inches from the centerline) of the weld in the safe-end at a distance of 2-1/8 inches from the shielding to the safe-end. Radiation level surveys 'were performed using a calibrated Cutie Pie before and af ter the . pilot demonstration to measure the dose rate at 30 additional locations on the loop piping. A!! measurements were at contact with eitner the insulation'or the exposed pipe. No particular attempt was made to locate " hot spots" in the piping or,inthe loop area. Also, - the dose rates from the test vessels were not measured directly, since the test vessels 4. 8 y .g9-m w p-- i 9

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+ 2 NEDC-12620-1 were' fully surrounded by lead at the time of the first survey and only partially shielded - (most 'of. the lead was removed for the pilot demonstration) during the second radiation ~ _ level survey. Two additional dose rate surveys were made.to resolve an anomaly between the gamma scan and dose rate data sets'(see Section 5.2 of this report for a complete discussion). The ,firstisurvey taken in September 1976, after the loop _ area had been. cleaned, consisted - merely of repeating the Cutie-Pie measurements at the thirty target locations. The second survey was_ conducted in December 1976_ with a directional hand-held-probe (Eberline HP220A) modified with additional shielding.- One inch of lead was added to the back of the detector and a 2-inch collimator with a 1-inch-diameter hole was added to the front. A series - of ~ IS measurements of the : loop piping was taken. Three measurements (1 minute each) were made at each location to establish the relative dose rates from the general background, the target pipe, and the environment behind the pipe. The geometric configurations for these three measurements are given in Figure 6. The dose, rates :were quantified by utilizing a scaler with digital readout. Additional background measurements were made at several locations to establish the sources of the residual dose rate.

5. RESULTS 3.1 SPOOL PIECE The experimental result,s of the before and af ter gamma scans for the four sections of the special spool piece are given in Table 4.

Greater than 90% of the Co-60 activity was removed from all four. spool piece sections during the pilot demonstration. Gamma ~ spectra for Section C (Dresden 2) are given in Appendix A. The count rate data (Table 5) for the 1-minute off--centerline scans indicated the residual activity was not uniformly distributed on a!! four sections. In each case, the bottom portion (as the spool piece was positioned in the loop) had a higher Co-60 concentration than the top portion. The_ l-minute gamma scan profiles for the Monticello and Dresden-3 sections (Table 6) confirms this conclusion, since the profiles for the lower portions definitely had higher concentrations of Co-60. 9 12

.~ = 5 _ g NEDC-12620-l' CONFIGURATION 8 CONFIGURATION A ' c HANDLE LEAD SHIELDING TUNGSTEN SHIELOING GM TUBE g .g f LEAD COLLIMATOR l. a 'I -l ~1 INCH DI AMETER HOLE ~ 2 I NO PLUG TARGET PIPE 2 ( LEAD BRICK T ENVIRONMENT CONFIGUR ATION C HANDLE-E \\ J -'I' LEAD SHIELDING TUNGSTEN SHIELDING --- O .GM TUBE LEAD COLLIMATOR r LEAD PLUG TARGET PIPE SCALE LE AD BRICK 1/4 inch = 1 inch ENVIRONMENT FIGURE 6. DRESDEN 1 DECONTAMIN ATION LOOP - RADI ATION LEVEL SURVEY CONFIGURATIONS 13 , -. ~ ,- + eN-www

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1 l NEDC-12620-1 TABLE 4. Spool Piece Gamma' Scan Measurements 2 Radioisotopic Concentration (u Ci/cm )' 1 Before - After Percent Section ' Decontamination Decontamination Removed ? 3.' -Monticello Co-60 2.9 0.0~ 97 (A) 2n-65 2.6- = 0.10, 96 Mn-54 0.06-N.D. t Quad Cities .Co-60 5.7 0.31' 95 (B): Zn-65 0.I 'O.056 44 5 Mn-54 0.15-0.012 92 [ Dresden-2 LCo 3.5 0.33 90 (C) Mn 0.07 N.D. -Dresden-3 'Co-60 8.9 0.83 90 - (D) ' Mn-54 0.24 N.D. 2n-65 N.D. 0.056 2-aNot Detected 1 1 ~ TABLE 5.~ Spool Piece Co-60 Distribution Net Co-60 Count Rate Average Co-60 Corp)entration (cpm) ( u Ci/cm I Section Top Middle Bottom Tea Middle Bottom Monticello. 75 159 415 0.03 .0.09 0.19 ~(A) Quad Cities 365 527 949 0.16 0.31-0.42 (B) Dresden-2 351 367. 636 0.16 0.33 0.28 - (D). Dresden '274 1510 2828-O.12 .0.88 1.22 l (D) Dresden-1 38 126 Safe-End -c j, 3 n in i J9 '*

NEDC-12620-1 TABLE 6. Spool Piece Gamma See Profiles a-s! Monticello Dresden-3 Distance from Co-60 ~ Co-60 a b Centerline. Net Count Rate Zn Net Count Rate Zn-65 (inches) (cpm) Co-60 (cpm) Co-60 +1-1/2 75 274 + 1 - 1/4 89 0.61 294 +1 78 0.27 334 +3/4 80 369 +1/2 93 0.44 474 +1/4 127 0.58 576 0 160 0.78 798 1 -1/4 195 0.87 1107 -1/2 -283 0.74 1527 0.15 -3/4 292 0.83 1995 -1 358 0.78 2528 0.09 10 1/4-438 0.76- '2816-0.08 1 1/2' 415' l.02 2728 0.08 Z" Co = - 0.90 before decontamination. bZn-65 not detected in gamma scan before decontamination. Calibration of the General Electric Cc(Li) pipe gamma scanning system for noncenterline geometries is possible, but has not been performed. However, even if this calibration were performed, quantitative interpretation of the data derived from a nonuniform and unknown distribution is. extremely difficult, if not impossible. The compleyity of this i problem can be appreciated by examining Figure 5 which shows the portions of the pipe. that are scanned in the centerline and 1-1/2 inch off-centerline geometries. From this figure, it can easily be seen that changing the detector location not only changes the i region being scanned, but also the shape of the pipe section being scanned. Average - Co-60 concentration values for these irregular sec*. ions can be estimated by assuming the pipe activity distribution is uniform. However, developing a concentration profile from a. series of these average concentrations is difficult; the areas overlap and are difficult to define mathematically. i Consequently, no attempt.was. made 'to convert the observed ' count rates.-from the Monticello and. Dresden-3 profile scans to microcuries per square centimeter values. However, the count rate profile for a uniformly distributed source was determined by- ~ standard slibration techniques utilizing a., Eu-152 source. The shape of this profile along !,1, with the net count rate profiles (normalized to unity at centerline) observed for the s, .Monticello 'and Dresden-3 sections is given in Figure 7. Calibration data were also ~ "! 3 '

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c NEDC-12620-1 developed at the Vallecitos Nuclear Center for the 1-1/2 inch off-centerline geometric configuration. Values for the Co-60 concentration for the eight 1-minute scans are given in Table 5. However, because of the errors associated with their uncertain geometries, short count times,' and nonca 'orm distributions, these 1-minute top and bottom scans I represent only an approximation, and should not be used to evaluate the decontamination l process quantitatively. The source of the residual activity on the spool piece can be determined by examining the radionuclide composition of the Monticello and Dresden-3 sections. Since the main condensor tubes a Monticello are made of-admiralty brass, there is a significant elemental zinc input into the reactor vessel. The zinc deposits on the fuel and becomes e activated. Zinc-65 is subsequently released to coolant and eventually becomes incorporated in the out-of-core deposits. Typically the Monticello out<f-core deposits contain roughly; an equal amount of Co-60 and Zn--65. Since the other reactors represented in the pilot test do not have Ldmiralty brass condenser tubes, their out-of-core deposit Zn-65-to-Co-60 ratio is considerably lower than that observed at Monticello. Consequently, the Zn-65 can be used as tracer to ascertain where the residual activity on the spool piece originated. Since the Zn-65--to-Co-60 ratios of the Monticello and Dresden-3 sections of the spool piece were essentially unchanged during the s decontamination, the residual activity must have originated from the same. respective sections of pipe. Also since the Zn-65-to-Co-60 ratio of the residual activity on the Dresderr3 piece was I order of magnitude lower than on the Monticello piece, most of the i$ residual activity. could not be from ineffective rinsing or deposition of the '1 circulating solvent. i Metallographic examinat. of the inside diameter surfaces of the four recirculat on loop i ' i t--inch bypass lines t.at comprised the spool piece have been performed at General Electric. All four lines were fabricated from the same heat of material manufactured by the same vendor. Numerous short, transgranular, randomly oriented,10-mil-deep penetrations were present on the inside diameter surfaces of all of this piping. These penetrations were present in the archive piping and were not due to exposure to BWR coolant. Figure 3 shows the penetrations in the inside diameter surface at 10X; Figure 9 a J cross-sectional view of the transgranular penetrations on the inside ' diameter surface at 10 and 500X. The oxide film containing the activated corrosion products' appears a relatively uniform thickness and coats the entire wetted surface includ7g that of the penetrations. Examination of a 'section of the ~ piping from Quad Citi, s I revealed approximately three-quarters of the activity was located in the penetrations. Thus, the four spool piece sections constituted difficult surfaces to decontamincte. 1 17 ) f 1 m. Z Z ~ 1' l. O l . i E T ^

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r NEDC-12620-1 a: ^ Apparently the solvent (NS-1) was able to permeate the penetrations and dissolve the majority.of the oxide film, but was only partially rinsed from the penetrations. The ~ orientation of the irregular geometries should not have affected the dissolution properties ' of the solvent. However, gravity flow from. the ' top penetrations would have aided in rinsing the upper penetrations, but not the lower penetrations.' Thus, an activity gradient would. have been. developed across the pipe with the higher values located on the - bottom surface. 1 The ~ 1-minute gamma scan of the Dresden-1 safe-end, which had not been previomly exposed to contaminated coolant, showed a residual Co-60 concentration approximately equal to that found on the loop' piping and on the top portions of the spool piece sections. Thus, it would appear there was a relatively uniform residual level of activity on the g end.e loor. This activity was most likely due to the rinsing procedure, since the top of the safe-end was a factor of 4 lower than the bottom pcrtion. 5.2 - PILOT LOOP 7 7 The radiation levels in the loop acea af ter the pilot demonstration appeared to be inconsistently high when compared to the quantities of radionuclides removed during the decontamination process. A plausible explanation for the high dose rates was external contamination; the loop area was known to have been grossly contaminated before the pilot demonstration and numerous leaks had occurred during the pilot test. A 2-inch-thick lead brick was attached to the pipe behind each target location to provide background shielding for the second set of scans. This procedural change was instituted to ensure the measurement of the residual activity'(anticipated to be low level) on the piping was not adversely affected by the high background in the loop area. On the initial set of scans,it l was assumed most of the activity in the loop area was contained within the pressure [ boundary of the loop and its location could be readily predicted. The four target locations were carefully selected to ensure there was no major radiation' sources (such as, pipe or vessels) ~1n the detector viewing cone. Normally, the count rate contributio'n from external contamination is minor when compared to the primary count rate due to' the re!stively intense source (the target piping) that is directly in front of the detector aperture. - Also~ the target pipe partially shields the detector further minimizing the background affect. Thus, background shielding was not provided for the first set of scans. 4 g The experimental results of the two sets of gamma scans of the pilot loop piping are given ~ in Table 7 At.three of the four target locations (1, 2, and 3) greater than 95% 'of the l-20 - 4 g m,..-,,--p sr-e t e-J'*

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n 1 NEDC-12620-1 1 Co-60 activity had been removed from the piping; only 32% had been removed from target Location 4. The gamma ray spectrums for target Location 3 are given in Appendix B. - Sections of piping were removed from. target Locations 3 and 4 and surveyed in the laboratory. Their dose rates, 3 and 25 mR/h, respectively, confirmed the gamma scan results wh. :h indicated the residual activity level at Location 4 was approximately one order of magnitude higher than at the other three locations.* The higher level at Location 4 was quite likely due to incomplete venting of the loop, which caused a waterfall ef fect in this vertical run of pipe. .1 1 l TABLE 7. Pilot Loop Isotopic Compositions 2 Radioisotopic Concentration ( 4Ci/cm ) Scan Number Before After Percent (year installed) ' Decontamination Decontamination Removed N. D. *b ' l 1 Co-38 0.11 (1969) Co 3.40 0.024 99 2 Cs-134 0.04 N.D. (1969) Cs(Ba)-137 0.05 0.012 76 ) Co-60' l.45 0.051 96 3 Co-58 0.05 N.D. (1974) Co-60 1.39 0.019 99 4 Co-60 3.26 0.583 82 (1974) aNot Detected b1.173 peak utilized due to calculator error on 1.332 peak. The results of the Cutie Pie radiation level surveys taken before and af ter the pilot loop decontamination are given in Table 8. The location of each measurement point is shown in Figure 10. ' The two sets of data should not be compared directly, since much of the lead - shielding was moved between the two measurement dates, However, even allowing for higher initial dose rates, the observed dose rate reduction was considerably different than the reduction which was anticipated from the results of the gamma scans of the loop - piping..Since the loop was known to be highly contaminated from pasc operational procedures, the high residual dose rate was generally attributed - to external contamination. Consequently, personnel from CECO, DOW, and Genera'. Electric decided. the best approach was to clean the loop area and repeat the radiaCon level survey. The 5 repeat survey (Table 3) taken by General Electric personnel in September 1976 indicated N. l 21

NEDC-12620-1 the area cleaning effort had resulted in essentially no reduction in the average use rate measured.at the 30 additional locations on the loop piping; thus, the anomaly had not been resolved. TABLE 8. Pilot Nop Dose Rate Measurements" Location Before Decontarnination Af ter Decontamination Af ter Cleaning 1974 Pipe. ~ (mR/h) (3/76) (mR/h) (6/76) (mR/h) (9/76) 'l '250 ~65 35 2 180. 65 30 3, 175 50-45 4 200 50 45 5 155 55 60 6-155 110 90 7 270 260 310 8 190 30 110 9 175 45 55 10 .220 45 40 11 125-30 30 12 225 30 35 13 250 75 65 14 150 70 55 t. 15 260 70 60 16-225 80 90 1969 Pipe 17 250 155 175 -18 450' 140 230 19L 500 210 170 20 380 180 140 21 420 120 70 22 450 120 80 23 410 145 125 24 '230-190 200 25 370 250 105 26 -300 100 100 27 200 200 280 28 180 70 70 29 210 80 70 30 160 100 90 "The sets of dose cate data should not be compared directly, since lead shielding wat moved between measurements. , - ~ ~ 22

NEDC-12620-1 D D A \\ O jj = 4 ico a i @ e , e. eie s < /r@ e s s / O ii e s/ e e b ( ) o owo 5 \\ \\ I ' og/ l ge i O O' / <. m ~ 23

NEDC-12620-1 In December 1976, a fourth radiation level survey was performed with the dual purpose cf measuring the relative dose rates from the loop piping at a larges nuinber (18) of locations and locating the sources of the high residual activ.ty. The results (in counts per minute) of the 18 measurements are given in Table 9 along with their correspond:ng locations. These data show there was a low level of fairly uniform residual activity around the loop and support the previous gamma scan results by providing measurements of the piping at 14 more locations. No hot spots were observed on the loop piping. TABLE 9. Pilot Loop Radiation Level Survey - Directional Probe Dose Rate (cpm) Location

Background

Pipe 6 6 57 7 300 78 8 76 38 9 62 4S 12 64 16 13 111 64 16 140 -1 17 294 58 18" 479 13 19 409 32 b 19A 1035 -12S 20 180 -9 21 133 30 22 90 6 23 286 -3 24 533 -24 26 117 53 27 263 10 29 127 47 "New Pipe - Not Contaminated b19A - Between location 19 and inlet to dynamic vessel. The background readings (configuration C of Figure 6) with the plug in ph. . vc..d a distinct pattern of activity (Figure 11) with major sources located around trw dynamit.; vessel and under the flanged end of the static vessel. Subsequent Cutie Pie measurements indicated extremely high dose rates (greater than 1 R/h) around the entire dynamic vessel. The dynamic vessel insulation, which was thoroughly soaked, was then removed from the vessel and placed in a plastic bag. The dose rate from the bag was 1.3 R/h. The outside of the vessel was wet and extremely dirty, with indications that spills had occurred over 24

N 2. m. DOSE nATE CONTOUR LINES .: x a:.. x :.. STATIC VESSEL \\ )- l '~ b ,O K l f170) g 7 f, y l ~ 1 L-----______________.___l s y;L (~ A ^ / ~ l~~~ N 1 ,f 't n 1 f f a / \\ \\ LE AD SHIELDING '/ o \\ N / r l DvnAu C ( 710 VESSEL' ~ N I I[~ N 17 l 3 n 3, , ' ~, Th x \\ li72 N y l Ul J. h ? l CRACK PitOPAGATION N \\'~ ~ / O \\ ,7 / VESSEL 1 i y \\ k

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/ 1'1 0 y i N- / ? s f XITV /CSN, ! 1 i i i i1 Oc-h l~~=c55 oo. ?.:~ - Y # i c I m Q w wn= wn SCA?.E ). DETECTOn LOCATION AND DtPECTION 1 inch = 26 inches BACKGROUND READINGS 800485 TOTALLY SHIELDED j GM TUBE FIGUnE 11. DRESDEN-1 DECONTAMINATION PILOT TEST COOP RADI ATION LEVEL SUnVEY - BACKGROUND DEADINGS 1 1 6

- NED C-12620-1 i the years. ' Wiping of the vessel walls reduced the dose rate by approximately 407 (from -'approximately 600 to 350 mR/h). The CECO personnel had previously measured the dose rate inside the vessel to be 230 mR/h and reported the inside vessel walls were clean.' ' A second hot spot (greater than 5 R/h) was located on the floor under the flange of the static vessel. - The dose rates from the vessel itself (determined with the shielded probe) were orders of magnitude lower. These data strongly support the hypothesis that the majority of the residual activity was external to the loop. However, absolute proof that there was no crud traps or hot spots remaining in the loop was not obtained. Absolute proof would have required cleaning the entire loop area to tha extent that the external contamination was below the activity level defined as a hot spot. This task would have required extensive and costly additional efforts and was not dyer ed necessary because of the preponderance of evidence supporting the external contamination hypothesis. In conclusion, the anomaly that existed between the gamma scan data and the dose rate measurement was due to contamination external (primarily on the insulation and floor) to ^ the loop. Thus, while the residual dose rate in a decontaminated area should be a realistic measure of the effectiveness of the decontamination, this normal yardstick is not a fair measurement of the pilot decontamination. The excessive external contamination present in the pilot loop area may or may not be typical of the rest of the Dresden-1 plant,- but its effect on 'he residual radiation levels will most certainly vary greatly throughout the t plant. Thus, this contribution, which cannot be affected by the decontamination, should not be used to evaluate the decontamination pilot test. External contamination should be I evaluated independently on a localized basis and appropriate cleaning efforts made so that ] ' the desired residual dose rates are obtained, t The Ge(LI) gamma scan results,; supported by the shielded dose rate measurements at ~ 18 locations, give a consistent. analysis that indicates that greater than 90% of the activity was removed from the loop and that there was a relatively uniform, but low, level of Co-60 distributed throughout the loop. 5.3 ERROR ANALYSIS A rigorous statistical analysis of the gamma scan data was not performed. However, the major errors associated with the Ge(Li) gamma scan measurement technique are counting b statistics (approximately 3% relative for the initial set of scans and 7 to 20% relative for 2d, y e-- 4 9 P

NEDC-12620-1 the second set of scans), the detector calibration (less than 10% relative), the detector positioning (approximately 10% relative), and the distribution of radionuclides on the piping (approximately IC% relative). Combining these estimated errors gives a - measurement of uncertaidy of approximately 15% relative for the initial set of scans and. 25% relative for the second set of scans. ACKNOWLEDGMENT The authors express their appreciation to D. Adam, D. O'Keefe, V. Chaney . and other members of the Radiation Protection unit of the Dresden Nuclear I Power Station for their cooperation and assistance in completing the loop survey in a timely manner; to G. Harrison of Dresden-1 technical staff for his assistance in scheduling, setting up, and performing the measurements. The authors thank E. L. Burley and J. Blok of General Electric for conducting the third _ and fourth radiation level surveys, respectively; G. F. Palino for developing the calibration data, and D. A. Hale for providing information on the loop and its operating history. The authors also wish to express their appreciation to J. H. Holloway and C. P. Ruiz for their assistance in establishing the experimental goals and in reviewing this document. i b. REFERENCES 1. Vandenberg, S. R., " Reactor Primary Coolant System Rupture Study-Quarterly heport No. 25," GEAP-10207-25, July-September 1971. 2. Hale, D. A., " Reactor Primary Coolant System Rupture Study-Quarterly Report No. 33," GEAP-10207-33, January-June 1975. 3. Palino, G. F.," Radioisotope Activities'on BWR Primary System Piping 1. Calibration of the General Electric Ge(LI) Pipe Gamma Scanning System," NEDC-12646-1, October 1976. ? 4. Anders, O. U., Private Communication, December 1976. 5. Burley, E. L., " Trip Report - Dresden,' September 8-11, 1976," December 1976. 6. Harrison, G. J., Private Communication, December 1976. A 27 I. :

NEDC-12620-1 APPENDIX A SPOOL PIECE GAMMA SPECTRA -SECTION A i 1 w A** 28

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l NEDC-12620-1 l I EXTERNAL DISTRIBUTION a No. Copies 1 T. D. Boyce (DIS) I l' i C. F. Chang (Argonne National Laboratory) 1 ' i! D. E. Harmer (DIS)* 2 i G. J. Harrison (CECO) Dresden Station I W.1. Kiedaisch (CECO) 2 g L-T. C. Quaka (CECO) 1 R. Staehle (O.S.U.) 1 W. P. Worden (CECO) 2 ij G. P. Wagner (CECO) 2 J. S. Graves (CECO) 5 kI (. t; t I L .J n .b '}}

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