ML20151G260
| ML20151G260 | |
| Person / Time | |
|---|---|
| Site: | 05000142 |
| Issue date: | 04/12/1988 |
| From: | Mclaughlin J CALIFORNIA, UNIV. OF, LOS ANGELES, CA |
| To: | Alexander Adams NRC, NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8804190302 | |
| Download: ML20151G260 (28) | |
Text
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- 1.os AscEt.Es . Rin E Rsint . uN DH G0 * # 4N FR 4NCisOO S ANT 4 84RB4R4 . 5 4%14 0RLZ DEPARTMENT OF COMMl'NI1Y S 4FETY RESEARCll & (XX't'PATIONAL SA6 TIT DIVISION LOS ANGELES, CALIFORNIA 9024 April 12, 1988 Mr. Alexander Adams U.S. Nuclear Regulatory Commission Washington, DC 20555 Attn Document Control Desk Docket Number 50-142
Dear Mr. Adams:
Enclosed are our responses to questions 9 through 15 (from the March 12, 1986 NRC letter to W. Wegst from H.N. Berkow) and the Phase I summary report on the dismantlement of the UCLA Argonaut research reactor. This facility is located in Boelter Hall of the School of Engineering and Applied Science (SEAS). You were correct during recent telephone conversations with Jeff Kleck and me that our January 12, 1988 submission was inappropriate as a Phase I plan. Responses to the remaining questions (cited above) and a request for quotation intended for prospective contractors for dismantlement and decommissioning comprised this submission. We are now developing our Phase II plan and should send it along in a few weeks. The "RFQ" has been formally interrupted in the meantime. If you have questions or require more information, please advise me here at (213) 206-6413 or Jeff Kleck in the Radiation Safety Off'ce at (213) 825-6995 or 52040. Sincerely Yours, s dht& }lt /& James E. McLaughlin Acting Director JEM:si i Enclosure
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- As indicated on next page STATE OF CALIFORNIA ) 8804190302 880412 COUNTY OF LOS ANGELES )
PDR ADOCK 05000142
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University of California Docket No. 50-142 at Los Angeles cc: Mr. Neil C. Ostrander, Manager Committee To Bridge The Gap Nuclear Engineering Laboratory 1637 Butler Avenue #203 School of Engineering and Los Angeles, California 90024 Applied Science University of California Mr. John Bay at Los Anneles 1022 Peralta Street Los Angeles, California 90024 Albany, California 94706 Attorney General Mr. James R. Heelan 555 Capitol Mall Director, Society Services Sacramento, California 95814 American Nuclear Society 555 N. Kensington Avenue California Department of Health La Grange Park, Illinois 00525 ATTN: Chief, Environmental Radiation Control Unit Roger Kohn, Esq. Radiological Health Section 524 lith Street 714 P Street, Room 498 Manhattan Beachs California 60266 Sacramento, California 95814 Mr. Daniel Hirsch Robert M. Meyers 3489 Branciforte Drive City Attorney Santa Cruz, California 95065 Lynn Naliboff Deputy City Attorney William H. Cormier, Esq. 1685 Main Street, Room 310 Office of Administrative Vice Santa Monica, California 90401 Chancellor University of California Roger Holt, Esq. 405 Hilgard Avenue Office of City Attorney Los Angeles, California 90024 200 North Main Street City Hall East, Room 1700 Christine Helwick, Esq. Los Angeles, CA 90012 ' Glen R. Woods, Esq. Office of General Counsel U.S. Nuclear Regulatory Comission 590 University Hall Region y ee on 94720 5 ba a e u te 1 Dean Hansell 302 South Mansfield Avenue Los Angeles, California 90036 l l 1
ei i . Attachment 1 Responses to Seven Questions for NRC on Phase I Decommissioning Plan (Ref 3/12/86 letter to W. Wegst (UCLA) from H. Berkov (NRC)) Question 9: For each task in Table p-5, "Task Identification", provide a description of the procedures for accomplishing the major activities. Any special health and safety considerations should be addressed for each task as appropriate. Include a table describing each task and associated exposure in person-rem and the total camulative exposure for the entire deco =missioning effort. Response 9: Procedures: Reactor core components were disassembled piece by piece using the NEL staff augmented by the health physicist assigned by the Radiation Safety Office. The work was controlled by that Office with a series of Radiation Work Permits that described the current radiation environment and the tasks permitted, prior to the next assessment of the radiation 1cvel. Personnel exposures were tracked on a daily (sometimes hourly) basis using pocket dosimeters. The table below reproduces the personnel exposures for the tasks listed in Table p-5 of the Decommissioning Plan: Phase I. Personnel Exposures ioerson-rem) By Task Cumulative
- 1. Planning and Review 0 0
- 2. Mobilization 0 0
- 3. Removal of External Equip 0.3 0.3
- 4. Disassemble Reactor Core 1.9 2.2
- 5. Package Materials for Transport 0.4 2.6 1
- 6. Transport 0.1 2.7 l
- 7. Assess Radiol Status 0 2.7 )
Data are from assigned personnel film badges. Tasks 4 through 6 were performed concurrently and the breakdown given is an estimate for the individual tasks. The highest individual dose during Phase I was 1.24 rem. A total of five persons received some radiation dose during the Phase I work. j l There were few special health and safety considerations, and few unique or special procedures were required. Protective clothing, dust masks, respirators, and high volume air sampling were employed as required. Bioassays (whole body counting and urinalysis) showed no measurable internal intake of radioactive material. 1 1 l
ss e . Question 10: Describe how pipes, drainlines and ductwork will be surveyed for contamination on the interior to assure conformance with Regulatory Guide 1.86 dose limits. If no appropriate access point for these surveys is available, indicate what survey plan would be used.to make the measurements necessary to identify potential contamination of the interior of pipes, drainlines and ductwork. Response 10: Inlets and outlets of drainpipes were wipe tested. .The exhaust system ductwork is accessible at the reactor room floor level (first floor), at the third floor, and at the eighth floor (roof). Wipe tests were made at those points to determine the need for decontamination. All surface contamination levels were found to be below the limits prescribed in Table 1 of U.S. N.R.C. Regulatory Guide 1.86. Fur:her surveys will be performed during , Phase II as increased access is gained to assess potential activation from neutron streaming in pipes and drainlines. Question 11: Provide a discussion of the final decommissioning alternatives, DECON and SAFSTOR. Include:
- 1. Cost estimates for each alternative.
- 2. Availability of funds.
- 3. Mejor tasks and schedules (in particular, the estimated data for completion of decommissioning or major interruptions in the plan.
- 4. Items subject to quality assurance.
- 5. Tasks that may be performed by a contractor.
- 6. The final radiation survey plan.
, 7. The collective dose equivalent (person-rem) for performing the selected decommissioning plan as compared to each alternative plan considered. Responso 11:
- 1. SAFSTOR is no longer under consideration. Based upon discussions with potential contractors, the cost of completing decommissioning is estimated to be in the vicinity of $200,000.
- 2. The School of Engineering and Applied Science will supply the funds; the Radiation Safety Office will provide appropriate health physics support.
- 3. Major tasks and schedules:
- a. Concrete demolition and removal 60 days
- b. Clean-up and survey: 30 days
- c. Estimated completion date: 90 days after
] NRC plan approval. 2-
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- 4. Items subject to Quality Assurance: The contractor will be responsible for the quality assurance of his equipment, and the UCLA Radiation Safety Office will be responsible for the QA of radiation instruments used to verify contractor measurements. UCLA understands that the final inspection will be done by a third party under NRC contract and assumes that the third party will be responsible for the QA of that inspection.
- 5. Contractor tasks: Concrete dismantlement,' packaging for shipment, transport of RAM and non-RAM off site, facility decontamination as required and surveys to assure complete '
decontamination.
- 6. Final radiation survey plan. The final radiation survey plan would be a part of the contractor's task. However, UCLA intends to perform independent measurements to assure conformance with appropriate federal and state requirementa and regulations.
- 7. Collective dose equivalent for DECON: This is difficult to estimate in the absence of specific details of contractor's demolition procedures. However, the cumulative dose equivalent should be below the 2.7 person-rem shown in response 9. This will be discussed with the contractor as necessary. ,
Ques 'on 12: Section 1.0 Background - Surveys in the areas indicated could include activity from fission products as well as activation-products. Please describe any fission products that would be a probable contaainant in the reactor area. If none is expected, please so state and justify your position. Response 12: The only "fission product" ever detected was Cesium-137. Its' presence is attributed to a liquid source once used to calibrate the secondary water effluent monitor in the early 1960's. Traces of this Cs-137 are still found in the sump. A great variety of > samples have been subjected to Ge(L1) analysis to identify the species present. No other fission products have been found. Question 13: (referring to Phase I plan) Page 2 - Sixth paragraph . In specifying items subject to QA, no mention is made of QA being applicable to radiation survey and sample analysis instrumentation. The licensee must assure the accuracy of all measurements as part of his final report to the NRC. Consequently, instruments used to measure levels of radiation for release of materials, equipment, structures, etc. is a prime item subject to QA. This must be addressed in the final report. ; l l l l
- i. i, Response 13:
Quality assurance of radiological instrumentation is the responsibility of UCLA's Radiation Safety Office. Routine calibrations of survey instruments are performed and documented. Periodic source checks of instruments used in decommissioning were performed. Liquid scintillation, whole body, and Ge(L1) counters are checked prior to each cr eration to assure that they are operating within past histo-ical parameters. Only NBS traceable sources are used in these QA functions and calibrations ara performed by qualified staff members of the Radiation Safety Of* tee. Question 14: Page 4. Section 1.4 This section requires further clarification. The staff is assuming that the DiCON alternative has been selected for the UCLA reactor decommissioning. If the SAFSTOR option is selected because the DECON "decommissioning activities may have implications that go beyond the reactor facility." then the licensee will be required to submit his plan for decommission under the SAFSTOR alternative. In any event, Section 1.4 needs further explanation in the Phase I Final Report. Response 14: SAFSTOR is no longer under consideration as a decommissioning alternative. Question 15: If tritium is identified, provide the origin and degree of tritium contamination within the reactor facility. Indicate how tritium decontamination will be performed. Response 15: Tritium was identified in the primary water in minute concentration of about 3.25 nC1/ml. This activity concentration was well below the applicable maximum permissible concentration (0.1 C1/ml) and was thus disposed to the sewer. The origin of tritium was traced to neutron activation. As outlined in the Phase I Report, the tritium activity was the largest contributor to the total activity in portions of the concrete shield nearest to the cove. Swipes taken for routine area surveys were randomly checked on the Liquid Scintillation Counter for any residual tritium contamination. No tritium contamination was detected in any routine survey sample. I
A 5 - O Attachment 2 l l REPORT OF UCLA REACTOR DECOMMISSIONING PHASE I s t f l l I i 1
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1.0 INTRODUCTION
The dismantlement of the 100 kW Argonaut reactor facility at the University of California, Los Angeles (UCLA) was planned as a two phase program due to large uncertainties in defining the radiological status of the facility and the estimated cost of decommissioning. The range of the cost estimates for the entire job, provided informally by contractors, was $350,000 to about
$1,000,000.
A Decommissioning Plan for Phase I was submitted to the Nuclear Regulatory Commission (NRC) on October 29, 1985. Phase I was ordered by the NRC on July 14, 1986. This report on the completion of Phase I is submitted in accordance with paragraph 3 3.8 of the approved plan. The principal objective of Phase I was to remove the reactor core structure and, hence, gain access to the interior of the biological shield and determine the distribution of neutron induced activity in the surrounding concrete. This assessment, as detailed in section 3 here, was carried out by spectral analyses of concrete and steel samples collected from several beam port plugs and a spherical shell computational model. Phase I dismantlement specifically included the removal of the following: reactor core-moderator, graphite thermal col umn , shield tank, peripheral equipment, fuel boxes, control blado system components, protruding parts of pipe and structures, and most of the concrete shield blocks. All these materials were either transported in accordance with the relevant parts of the Title 49 of Code of Federal Regulations (CFR) or released for unrestricted use in compliance with the criteria of U.S. Atomic Energy Commission Regulatory Guide 1.86. To complete the dismantling and decommissioning of the reactor f acility, UCLA intends to submit a Phase Il plan to the NRC and distribute a Request for Proposal to potential commercial bidders for performing the work in the ; near future. l The present report is divided into five sections that summarize Phase 1 and i detail the various dismantlement steps. Section 2 describes the actual j operations performed while Section 3 explicates the model calculations used I in assessing the distribution of radioactive material in the biological j shield. Section 4 provides na account of the radiation protection program. 1 The concluding swction discusses financial costs and future plans of UCLA in the final Phase II decommissioning of the reactor facility. i 1 1
2.0 DESCRIPTION
OF DECOMMISSIONING OPERATIONS Af ter receipt of the order by the NRC dated July 14, 1986 authorizing the. , dismantling and disposition of various component parts', Phase I operations ! commenced in mid July 1986. The reactor internal structure was removed by four UCLA staff employees with occasional help from two other staff members. The names and technical qualifications of these personnel were described in our responses to NRC questions 1 and 6 on March 21, 1986. The tasks were controlled by the ; issuance of work permits. ' 2.1 Removal and Transfer of Graphite and Lead from the Reactor Cores l Based on the preliminary radiation survey of samples of graphite and lead from the reactor core (see Appendix B of the Phase I Decommissioning Plan), it was anticipated that unstacking and packaging the core material would require four to six weeks. The work took a longer period of time because ; extensive sampling of the graphite was needed to account for variations in both total activity (Co-60, Eu-152, Eu-154) and in the isotope composition I (ratio of Co-60 to total Europium). On September 16, 1986, except for a few small pieces (odd shapes) of graphite, all the graphite and about 80 % of the lead was transferred to ; Texas A & M University as "Low Specific Activity Material (LSA)" in conformance with the relevant parts of Title 49 of the Code of Federal Regulations (CFR), title 10 CFR 71, and NRC regulations. '. A majority of the remaining 20% of lead was surveyed and released to other UCLA organizations (9.g., Neutron Therapy Cyclotron, etc.) for use as shielding. At present, about 5000 lbs of radioactive lead (lead shots, odd pieces, etc.) is being stored in the reactor facility for eventual disposition. Land burial is not possible at this time. 1 2.2 Peripheral Support Structures: During the latter part of 1986, the fuel boxes, reactor control blades, control blade shrouds and the protruding portions of embedments were removed. The shield water tank was removed and cut into suitable pieces for segregation of tne lower tank radioactive parts from the upper non-radioactive parts. The radioactive metallic parts and other radioactive pieces of pipes and structures were packaged and shipped to our land burial site operated by U.S. Ecology Company at Hanford, Washington on February 10, 1988. The metallic parts consisted of aluminum, magnesium, and steel (structural and stainless). 2 3 Removable Concrete Snield Blocks: There were a total of twenty nine removable shield blocks (see Fig. 2.1). Of these, twenty two concrete shield blocks (46 tons) were identified'as non-radioactive and were celivered to the Wildlife Waystation in San Fernando, California, during September, 1987 for use in soil erosion j t
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prevention. Of the remaining seven, four have not been removed and still reside in the reactor f acility. These are identified as C-8, C-9, A+B, and C+D in Fig. 2.2. The other three blocks identified as C-6, C-7 and C-10 were removed as L.SA material. A variance was sought from the State of
'n'ashington to permit burial of these three blocks without additional packaging. The variance was granted on December 8,1987, and these blocks were shipped for burial to the U.S. Edology site at Hanford on February 10, 1988.
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3 0 ASSESSMENT OF RAD 10hCTIVE MATERIAL IN THE BIOLOGICAL SHIELD A diamond saw was used to section concrete neutron beam port plugs. These sections were collected in 1986 and early 1987 to determine the distribution of radioisotopes ih the biological shield. Special sampling and counting continued throughout the summer to refine estimates of the distribution in regions not well represented by the beam port plug data. 3 1 Assumption: It kas assumed that a spherical shell model would adequately describe the majority of distribution of radioactive material in the biological shield. 3.2 Tne Spherical Shell Model: The isotope distribution model was basea upon twelve samples of concrete taken from three beam port plugs containing heavy concrete like that of the biological shield. One inch disks were sectioned from the plugs at six inch intervals. The aluminum casing was stripped from the perimeter of the sections; each section was then weighed and counted on the Ge(L1) spectrometry system. Tne beam ports were in the same horizontal plane, near the reactor core mid-plane. The samples were identified by their beam port location (N,N'n', SE) and by the distance (along the beam port axis) of the sample from the interior facts of the biological shield. The "N" port is perpendicular to the north f ace of the core, the other two enter the core diagonally as indicated by their compass directions.
, The Ge(L1) counting system was calibrated for two sample geometries. The standard source was placed at the same distance from the detector as the bottom surf ace of the sectioned disks and secondly, the source was placed on the top surface of a one inch thick sect. ion of similar, non-radioactive concrete. An empirical self-shielding f actor was determined from the dual calibration data. The counting standards: Na-22, Mn-54, and Co-60 were "NBS traceable" standards.
3.2.1 Correlation of Data: The beam port plug data was fitted to the following equation: BCo Ro e(p(Ro - R)) C(x,y,:) = (I) where: C(x,y,r) = the total radioisotope concentration at point x.y,z, C o = 0.351 wCi/g, R o = 38 inches, B = 4.13, u = 0. R =(x}70pgrinch,gd +y +zj d
The numerical value of oR , while arbitrary, happens to coincide with the perpendicular distance from the core ceriter to the north (and south) faces of the monolithic concrete. Any positive value (including zero) could be chosen by adjusting C o and R o so that C R exp(p Ro) has a constant oo numerical value (about 17 Cl-inches /g) . The radius R was measured from the center of the reactor core cavity which once contained the core and the graphite moderator. The dimensions of the rectangular void were 76 inches (N-S), 60 inches (E-W), and 56 inches high. R is the distance from the center or origin to any point x.y,z. The correlation function chosen is that of a solution to the point source neutron diffusion equation. The observed value of u is 1/L where L (2.70 inches) is the diffusion length calculated for thermal neutrons in the concrete. The values of p ando C were initially chosen by a least squares fit of all of the distribution data. There were, however, four extreme points; two were underestimated and two were overestimated by the model. The value of o was adjusted so that no point would be underestimated by more than 20 C percent. 3.2.2 Pure Beta Emitters: To determine the activity concentration of pure beta emitters such as Tritium and Carbon-14, five samples of concrete from the south beam port (1.3 inches from the face of the shield) were analyzed on a Liquid Scintillation Beta Spectrometer. Each of the five samples were cut and grinded into fine powder and then analyzed with liquid scintillation gel solvent. The Ge(Li) spectrometry data from other samples and the correlation in equation 1 was then multiplied by the constant B to account for the contribution of tritium and carbon-14 Thus, at a point x y,z from the center of the core, the correlation gives the total radioisotope concentration in the concrete. Fig. 31 illustrates the correlation of the horizontal beam port data. The radioisotope concentration shown on the vertical axis is the total concentration of the mixed activation products. The radioactivity composition of the mixed nuclides in the fixed concrete is approximately as shown in Table 3.1. The maximum Co/Eu occurred in sample Nei-18, the minimum in sample NW-24. These two are unusual; all other samples had a cobalt-60 composition bounded by the value 9.32 0.7% in February, 1967. There is no apparent relation between variations in relative composition and differences between the measured and calculated activities of the samples.
Table 31. Relative isotopic ccmpositions within the concrete. Nuclide Half-life Relative Composition (years) (percent) Cobalt-60 5.27 9.0 , Europium-152 13.6 13.6 Europium-154 8.6 '1.0 Manganese-54 0.856 0.3 Cesium-134 2.06 0.2 Tritium 12 35 61.8 Carbon-14 5730.0 14.0 3.3 Limitations of the Model: While the spherical shell model closely predicts the activity concentra-tions in different regions, it has limitations. For example, in the region east of the core which contained the graphite thermal column 60 inches wide by 56 inches high by 44 inches (east-west), and the region directly above the core which contained graphite over a square area 36 x 36 x to inches thick, the model underestimates induced activity. Also, the model over predicts the radioactivity in the concrete that is located adjacent to the structural steel or surrounded by steel.
. 3 3 1 The Modifying Factor: All of the removable concrete blocks had angle iron edges of dimensions, 2 x 2 x 0.25 inches. Samples of a.igle iron edge strips and adjacent concrete were taken from the mid points of the upper, outer edges of block C-10. These points are shown in Fig. 31 and summarized in Table 3.2. The sample locations were outside of the region of graphite effects. Cocparison of columns C and D in Table 3 2 indicates that observed values of activity concentrations in concrete adj acent to steel were less than predicted by the spherical shell model by a ratio of about S. The effect is attributed to a local neutron flux depression caused by the relatively large neutron absorption cross section of iron. The effect is most pronounced when the steel "shadows" the concrete as is evident in the observations of a rectangular beam port plug in block C+D. Tne plug constructed of heavy ocncrete was encased in a 0.25 inch steel plate jacket, and the ratio of activity concentration in the steel Jacket to that of the encased concrete was about 30.
Fig 3.1 Total Radioisotope Concentration -
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Table _3.2. _ Samples of Steel and Concrete from Block C-10. Distance Steel Concrete Ratios a C d pe A R (inches}b B C B/C B/D N 56.64 375 375 983 1.0 0.4 S 56.64 310 335 983 0.9 03 E 48.41 1260 1086 24123 1.2 0.05 W 48.41 3860 2870 24123 13 0.16 8 The edge orientation of the sample identified by the compass direction from the center of the block to the edge. b Tne distance of the sample point from the center of the core, inches. C The observed Cobalt-60 concentration in the steel,,pC1/g. d The observed concentration (total) in the concrete, pCi/g.
- The concentration of concrete at R inches as predict 9d by the shell codel, pC1/g.
Th us , it was concluded that concrete samples taken in close proximity to steel would be unreliable for estimating the activity concentration of bulk concrete at the same location. A semi quantitative argument for a constant, average ratio of Cobalt-60 concentration in steel to that in .oncrete exposed to the same flux can be made by assuming that the iron L. magnetite and the iron in steel both carry about tge same atomic ratio of cobalt to iron. To achieve a density of 200 lbs/ft , the concrete cust be loaded with about 26 weight percent iron (in the form of magnetite) and will have a Co-60 concentration of about 26 percent to that of steel under similar irradiation conditions. From the horizontal beam port sections, Co-60 comprised about 9 percent of the total activity found in concrete. Thus, the ratio of Co-60 in steel to total activity in concrete should be about 0.09/0.26 or 0.35, a number in reasonable agreement with Table 3 2, column B/D, sples N and S. Thus, a modifying f actor of 0 35 was employed to estimate the radioisotope concentration in the concrete of blocks C-8 and C+ D (see Figure 2.2) . The most direct observations are of the steel, and the concentration of the concrete at that location in the absence of steel, is taken as 1/0.35 or the Cobalt-60 concentration of the steel. 3 3 2 Craphite Effects: An overview of the effect of the graphite volumes in the reactor core is shown in Figures 3 2, 3 3, and 3.4. Finer details of selected regions are shown in Figures 3 5 and ~ 3.6. The region of graphite influence is taken to be that represented by a projection from the core center to a point within the region which lies partially within a graphite volume. For a point in such a region, the exponent of equation 1 is replaced with : Verfrea.l C-( I
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-R+R)-yg R g (2) where, y, R,o and R are the same quantitles as in equation 1. R g is the portion of the distance which lies in graphite and v is an eTrective attenuation coefficient in graphite along that path. For the thermal column region, y equals 0.1163 per inch and in the vertical direction y equals 0.0289 per inch.
The value of v was determined from the concentration computed from a radiation measurement of the end of a beam port plug in block C+D which, when combined with observations of the relative intensities of the Co-60 and fu-152 spectra, yielded a concentracion of 30 nC1/g of Cobalt-60 in the steel. Using the modifying factor 0.35, the activity concentration in concrete at that location (in the absence of steel) was estimated to be 86 n01/g. With R equals 74 inches and R g equals 44 inches, the value of y equals 0.1163 satisfies equation 1 using the exponent form in equation 2. The value for y in the vertical direction was determined from the measured Cobalt-60 concentration of 130 pC1/g using a 60 gram sample of steel cut from the lower end of the tenter vertical hole liner at 66 inches above core center. Using again the factor of 0.35, with R - 68 inches and Rg - 10 inches, y - 0.02S9 per inch. Neither value of v corresponds to the theoretical value of 1/L for thermal neutrons in graphite. In both cases the graphite volume should be viewed as a channel with leakage and absorbing walls, thereby increasing the attenuation (v > 1/L) . The value y as 0.0289 per inch for the vertical region is somewhat too small for this interpretation. The graphite contained three vertical penetrations (irradiation ports). Due to poorly machined structure of these ports, the graphite was probably more transparent to neutrons than a solid structure of the same volume. 3.4 inventory of Radioactive Isotopes in the Fixed Concrete (excluding l embettents) of the Biological Shield ' Tne inventory of radioactive material in the concrete of the biological shield was deterrcined by numerical integration along the radius of the spherical shell model taking into account each fraction of the spherical area wnich intersects concrete. The results are shown in Table 33 The ! table includes the radioactivity concentration, total activity, and I concrete volume for all concrete above the reactor floor level as well as I the 52 cubic feet which is 22 inches below the floor level. Note that the inventory (total activity) in the table does not include contributions from embedments which are localizea and are identified separately in the following section and in Table 3 4. l 1
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Table 3 3 Radioactive Material Inventory in the Concrete (excluding embedments) of the Biological Shield and Substructure r C(r) Inventory V-l uge (inches) (pC1/g) (pC1) (ft") _r . . 74 1.222(-06)a 3 370(+00)D 34.0 70
- 5.675(-06) 3.855(+01) 160 3 66 2.~644(-05) 1.'823 ( < 02 ) 27'.3 62 1.236(-04) 7.669(+02) 368.0 .
58 5.607(-04) 3;129(+03) 451.5 54 2.739(-03) 1.261(+04) 522.8 50 1.299(-02) 4.936(+04.' 561.7 46 6.207(-02) 1.775(+05) 625.6 42 2.986(-01) 5.716(+05) 654.1 36 1.450(+00) 1.637(+06) 670.5 34 7.116(+00) 4 336(+06) 678.9 32 1 565(+01) 6.619(+06) 681 3 30 3.544(+01) 9.667(+06) 682.1 29 5.307(,. ) 1.123(+07) 663.0 28 7.959(+01) 1.165(+07) 683 1
" Read as 1.222 x 10-06 D Read as 3 370 x 10+00 Table 3 4 Radioisotope Inventory - fixed Concrete Isotope Half-lir'e Concrete Steel Total (years) 5 (C1) 1 (C1) (C1) '
Mn-54 0.665 03 0.04 0.4 0.01 0.03 Co-60 5.27 9.0 1.07 18.1 0.16 1.25 Cs-134 2.06 0.2 0.02 -- -- 0.02 Eu-152 13.6 13 7 1.62 -- --- 1.62 Eu-154 6.6 ~1.0 0.12 -- -- 0.12 H-3 12 3 61.8 7 32 -- -- 7 32 C-14 5730.0 14.0 1.66 -- -- 1.06 Fe-55 2.7 ~-- ~-- 75.1 0.76 0.76 Te-99 214000.0 -- -- 6.5 0.07 0.07 100.0 11.65 100.0 i.02 12.87 l 1
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3 5 Embedments in Fixed Concrete The major iron or steel embedments in the thled concrete are six beam por't liners and the remaining (non protruding) portion of the reactor framework _. Other embedments (rebar and electrical conduit) are diffusely distributed.' Tne mass is small relative to the mass of iron in the magnetite concrete, and these lesser embedments are not separately analyzed. The contribution of embedments to the radioactive material inventory is based upon. measurements of steel samples from the upper edging ~ strips of block C-10. The dominant _at ta-emitting radioisotope found in structural steel is -Cobalt-60 wi th Ha. .sese-54 present, to the extent- of about 2% of the Cobalt. The Europium was found in concrete, not in steel. 3.5.1 Monolith Embedments: The principal embedments localized in the c.onolith are the six steel beam port linces. Two are six inches in diameter-and aligned along ~,he north-south horizontal center line of the core. The other four are 4.625 inches in diameter and enter the core diagonally near the edges which ~ join the four vertical faces of the core. The diagonal ports are treated approximately by assuming that they lie along radial lines from core center, but differ pairwise in the distance of tueir closest approach to the core center. The spherical model yielded an estimate of the total activity in concrete columns substituted for the li_ner_s, and these are converted to Co-60 activity in the steel using the 0 35 f actor previously described. The relative fractions of radioisotopes in steel given in Table 3.4 were found froa, analyses of steel samples (south beam port) on a Liquid Scintillation Beta Spectrometer and a Ge(Li) Spectrometer. Based on the relative percent of Cobalt-60 in steel, the N and S beamport J iners were fou'Id to have a
' total activity of 9,37 mci shile the other four were determined to have a, total activity of 1 38 mci.
3.5.2 Pedestal Embedments: A concrete base or pedestal is located beneath the reactor structure. The maj or embedment in the pedestal was the remaining (non protuding) part of the control blade support structure. Its components were 200 inches of American Standard steel channels (five inches by 6.7 pounds / foot) and 42 inches of five inch angle iron, (10.25 pounds / foot). The embedment is flush wi th the pedestal surface. The inventory was estimated by calculating the inventory in a concrete block 24 inches wide by 76 inches long by five inches high. If placed wi th ' the center directly below the' core _ vertical center line, the average conce;1tration was estimated to be 7.04 pC1/g of concrete. Using the praviously mentioned factor of 0 35,' the corresponding Cobalt-60 concentration of steel disposed in the same geometry would be 2.67 pCi/g. The total activity in that steel would be 14.77 pCi/g. The total mass of the embedded framework is 148 pounds and the total activity of the two channels is 1 C1. Tcble 3.4 gives a summary of the total radioactive material inventory in the fixed concrete. 1
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- The residual concrete biological shield consists of a monolithic structure
-M and four removable blocks. The characteristics of the bloct:s- are shown in Table 3.5.- The,prinicpal embedments in the blocks are stee' in the form of reinforcing bar, angle iron edging, beam port liners? and plugs.
Table 3.5.' Concrete Blocks df the Biological Shield Block ID- Size Density Weight Inventory (inches) Voly)e (ft (1bs/ftS) (tons) (Total) l l A+B #177 69x66x20 52.71 258 7.11 < 22.0 pCi C+ D #177 69x66x20 52.71 258 6.77 0.03 C1 C-8 #289 90x60x30 93;75 213 10:00 - 0.07 Ci ! C-9 #293 84x48x30 70.00 '258 8.94 4.57 Ci 4.67 Ci-l The total on-site inventory at initiatiUn of. Phase II is shown in Table l (, 3.6. The spectral analyses are as of February, 1988. - 1 Table 3.6. Total Radioisotope Inventory Nuclide Half-lif e Concrete Steel ~ Total (years) % Ci % C1 Ci Mn-54 0.865 0.3 0.05- 0.4 0.01 0.06 Co-60 5.27 9.0 1.4 16.'1 0:3- 17 l cs-134 2.06 0.2 0.03 -- -- 0.03 Eu-152 13.6 13 7 2;2 -- -- 2.2 ! Eu-154 '8;6 ~1.0 0.2 -- -- 0.2 ) H 12 3 61.8 9.8 . 9.8 ) C-14 5730.0 14.0 2.2 -- -- 2.2 l Fe-55 2.7 -- -- 75.1 13 13 i To-99 214000.0 , -- ,-- 6.5 0.1 . 0;1 100.0 15.9 100.0 -1.7 17.6 18 - n __,.,-._.-._-,,--....,._m____,.-,.,-,_~.~_...r , _ . . . . . ~ . ...-...-.,_.._,,_,_,.u- ,,,,,-m-_,_y_._...._.,,y. --.
4.0 HEALTH ANb SAFETY CONSIDERATIONS The dissantlement work was controlled by the UCLA Radiation Safety Office with a series of Radiation Work Permits that described the current
, radiation environment and the ~ tasks permitted. Personnel exposures were tracked with pocket dosimeters ,used by personnel- for up to one day.
Radiological instrumentation maintenance was the responsiblit/ of a part of the Radiation Safety Office. Routine calibrations of survey instruments were performed and documented. Periodic source checks of instruments use.d in decommissioning' were performed. Liquid scintillation, whole body, and
~
Ge(L1) counters were checked prie 'to each operation to assure that they were operating within past historical parameters. Only NBS traceable sowces were used in calibrating these instruments and in their operational "QA" checks. 4.1 Collective Dosa-Equivalent Table 4.1 shows the estimated and actual collective dose-equivalent of the personnel for various Phase I tasks. The actual valuec were derived from a commercial, accredited, personnel radiation badge service. Tasks 4 through 6 were performed concurrently and the breakdown given is an estimate for the individual tasks. The highest individual dose during- Phase I was 1.24 r em . Note that bioassays (whole body counting and urinalysis) showed'no i mcasurable internal intake of radioactive material. Table 4.1 Collective Dose-Equivalent (Person-Rem) of Personnel for Various Tasks Task Activity Estimated Actual
- 1. Planning and Review 0 0
- 2. Mobilization 0 0
- 3. Removal of External Equipment. < 1.0 =0 3
- 4. Diassemble Reactor Core 7.0 1.9
- 5. Package Materials for Transport 2.0 0.4
- 6. Transport Materials from Site 0 0.1
- 7. Assess Radiological Status _.1.0_ 0
= 11. 0 2.7
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4.2 Area and Surface Surveys - Periodic surveys were performed through out_ Phase I -through the use of portable'. survey meters, various counters for collected swipes including _a liquid scintillation counter, and other radiatjon tsonttoring equipment to assure that the surface contamination levels of released material stayed well below the "Removable" values of Table 1 in NRC Reg. Guide .1.86. Af ter the removal of the reactor internals, the residual radiation level was about 1 R/h at floor level and 600 mR/h at waist level in the center of the f,ormer' core region. These levels are attributed to structural steel (80 pounds of 5 inch channel) embedded in the underlying concrete pedestal and the 6 inch diameter steel beam port liners embedded at the horizontal mid- , plane of the biological shield. At the conclusion of Phase 1, swipe sample measurements of the drainlines, fuel storage pits and decontamination facilities all showed surface levels well below the "Removable" values of Table 1 in Reg. Guide 1.86. Due to the _J potential for spreading _ radioactive material contamination' in Phase II, sludge sampling of the sump will be performed during dismantlement and removal of the concrete structure. 4.3 Particulate Monitoring: Two high volume air samplers and a stack effluent monitor were used through out Phase I to monitor radioactive particulate flow. Protective clothing, , respirators were employed as required. ' I I l l l l
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. " . .*e 5.0 FINANCIAL COSTS AND FLITURE PLANS The Decommissioning Plan for Phase I stated an estimated cost of $65,000, exclusive of internal administrative and supervisory ~ costs. The actual ' . costs were more nearly $23,000, exclusive of internal personnel costs. The principal cost items were purchased waste packaging, special tools, hoist slings, shipping and land burial. Personnel costs based upon 4.5 man years l (7/1/85 to 1/1/87) suggest an additional cost of $180,000 for worker time.
For Phase II, SAFESTOR is no longer under consideration.at this ^ time. UCLA . intends to complete dismantling and decommissioning via the DECON route and will rely on- the successful responder to a Request for Proposal to be distributed in the near future. The School of Engineering and Applied Sciences has approved the estimated cost of about $200,000 (incrementally above the cost of. Phase I). The Contractor's task in Phase II will include the following:
- 1. razing the concrete tronolith and pedestal to floor level, 2.-excavation of the approximately 52 ft 3 op . concrete down to about 22-inches below floor level,
- 3. removal of four large concrete blocks,
- 4. packaging materials for appropriate land burial,
- 5. transfer of such material.to burial site,
- 6. decontamination of the reactor rooci and nearby areas, and
- 7. final release survey of the facility.
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