ML20127K282

From kanterella
Jump to navigation Jump to search
Final Decommissioning Rept for Boelter Reactor Facility Dismantlement & Final Decommissioning Project
ML20127K282
Person / Time
Site: 05000142
Issue date: 12/31/1992
From:
CALIFORNIA, UNIV. OF, LOS ANGELES, CA
To:
Shared Package
ML20127K274 List:
References
NUDOCS 9301260087
Download: ML20127K282 (61)
v • d • e


Text

_

FINAL DECOMMISSIONING REPORT FOR TIIE IlOELTER REACTOR FACILITY DISMANTLEMENT AND FINAL DECOMMISSIONING PROJECT N

h

_ g*,a.% OY(

  • i!!i<

/ SWil .=.: r:=511::: ..~ / I i O [ + IlCLA 1 UNIVERSITY OF CALIFORNIA LOS ANGELES, CALIFORNIA DECEMBER 1992 i j ) 9301260087 930104 DR ADOCK 0500 2

i FINAL, DEC051AllSSIONING ltEl'OllT t F O lt Tile t I!OEllrEll ItEACTOlt FACil,lTY DISAIANTI EAIENT AND FINAL, DEC051MISSIONING l'RO LECT 1 l DOCKET NUA111Elt: 50142 I f CENSE: 11-17 1 i i i 'T 'Op $ LWO A g eri.e 1 SEVEE i ~/ m.v.: w 'Dd N[# iii@. 'T1

- avr u

'o ?i)? /, (:=2si IGHT UCLA i t l i Report Prepared By l l Scott State, Dames & Moore Amir Huda, UCLA Joseph M.Takahashi, UCLA UNIVERSITY OF CALIFORNIA l LOS ANGELES, CALIFORNIA DECEMBER 1992 i

Table of Contents Page No. INTRODUCTION. I tlACKGROUND 2 Reason For Deconunissioning........................ 2 Management Approach.. 2 Schedule......... 3 SITE DESCRIPTION............ 5 Type and Lt.c ation of Facility............................................. 5 O wne rs hi p........................................................... 5 Facility Des cript ion.................................................... 5 Buildings....... 5 Grounds 9 OPERATING !!! STORY............ 10 Licensing and Operations....... 10 Prcc e s se s Pe r fonned.................................................... 10 Waste Disposal Practices............... 10 DECOMMISSIONING ACTIVITIES... 11 Objectives............... 11 Results of Previous Surveys....... 11 Project IIcalth and Safety Program..... 16 Decontamination and Demolition Procedures............................. 16 Routine Activities 16 Operational Surveillance........................................... 16 Air Sampling Surveillance. 17 Radiological Protection.......... 17 Personnel Exposure Monitoring ,,................................... 18 Waste Screening 18 Specific Project Tasks......... 18 Removable Shielding B locks........................................ 18 Decontamination & Removal of Shower & Sink 19 Monolith Containment and 11VAC Contamination Control 19 Steel Support Rail Removal............................ 19 B eam Port Removal.................. 20 Activated Concrete Removal 20 Major Contaminants identified.,..................................... 21 Activated Concrete Packaging 22 Process Pit Containment. 22 Process Pit Holding Tanks Removal. 22 Process Pit Piping Removal. 22 Removal of Sludge in Sump........................................ 23 Process Pit Decontamination 23-Clean Concrete Removal..... 23 Contaminated Drain Line Removal 24

LSA Shipments 24 Final Site Cleanup. 25 FINAL SURVEY PROCEDURES. 26 Sampling Parameters 26 Background /llaseline Levels identified... 26 Guidelines Established...... 27 !h uipment and Procedures Selected.. 27 t Instruments and Equipment... 27 Instrument Use Techniques 28 Procedmes Followed 29 Surveying Organization.. 29 SURVEY FINDINGS......... 30 Techniques for Reducing / Evaluating Data.................. 30 Statistical Evaluation....................... 32 Comparison of Findings with Guideline Values and Conditions....... 32 Control Room (Room 2001) 33 Count Room (Room 1005) 33 Storage Room (Room 1003) 34 Reactor Room Floor (Room 1000) 3% Reactor Room Catwalk.... 35 Reactor Room Menanine... 36 Reactor Room Ramp..... 36 Reactor Room Walls. 36 Process Pit....... 37 Fuel Storage Pits............. 38 Floor Drains and Piping inlet / Outlets............ 38 Soil S am ple....................... 39 Verification of Surveys by UCLA Radiation Safety OfGee......................... 39

SUMMARY

41 REFERENCES. 42 APPENDICES l

INTRODUCTION This repon describes the final phase of decommissioning activilles including release suncys that were perfonned at the UCLA Argonaut Reactor facility for its release for unrestricted use. The Final Phase activities were undenaken in accordance with the requirements of Section 2.5 of the Atomic Safety and Licensing floard Settlement Agreement dated September 30,1985 and the NRC l'hase 11 Order dated July 28,1989. 4 ) ) ) ) ') ) i 1

) IIACKGROUND Reawn For Decommissioning ) UCLA announced its intca: to decommission the lloclier Research Reactor on June 14, 1984. Chancellor Charles Young noted that the deconunissioning decision was reached solely as a result ) of the changed circumstances affecting the academic benefits and escalating costs of continued operation of the reactor facility in a press release that day. Management Approach ) The decommissioning of the UCLA Reactor facility was carried out in two phases. The Decommissioning Plan for Phase I was submitted to the Nuclear Regulatory Commission (NRC) on October 29.1985. Phase I commenced after approval of the plan by the NRC on July 14, ) 1986. A report on Phase i decommissioning activities was submitted to the NRC on April 12, 1988. The first phase involved removal of the following:

  • reactor core moderator a graphite thennal column

) . shield tank . peripheral equipment a fuel boxes . control blade system components . protruding pans of pipe and structures . most of the concrete shield blocks ) The Phase !! Plan was submitted to the NRC on June 10,1988. The NRC requested additional infomiation in letters dated October 12,1988 and March 14. 1989. Supplementary infomiation wa's provided by UCLA in letters dated December 7,1988 and March 31,1989. The NRC Order ) for Phase 11 was issued on July 28,1989. The second phase included the following: . demolition of the reactor monolith . removal of the remaining shield blocks . removal of all process equipment . decontamination of all remaining facilities ) = a tennination survey The final survey was perfonned by Nuclear Energy Services (NES) with oversight pmvided by the UCLA Radiation Safety Oftice. Figure i presents the overall organizational structure for the project and final survey. UCLA retained Dames & Moore to provide radiological engineering 2 )

O oversight and to ensure compliance with on site health and safety procedures. QA/QC activities were perfonned independently within the UCLA Radiation Safety Office. IndepenJent QA audits were also perfonned by both Dames & $1oore and N!!S. O The management team pissessed sulncient training and expenise to successfully complete the project and to ensure worker and general public health and safety. The UCLA Radiation Safety Officer, hir, Joseph Takahashi, served as the Owner's Representative. hir. Takahashi is a certified health physicist and is responsible for campus wide radiation safety. The QA/QC manager was hir. Amir lluda. hir.11uda possesses a h1 asters Degree in Nuclear Engineering and has worked in the UCLA Radiation Safety 010cc for over 4 years. The radiological engineer, hir. Scott State, Oi was provided to the project by Dames & hioore. h1r. State has a h1 aster Degree in Nuclear Engineering, is a licensed professional engineer, and also previously held a reactor operator's license on the Argonaut type reactor. O The NES management team consisted of hir. Lee Penney as the Site Supervisor, hir, William Needrith as the Operations Supervisor, and hir. Eric Alelquist as the licalth Physics Supervisor. hir. Penney and hir. Needrith have acquired extensive D&D experience over the past several years including the D&D of the Berkeley Research Reactor. hir. Abelquist possess a hiasters Degree O in llealth Physics and has previous D&D and radiological controls experience. Schedule O NES began mobilization on August 3,1992. Ilealth and Safety training was performed during the week of August 10. Decommissioning operations commenced on August 13, with activated concrete removal. This was completed on September 22,1992 with the decontamination of the process pit. The final release survey began on September 19,1992 and was completed on October 15, 1992. Appendis A details the project schedule. O O 3 9

) ) Chancellor ) Executive Vice Chancellor Administrative Vice Chancellor Capital Programs Vice Chancellor 1 Dean, School of Engineering Assistant Vice Chancellor Director of ) & Applied Sciences Business Operations Administrative Officer Radiation Safety Officer Senior Project Manager Owner's Representative Capital Programs 1

7....................a l

QA/QC Manager Radiological Engineer, NES, Site Supervisor +------- +--- Consultant ) Operations licalth Physics Supervisor Supervisor k

Penhall, r'

Subcontractor Supervisor i > refers to the flow of documentation generated by NES/IES. Figure 1: Iloelter Reactor Final Decommissioning Project Organization Chart t 4 r

1 SITE DESCRil' TION Type and Location of Facility The facility that was decommissioned was a 1(X) kW Argonaut type research reactor. General reactor research and teaching were perfonned at the facility including activation analysis and reactor physics instruction. The facility is located on the UCLA campus in Los Angeles, Califomia on the ground floor of Boelter llall. Figun:s 2 & 3 provide a generallayout of the reactor facility and associated work areas. Those facilities that were within the scope of the D&D project are shaded. Ownership The reactor facility is owned by the University of Califomia and was licensed with the U.S. NRC under a facility license R 71. Facility Description The facilities that were decommissioned were fully housed within Boelter IIall. In the case of this project, there were no other gmunds. Buildings D The Reactor Facility is a 2 story, reinforced structure approximately 75 x 49 feet in plan and 27 feet high. Major columns for the Reactor Facility rest on 10 x 10 foot poured, reinforced concrete footings. All exterior walls are load lxaring and are of 12 inch thick, reinforced concrete, faced with Roman brick where exposed. D D 5 1 D

O Z A d e a ~= c:) Q TOKAMAK 1000A e

)

d

)

O L O e=. IRl R- [ f u L { REACTOR {_g 1000 I r g \\ O O 3 ct3 q 'N / i L. J l ') C: ~ l g 1003 l Storage t l -1005-Room. O _j. Counting g 7 Room 1(XM l Transformer Vault Jl n C c 0 e Shaded Area: 2"d Phase of Decommissioning Project Figure 2: lloelter IIall First Floor, Reactor Area O 6 6 -~

Z i!!!!in ;

Pi j, lli i

hh:ll! 'l l

)

.) TOKAMAK

)

n ' ] 3 liiilmilimliululin 4 1- [ int !l!I KNH I T I: ,.I g) b 4W r REACTOR e 3 1 L e O O ( :) ci s pv y 0:- b!!$hk$sua.M v

)

v n+ EfdM!I:sMi )_ -u 001 N'introl

  • =O"

'=O=" ~ ~ '~~ l Room I r.3 r3 g Lightly Shaded Area: 2*' Phase of Decommissioning Project Figure 3: Iloelter llall Second Floor, Reactor Area 7

O The interior wall separating the reactor high bay f rom the remainder of the building is also load bearing and is of reinforced concrete.18 inches thick. All other interior walls are nonbearing, S inch. cement block walls. 4 A process equipment pit is located directly north of where the reactor was located. This pit is 9 x 12 feet in plan exclusive of an extension under the floor to the west that contained two retention tanks. Directly east of the process pit is a 5 x 6 matrix of 30 galvanized steel lined fuel storage pits buried in concrete. Each pit is 78 inches deep and is stepped once at 30.5 inches from the pit bottom. Below the step, the pit inside diameter is 8 inches. Atove the step, the pit inside diameter is 10.25 inches. O The two adjacent rooms on the first floor of the building numbered 1003 (storage room) and 1005 (count room) were also pan of the decommissioning project. 'These rooms contain under 1000 square feet of combined floor space. Each room has a smooth e concrete floor. On the second floor of the building is the fonner control room and a locker room which were included in the scope of the decommissioning project. The control room is less than 500 square feet and contains a tiled floor. The adjacent locker room contains a shower and bathroom with primarily ceramic surfaces. O None of the areas that were decommissioned contained expansion joints or floor penetrations that were difficult to decontaminate. Piping used for drain lines is contained O within the floor of the reactor high, bay. The openings of the piping were surveyed as described later in this rerwn. All areas except the reactor high bay were decontaminated in such a manner that wall and g floor surfaces were not damaged. In the high bay, it was necessary to demolish the reactor monolith which resulted in leaving a significant ponion of the floor area in a pitted condition. After release of the facility for unrestricted use, the high bay ikor will be restored to a condition suitable for using the facility for other research purposes. 0 8 9

) Grounds ) No grounds outside Boelter llall were within the scope of die decommissioning activities. The area decontaminated is not subject to concems related to geology, hydrology, seismology, meteorology, or population with respect to the extent of contamination. Southem Califomia is seismically active but no major events have occurred which would I affect the area or extent of radiological contamination. The physical size of the areas surveyed was fully within the building area as described in the previous section. D D D D D 9 D 1

) OPER ATING lilSTORY ) The operating history of the reactor extends from initial criticality on October 21.1960 to the notice of intent to decommission by Chancellor Young on June 14. 1984. Licensing and Operations ) The facility was operated from the time of initial criticality until the decommissioning order for the purpose of n:scarch and teaching. Due to the academic mission of the reactor,it was operated ) "as needed" for class instruction and experimentation. As such, it was rarely operated for more than a few hours in any one day. The reactor was licensed by the NRC (license R-71) prior to inillal criticality with a limitation on i ) power output of 10 kilowatts. Slight modifications were made to the reactor and licensing amendments were approved that allowed operation up to 100 kW in October of 1963. All licensing and operations records have been retained in the fonner control room, b Processes Performed The processes performed in the reactor facility were limited primarily to reactor operation. The purpose of the reactor operation was teaching and research as described previously. The by- - ) product of the operations was activation of, reactor core and structural material and radioactive contamination in the reactor process system. ) Waste Disposal Practices Waste disposal practices were limited to those associated with the nonnat operations of a facility of this type. No waste disposal practices impacted the decommissioning status of the facility. ) ) )' . ~

DECOMMISSIONING ACTIVITIES Decommissioning activities were perfonned by NES and their subsidiary IES with oversight from the UCLA Radiation Safety Office. All significant activities associated with the project are described below. Objectives The objective of the decommissioning effon was to dismantic and remove the Boeller Research Reactor, decontaminate the reactor room and ancillary facilities, and release the facilities for unrestricted use. Results of Previous Surveys A walk-through survey of the Boc!ter Reactor facility was perfonned by NES prior to initiation of the final decommissioning to confinn radiological conditions that UCLA had reponed in the interim between Phase 1 and Phase 2 activities as shown in Figures 4 through 7. Each shield block was surveyed for radiation and contamination levels and the results wen: documented. No smearable contamination was detected on any of the shleiding blocks. Shiciding block C 9 exhibited the highest radiation level with a contact exposure rate of approximately 40 mR/h. The five shleid blocks that remained in the facility after Phase i effons were roped off and the area posted as a " Radioactive Materials Area." A general area removable contamination survey was conducted to detemiine the protective clothing requirements. A baseline expo ure rate survey was also perfonned thmugh out the facilities to detennine the proper radiation trea posting requirements. A detailed radiation survey was perfonned within the monolith containment to detennine radiological conditions present during initial demolition in this area. A removable contamination survey was performed to identify the nature and extent of removable contamination within the monolith. Baseline air samples were also obtained from within monolith contairunent. Both the smears and air samples were at background levels. I 11 L

O l l 1 N 5 10 pIUh 4,y, 4 10 pR/h 9 i 9 i c e ea ROOh! 1005 4 9 3-9 pl W G Work Ilench E G a 2 Work Hench d 310 pl% 3 8 pR/h I E 4-11 uR/h g l410 lpR/h ,c da 2 ta ROOh! 1004 ROOh! 1003 O Transfonner Vault fe Y b 5 3 1 4 e 1 l o g l 5-14 pR/h O O u hieasurements made with Victorcen 450P, S/N 1366, on April 24,1992 at one meter above foot level. 4-11 pR/h 0: Swipes counted on Liquid Scintillation Counter. All < h1DA (30 dpm) l Figure 4: Radiation Levels in Storage Room,1003, and Counting Room,1005 g 12 9

) ) -1 I r 3 )- ROOM 1000 46 55 pR/h -O ) Fuel , _m.,1,, Storage nim + . Beam Ports Pits kernnynhle Birrk nA Covered ~ Process -74 mR/h O 16 25 pR/h

]

Pit g-g ). 4 ) 3.3 mR/h O-0 W 3:. 0.89 mR/h l o I

  • l Measurements made with Victoreen 450P, S/N 1366,' on April 24,1992 at one meter above foot level.

0: Swipes counted on Liquid Scintillation Counter. All < MDA (30 dpm). ). Figure 5: Radiation 1.csets in Reactor Room,1000 - 13 l-._ I

a O: e' J I r 714 pR/h 230 pR/h f 3 e' Fuel 9,, m,, s..,w,, Storage mm+ Pits acmovabic 131ock N e: A Covered . Removable g@ Process 25 36 pR/h 40 50 pm } Pit O Illock e. ~ q w / 3 g,rw; e I I ]O - 10M mRA) 1 r i I i i i i e l l Measurements made with Victorcen 450P. S/N 1366, on April 24,1992 at one meter above foot level. O: Swipes counted on Liquid Scintillation Counter. All < MDA (30 dpm) -e' ~- Figure 6: Radiation I.esels in Reactor Room, Upper I.evel l l l l l e l 14 T e u l

) -f I l l I 410 pR/h 411 plyh ) 4-10 pR/h ) ~ ROOM 2001 0 513 pR/h ) 415 pR/h ) 412 pR/h 410 pl% .c il ) g V U 4 C Change Room J U 410 pR/h { .310 pR/h )-

l 1

i Measuret is made v'ith Victorean 4;.0P, S/N 1366, on April 24,1992 at one meter above foot level, g O: Swipes counted on Liquid Scintillation Counter. All < MDA (30 dpm) Figure 7: Radiation I.evels la Control Room,2001 p 15

O A radiation survey was perfonned to identify contact exposure rates on four sinactural steel rails within the pedestal of the monolith. IJsing a teletector, the highest contact es[we rate sneasured was 11(O mR/h. This reading was in agreement with results of previous surveys. O Project Ilcalth and Safety Program All NES personnel, including on-site subcontractors, were trained in accordance with the NES Radiation Worker Training manual and received a general safety briefing, in addition, all site personnel were given an orientation on University specined guidelines by a University representative. llealth and Safety Site Specine Training was administered to both NES and Penhall personnel. A fifty question test was given to all personnel at the conclusion of the training. O Decontamination and Demolition Procedures The decontamination and demolition activities were perfonned by NES through their subsidiary O lES. IES in tum hired subcontractors for concrete demolition (Penhall Company) and waste transportation (Environmental Management & Control). NES/IES has successfully perfonned numerous radiological D&D operations including a similar project at the fonner Berkeley Research Reactor All activities were perfonned with oversight by the UCLA Radiation Safety 4 Office with the aid of Dames & Moore whom the University retained to provide a radiological engineer to support the day to-day D&D operational efforts. Specilie activities that comprised the D&D project are detailed below, General / routine activities are first presented followed by individual work scopes in the approximate chronological order they were perftnned. This is also g shown in Appendix A. Routine Activities O Operational Surveillance Routine radiation and contamination surveys were perfonned daily within the radiologically controlled area (RCA), as well as the count room, control room, storage rmm and reactor oom ramp area. These surveys were conducted to verify that 16 9

) contamination haa not spread from the contivlled surface contamination area (CSCA) during decommissioning operations. The surveys aided in the substantiation that ) administrative and engineering controls implemented during operations were adequate. Air Sampline Surveillance ) Baseline air inonitoring was perfonned for the reactor room ventilation exhaust duct on the third floor of Boeller llall. Both pre and post air sample results exhibited no activity alsove background levels. )- Fifty one (51) air samples were collected over the course of the decommissioning project. Air samples with initial results near or exceeding iE 11 pCi/ml for alpha emitters and IE-10 pC /mi for beta / gamma emitters were recounted within 24 hours to allow for decay of ) short lived naturally-occurring radionuclides. The air samples collected during activated concrete cemolition were allless than the maximum pennissible concentration (MPC) for each radionuc'ade present. The dominant radionuclides detected were Eu 152 and Co-60. ) Eleven air samples that were collected during concrete demolition within the monolith contairunent were also analyzed with a liquid scintillation counter (LSC) to detennine the 113 and C 14 concentrations in air. The LSC analysis resulted in overestimated quantitles of It 3 and C 14 because of the presence of Co-60 and Eu 152 in the air samples. Even ) with the gross overestimate, the concentrations of 113 and C 14 were significantly below their respective MPC's. ) Radiological Protection All decommissioning work within the radiologically controlled area (RCA) was perfonned under the issuance of Radiation Work Pennit (RWP). The RWPs provided the radiation workers with ) the radiological conditions under which work in the RCA was to be perfonned. - The RWPs required the use of engineering controls and protective clothing, as necessary, to ensure that the work was accomplished in a radiologically safe manner while maintaining personnel radiation exposure as low as reasonably achievable (ALARA). The RWP was prepared by the IIcalth Physics Supervisor (ilPS) based on expected and surveyed [ 17-

O conditions. Each radiation worker making an entry to the RCA was prosided with both a film badge and a self reading dosimeter (SRD). The cotry.uid exit SRD readings were documented 8 and reviewed by the llPS on a daily basis. In addition, pre and post whole tudy counts were administered to all personnel. An NES/IES administratise limit of MK) mrem / person was established for diis project. Approval O from the Radiological Services Depanment Manager was required before projc(.t personnel could exceed diis limit. Personnel Extmure Monitoring g The University issued dosimetry and adminisiered entry and exit whole body counts for all site personnel. The total man rem for the pmject was 1.17 rem, as detenuined from commercial film badge results. The whole txidy counts were negative for all personnel from the decommissioning project. Waste Screening The liquid radwaste generated during decontanonation and demolition of the structures, equipment, and floors was collected, monitored, and then released to the sanitary sewer, if under the limits of 10CFR20. The major bulk of solid waste, such as concrete rubble, metallic embedments, etc., was surveyed according to the release criteria of Regulatory Guide 1.86 and the 5 prem/h over background at i meter. Based on the results, dicsc v.cr either packaged and shipped to a licensed burial site or disposed as clean waste. e Specific Project Tasks Removable Shiciding Blocks O The live removable shield blocks were removed from their positions on the monolith and l placed in the southeast comer of the reactor room. They were subsequendy shipped off-6 site for disposal as radioactive waste. l gg

Ikcontamination & Remo"al of Shower & Sink A shower stall and sink were removed from the reactor room (refer to Figure 5). These two items were sun eyed to detennine the extent of any contamination. The sink required minor decontamination with a cleaning solution and was released for clean waste disposal following a survey, The shower stall was surveyed and disposed of as clean waste. Monolith Containment and ilVAC Contamination Control The monolith contamination control envelope was constructed with 2"x 4" framing and 10 mil plastic sheeting. The plastic was fastened to the monolith with spray adhesive and duct tape. Two high efficiency particulate air (llEPA) units were connected to the containment Adequate ventilation was ensured by measuring for negative pressure within the containment. After the activated concrete was removed, all materials used for contamination control were disposed of as clean or contaminated waste as appropriate. The reactor room heating, ventilation, air conditioning (IIVAC) exhaust duct on the east wall was sealed with two llEPA tillers. The four HVAC ceiling inlets were covered with cloth media to protect the system from contamination. Steel Support Rail Removal A detailed radiological survey was performed inside the monolith prior to the n'moval of the four steel rails imbedded in the pedestal below the location the core was previously positioned. An ALARA review and mock-up training were conducted for personnel perfonning the removal of the steel rails. A remotely operated demolition ham'mer was used to break up the concrete surrounding the embedded rails. The portion of the rails not being worked on was shicided with lead bricks, The steel rails were kmsened and eventually freed from the concrete by the demolition hammer. The demolition hammer was used to dismantic the four separate pieces at each comer. An oxy-acetylene cutting torch was then used to cut the rail into manageable. sectiorts. ) 19

nV The rail pieces were double-bagged and taped. Workers handling the steel rails v. ore finger ring dosimetry. The rail sections were passed through the monolith containment and placed in the process pit for temporary storage. The process pit covers were placed over 9 the pit and the surface was surveyed to ensure that the process pit concrete covers provided adequate shielding. The steel rails were later disposed of as low specific activity (LSA) waste in LSA shipment #3 as described later in diis section. Beam Port Removal 9 The six experimental beam ports that penetrated through the monolith wem removed from outside of the monolith using concrete core boring equipment. The water generated during the coring operation was collected in 55 gallon drums, sampled and disposed of 9 appropriately. The coring equipment cooling water was recirculated to the extent possible. Following removal, the beam ports were surseyed to estimate the depth of concrete activation inside the monolith. This activation depth guided subsequent activated concrete O removal. The beam port cores were disposed of as LSA waste. O Activated Concrete Removal As detemiined from the beam port characterization, a predetermined amount of concrete g from the monolith surface was removed. The amount of concrete remc.ved was based on an assumed spherically symmetrical activation pattem. The concrete was removed using a hiini htax hydraulic ram. Water was continuously sprayed on the concrete being removed to minimize dust generation. Reinforcement steel encountered during the O concrete removal was cut as necessary with an oxy-acetylene cutting torch. Once the predetennined amount of activated concrete had been rerouved, the monolidi was 9 surveyed with a Ludlum 2221 and 44-40 shielded Geiger.htueller (GN1) pancake pmbe to better characterize remaining activation locations. With the hot spots identified and 20

? marked, funher activated concrete removal was perfonned. This iteration of surveying the monolith and decontaminating hot spots was continued until the release criteria werc ) achieved. Maior Contaminants identified ) The major contaminants were products of neutron activation due to reactor operations. Based on laboratory characterization, Co-60 and Eu 152 were found to be in the highest concentrations. A concrete chip representative of the interior monolith walls was collected ) and analyzed. The analysis identified the following radionuclide content: Isotope Activity (nCi/c) 'FI 1,494.00 ) "C 502.00

  • Co 2,575.00

'"Cs 3.82 '"Eu 5,481.00 2"Eu 293.00 5"Eu 45.00 ) A sample of the steel rebar from the monolith was also analyzed and resulted in 19,600 pCi/g of Co-60. ) A concrete sample was also taken from one of the removable blocks (block C-9) an.1 similar results were obtained: Isotone Activity (pCi/c) 'F1 'l.750.00 "C 348.00 "Co 3,809.00 "Cs 20.80 "2 ) Eu 23,090.00 '"Eu 1,263.00 "Eu 62.00 s 21 .). -I

O! Activated Concrete Packacine The activated concrete rubble was loaded into LSA containers within the containment. 9 The LSA containers were manually filled by Penhall employees using shovels. Fully loaded containers were stored in a designated radioactive materials area (RMA) until they were scheduled for shipping. 9 Process Pit Containment A contamination control enclosure was constructed that covered the process pit. HEPA ventilation was installed within the process pit containment. In addition, a step-off area was established at the entrance of the process pit containment. Process Pit Holdine Tanks Removal 9 Two holding tanks were removed from the process pit. A small quantity of residual water was present in the holding tanks which was drained into the sump through the existing plumbing. To prevent overtlow of water from the sump, the water in the sump was 8 simultaneously pumped into 55 gallon drums. The water was sampled and disposed of appropriately. 9 Once de-wMered, the holding tanks were disconnected from the piping and mounting brackets. The tanks were then cut within the process pit containment and disposed of as clean waste following a detailed survey. 9 Process Pit Pipine Removal Water was drained into the sump from the process pit piping. The process pit piping was then removed and cut into manageable lengths within the containment. Each pipe section was placed in a drum, transported from the containment, and surveyed. All piping was classified and disposed as either clean or radioactive waste as appropriate. 9i 22 9' \\

).- Removal of Sludee in Sumn ) Water contained in the sump was pumped into a 55 gallon drum and surveyed to detennine the correct disposal option. The sump pump and associated piping were then removed. The equipment was packaged and disposed of in an LSA container. ) The sludge at the sump bottom was mixed with water and vacuumed into a 55 gallon dmm. The sludge was then allowed to settle to the bottom of the drum. The waterin the drum was pumped off and surveyed. The sludge was then solidified and disposed of as LS A waste. ) Process Pit Decontamination ) The process pit floor and walls wem surveyed after removal of all hardware. Ilot spots were identified and decontaminated. The method of decontamination was to hose down the walls and floor with water and to then remove the residual water with a wet / dry vacuum. The sump floor and walls' n quired further decontamination efforts. The floor and walls were scabbled with a bushing hammer and the resulting debris was vacuumed, collected and disposed of as LSA waste. ) Clean Concrete Removal ) The interior of the monolith was cleaned of all rubbled concrete after preliminary surveys indicated the release criteria had been met. The monolith was then gridded and "smveyed in detail for free release. The survey consisted of an initial 1007e scan of the monolith interior walls wiih the Ludlum 2221 and 44-40 shielded GM pancake probe. The hichest ) 2-direct reading was reponed within each grid. It ranged from 196 to 4549 dpm/100cm 2 on the south face and 392 to 4941 dpm/l00cm on the north face. Smears were then taken on all monolith surfaces and ~257c of the smears taken inside of the monolith were i counted on a liquid scintillation counter. Smears that were counted with the Ludlum 2929 23 1 ~

O were reponed as less than minimum detectable activity (MDA) except for an area on top of the monolith which read 447 dpm. Smears counted on the LSC showed a maximum reading of 12.3 dpm for '11. After confinnation that all activated concrete had lven 4 removed, the monolith walls were demolished and the concrete pedestal was removed. A larger, more powerful, demolition hammer was used to efficiently dismantle the " clean" monolith. Water was again used for Just mitigation. 9 The concrete mbble was removed from the reactor room with a Bobcat bucket loader to the laydown area where it was surveyed for exposure rate. The background dose rate was determined in the laydown area prior to removal of any concrete from the reactor room and then the highest Bieron Micro Rem meter fluctuation was recorded for each bucket-load of clean concrete. The concrete was then loaded into roll-off boxes, surveyed for one hour with a Reuter Stokes envimnmental monitoring pressurized ion chamber to ensure acceptable exposure rates, and disposed of at a concrete recycling facility, The 9 average background exposure rate was reponed as 11 pR/h with readings betwoen 10 to 13 pR/h for the concrete surveyed. The steel rebar embedded within the monolith was surveyed and disposed of as clean waste. Contaminated Drain Line Removal The drain lines embedded in the reactor pedestal were removed using the demolition hammer. The drain lines were disposed of as LSA waste. Removal of the lines resulted e in a hole that penetrated the building foundatior. Shoring was insened into the hole created in the process of removing the drain lines to prevent the foundation fill material from eroding below the reactor room floor and sliding down into the hole, 9 LSA Shir>ments Five LSA shipments were made during the Boelter Reactor D&D project. Waste shipments #1 and #2 consisted of the removable concrete shield blocks. The live shield O blocks were cach contained within a strong, tight wooden box and surveyed to ensure that no loose contamination existed on the outside of package The overhead crane and forklift were used ta remove the blocks from the reactor, room and plaec them on the 9 24 9

) transpon vehicle. ) Waste shipments #3

  1. 5 consisted of primarily rubbled concrete packaged in LSA containers. Once tilled cach LSA container was weighed, surveyed and shipped to the Beatty, NV burial site. Four containers constituted a typical LSA waste shipment. The radionuclide content data necessary for preparing the shipping manifests was calculated

) using isotopic analysis perfonned by a commercial laboratory. The steel rails were placed into one LSA container on top of approximately 18" of concrete. A wooden frame was constructed to position the rail at the center of the j ) container and to ensure minimal movement during transportation thus minimizing personnel exposure. The remaining portion of the container was filled with activated concrete. The LSA container containing the rails was part of waste shipment #3. The beam ports were inserted into LSA containers during routine filling operations. A ) portable band saw was used to cut beam ports into appropriate lengths to fit in the containers. 3 The dates of the LSA waste shipments were as follows:

  1. 1 08/18 S 2 l
  2. 2 08/25 S 2 h
  3. 3 Oo/02S2
  4. 4 09/14S2
  5. 5 09/23S 2 P

Prior to the conclusion of NES/IES decommissioning activities, all wastes were shipped off-site and disposed of appropriately. All records of contaminated and clean waste shipments were turned over to the University for storage and archive. 4 Final Site Cleanun Upon conclusion of the decontamination and final survey activities, the site was cleaned to a condition that met or exceeded the as.feund housekeeping of the site. i J t 25 L

i d FINAL SURVEY l'ROCEDURES The Boeller Reactor facility was surveyed to demonstrate compliance with the established release criteria. Several hot spots were identified during the reactor room floor survey and were decontaminated. A bushing hammer was used to scabble the concrete and the debris was vacuumed and disposed of as LSA waste. All other areas required little or no funher decontamination. 1 Sampling Parameters The general appmach consisted of dividing the surfaces (i.e., floor, walls, ceiling) into a i predetennined grid pattem. Typically, the floors and lower walls (up to 2 meteis) were divided ] into i meter by I meter grids, while the upper walls were 2 meter by 2 meter grids and the ceilings 3 meter by 3 meter grids. The water in the process pit sump was assayed with the liquid scintillation counter. The water j was disposed of in the sanitary sewer system if the radionuclide concentration was below NRC effluent release guidelines for each radionuclide present. Only tritium was identified in the sample and at levels which permitted release into the sanitary sewer. i Soil samples collected beneath the pedestal were sent to a commercial' laboratory for gamma j spectrometry and analysis for H-3 and C-14. The radionuclide concentrations in the soil indicated that the soil had not been contaminated by reactor operations. Background / Baseline Levels idennfled -Background dose rate measurements were taken with the Bicron Micro-Rem meter in unaffected locations within the Boelter reactor facility and in adjacent University facilities with ~similar t constmetion materials to that of Boelter Hall. The background dose rate was calculated using the equation in Section 3.4 of " Final Survey Procedure" (Ref.1) and was equal to 13 prem/h. Background detennination for fixed and removable contamination surveys is discussed in the section on techniques for reducing / evaluating data later in this report. 26 1

O Guidelines Established The Boelter reactor deconunissionmg was perfonned within the guidelines established in the O consent order between UCLA, the NRC, and the Committee to Bridge the Gap. The radioactisc release criteria established for the Boelter reactor facility dismantlement include compliance with the surface contamination levels presented in the NRC Regulatory Guide 1.86, "Tennination of 9 Operating Licenses for Nuclear Reactors"(Ref. 4). For beta gamma emitters, the average fixed 2 plus removable contamination levels could not be greater than 5000 dpm q/100 cm and 2 removable contamination levels could not be greater than 1000 dpm pq/100 cm. Furthennore, l 2 the maximum surface contamination level, applied to an area of not more than 100 cm, could not g 2 exceed 15,000 dpm pq/100 cm, In addition to the NRC Regulatory Guide 1.86 requirements, dose rates were not to exceed 5 prem/h above background radiation levels, measured I meter from the surface of interest. 9 Equipment and Procedures Selected Survey measurements for fixed contamination, removable contamination and dose rates (floor grids only) were obtained for each survey grid using the instruments and techniques. described below, instruments and Equipment Two Ludlum 2221 meters with associated 44-9 GM pancake probes were used during the final release survey, A Tc-99 electroplated beta source (2060 i 80 dpm) was used to detemiine the efliciency of all GM pancake probes and the Ludlum 2929 smear counter O used during the project. A jig was used to ensure a reproducible geometry during the initial efficiency detennination and for the subsequent daily instrument perfonnance checks of the 44-9 probes. The efficiency for the Ludlum 2221 (serial # 73700) and 44-9 (serial # 066761) was 0.2110.0088 counts per disintegration, while the efficiency for the O Ludlum 2221 (serial # 73683) and 44-9 (serial # 057871) was 0.18 0.0077 counts per disintegration. The elliciency for the Ludlum 2929 smear counter using the Tc-99 source was 0.1510.0076 counts per disintegration. 27 O

Appendix B illustrates an example of the efficiency uncertainty calculatio'n using a propagation of errors. The Tc-99 source was chosen because it emits radiation of the level and type that was expected at the Boelter Reactor facility. Tc-99 provided a conservative estimate of the GM pancake probe efficiency for the two prevalent radionuclides at the Baciter reactor i site, Co-60 and Eu 152 (based on the isotopic analysis of the activated monolith concrete). The conservatism resulted from the fact that the beta endpoint energy of Tc 99 (292 kev) is less than that of Co-60 (318 kev) and Eu 152 (696 kev). The higher beta energies of Co-60 and Eu 152 resulted in higher actual efficiencies than that detemiined with Tc-y 99. LSC counting vials were analyzed on a Packard Model 2500 TR liquid scintillation ) counter with a dual-label protocol for 11-3 and C 14 (named "311 14C-OPEN WINDOW"). Three regions wen: established for the smear analysis, Region A (0 - 12 kev) for 113, Region B (12 156 kev) for C-14, and Region C (150 2000 kev) for higher energy beta emitters. The LSC data output contains the respective net counts in each region. f instrument Use Techniaues Initially, a 100% surface scan survey was perfonned for each grid with a Ludlum 2221 ratemeter/ scaler and 44 9 GM pancake probe. Then five (5),30 second direct beta-gamma contamination readings were taken within each grid using a Ludlum 2221 meter and 44-9 GM probe. These measurements were unifonnly spaced (i.e., similar to the ) pattem of a live on a die), with the scan survey serving to identify the highest direct contamination reading location for each of the five documented grid locations. Two smears (1.75 in. diameter cloth sampling smears) were taken within each grid, one ) -of which was taken at the location of the highest direct contamination reading. The smears were counted on the Ludlum 2929 with phoswich detector. detector for 30 seconds. In addition, one moistened paper smear was obtained per every four grids for detection oflow energy beta emitters with a liqui i scintillation counter (LSC). The LSC 28 )

l O analysis functioned primarily as a screening tool to ensure that the surface contamination levels of 113 and C-14 were below the release criteria. 4 The liquid scintillation smears were placed in 20 mi glass vials and prepared for counting. One milliliter of a 50%-50% water and alcohol solution was dispensed on the smear paper to facilitate clution of the collected activity. Apprmimately 10 ml of Ultima Gold XR 9 scintillation cocktail was added to each vial. The vials were shaken vigorously and stored for about an hour to allow for any photoluminescence to decay. The LSC counting protocol yielded accurate activity results as long as only 113 and/or g C-14 were present in the sample. However, when higher encre beta emitters were present (e.g., Co 60. Eu 152) the activities reponed were overestimated due to the spilldown of counts into the lower energy regions. For this case, conservative activities were reported for 112 and C-14. O Dose rate measurements were taken for each lloor grid (and the lower walls of the Reactor Room) with a Bicron Micro-Rem meter. The dose rate measurements were taken at approximately I cm and i m from the surface. Procedures Followed O The survey techniques used for the unrestricted release of the Boelter Reactor facility are covered in NES Procedure 82A8021," Final Survey Procedure" (Ref.1). This procedure satisfies the requirements of NUREG/CR-2082 (Ref. 2). O Surveying Organi:arion The surveying organization consisted of Mr. Abelquist in the role of lead surveyor assisted by Mr O Needrith, one senior radiation technician, Mr. Patrick llorkman, and two junior radiation technicians, Mr Mark Wachowski and Mr. James Castle. All final surveying was overseen by the UCLA Radiation Safety Office and Mr. Scott State of Dames & Moore, the radiological engineer retained by UCLA for the decommissioning project. O 29 Oj

) i SURVEY FINDINGS ) Techniquesfor Reducing l Evaluating Data The minimum detectable activity (htDA) was calculated for both the fixed contamination survey instmmentation (i.e., Ludlum 2221 and 44-9 Gh1 pancake probes) and the smear counter (i.e., Ludlum 2929). The h1DA was calculated by the following equation (Ref. 3): 2.71 + 3.29 + MDA = (cficiency) (E# ") [ 8 100 cm

where,

) R. = Background counting rate (cpm), T = Background count time (min), and ) T, = Sample count time (min). The h1DA for the Ludlum 2221 and 44-9 Ght pancake probe was calculated in the same units as 2 the fixed contamination results (dpm/100 cm ). As an example, the h1DA for the Ludlum 2221 ) (serial # 73683) and 44-9 (serial # 057871) can be calculated for a background counting rate of 50 cpm; 2.71 50 cpm, 50 cpm , 339 yg, 0.5 min T 1 min 0.5 min (2) 2 (0.18 c/ dis)( 15 cm 100 cm2 ) or, U" MDA = 1690 (3) 2 100 cm ) 30 )

O Thut the MDA is approximately one-third of the average surface contamination level (5000 dpm 2 p yl100 cm ) in NRC Regulatory Guide 1.86 (Ref. 4 9 The h1DA was calculated in a similar manner for the Ludlum 2929 smear counter, llowever, no 2 correction for probe area was necessary to convett to units of dpm/100 cm since the smeared 2 surface area was approximately 100 cm. Also, the background count!ng rate for the Ludlum 2929 9 was detennined by a thirty (30) minute count (typical background was about 80 cpm). The h1DA 2 for the smear counter was approximately 320 dpm/100 cm, about one third of the removable surface contamination level (l(X)0 dpm -y/100 cm') in NRC Regulatory Guide 1.86. 9 The h1DA for the liquid scintillation counter was extremely low due to the low background counting rates (e.g.,12 cpm for H-3) and high counting efficiencies. Each LSC smear, as well as the background smear, was counted for ten minutes. The efficiencies for H-3 and C 14 of 67% and 96%, respectively, resulted in extremely low tainimum detectable activities (about 8 dpm/100 9 2 cm for H 3). The survey data were recorded onto the r.ppropriate survey fenn. The direct contamination O readings were converted to dpm/100 cm, using the following fom1ula (probe area for the Ludlum 2 44-9 Gh1 pancake probes was 15'cm ): d)m/100 cm2= I" " 9"' ~

  1. f*-

(eficiency) (E' "") 100 cm2 I The average background count rate for each Ludlum 2221 and 44-9 Gh1 pancake probe was detemiined by a series of three 1 minute counts. Dch direct measurement of fixed contamination was thirty seconds in duration. If the dircer measurement (correctedfor a i minute count time) resulted in a value less than or equal to the &tekground count rate, the net count rate was given a value ofI cpm. For example, if the Ludhun 2121 and 44-9 probe had a background count rate of 50 cpm and a 30 second direct measurement resulted in 23 counts (gross count rate equals 46 cpm), the net count rate would be given a value ofI cpm (as opposed to the actual net count rate of-4 cpm) and then converted to dpm!!OU cm' by dividing by the efficiency and correctingfor the 31 6

probe area. Although this practice bias:s the average stuface contamination restdts, the effect is the documentation of conser 'ative survey results. The average of the five direct contamination ) measurements was calculated for each grid and compared to the average surface contamination 2 level (5000 c; Bq/100 cm ) n NRC Regulatory Guide 1.86. Statistical Evaluanon Limited statistical evaluation was perfonned to analyze the survey findings. The standard errors in each of the average surface contamination levels were calculated by propagating the ermr in ) the variables of Equation (4). Specifically, the errors in the gross count rate and background count rate were obtained from the application of Poisson statistics and the method of determining the error in the counting efficiency is illustrated Appendix B. No error in the active area of the pmbe was assumed. All survey values were less than the limits established for release for ) unrestricted use when comparing the limit against the measured value plus two standard deviations. More detailed statistical analyses were deemed to be unnecessary for this project. Comparison of Findings with Guideline Values and Conditions B The fixed contamination results for all survey grids were less than the average surface 2 contamination level (5000 dpm pq/100 cm )in NRC Regulatory Guide 1.86. The highest average surface contamination levels were found in the process pit sump, with readings of 2260 t 244, 24201249, 20501237, 22301244, and,2600 253 dpm p-//100 cm, respectively, for the 2 north, east, south, west and floor surfaces. Overall, there were very few fixed contamination rea' dings that exceeded the MDA for the instrumentation. A summary of the fixed contamination y readings for the floors are listed in Appendix C. The maximum and average readings for each grid are listed. The removable contamination survey consisting of tbc two cloth smears per grid resulted in all E measurements less than MDA No alpha contamination was identified on any of the smears. As stated earlier, one of the two smears per grid was taken at the location of the highest direct contamination reading. Ilowever, contradictions to this procedure appear for the reactor room floor survey. This is because the smears were taken prior to the " hot spot" decontamination, and 32 S l

l O i the highest direct reading prior to decontamination was not always the highest direct reading following the decontamination effort. O The LSC smear results were all below the removable surface contamination levels (ifXX) dpm - 2 y/100 cm ) in NRC Regulatory Guide 1.86. The highest tritium surface contamination level 2 identified was 516 56 dpm/100 cm The standard error in the tritium activity was calculated 9 by assuming a conservative error in the LSC cfficiency determination of 10?c and for calculational 2 purposes assuming no ermr in a smeared area of 100 cm, The dose rate survey resulted in all measurements being less than 5 prem/h above background (13 g prem/h). The highest dose rate measurement was 16 prem/h. A discussion of results from each room / location surveyed for release is given below. A summary of dose rate for each floor grid is listed in Appendix C. O ~ Control Room (Room 2ml) The fitnl release survey of the Control Room consisted of 97 grids. The Control Room was considered to have a low potential for radioactive contamination and therefore, larger grid sizes were chosen for the floor and walls (2m x 2m) and the ceiling (3m x 3m). The fixed contamination results were all less than MDA, with the exception of some grid locations in the bathroom. Gamma spectrometry analysis attributed the readings on the lloor and wall tiles to naturally occurring radioactivity. The removable contamination survey resulted in all measurements less than MDA with the Ludlum 2929 and LSC. The twenty dose rate measurements at 1 meter ranged from iI to 15 prem/h and averaged 9 12.6 prem/h. Background dose rate was 13 prem/tt Count Room (Room 1005) O The final release survey of the Count Room consisted of 138 grids. The Count Room was considered to have a moderate potential for radioactive contamination due to its proximity to the Reactor Room. The grid plan was as follows: ikx)r and lower walls 9-33 9 l

) (below 2m) were im x im grids: upper walls and ceiling were 2m x 2m grids. The fixed contamination survey resulted in all measurements less than blDA for the floor grids. While several grid locations on the walls were slightly wove MDA, the averaged fixed contamination was below h1DA. All smears were less than htDA with the Ludlum 2929 and LSC. The 48 dose rate measurements at I meter ranged from 12 to 16 prem/h - ) and averaged 13.5 prem/h. Background dose rated was 13 prem/h. The laboratory bench in the center of the Count Room was surveyed as a separate item. ) Both the fixed and removable contamination measurements were less than MDA.- l Storace Room (Room 1003) ) The final release survey of the Storage Room consisted of 74 grids. The Storage Room was considered to have a moderate potential for radioactive contamination due to its proximity to the Reactor Room.' The grid plan for the Storage Room was the same as for the Count Room ( Im x im grids on the floor and lower walls: 2m x 2m grids on the ceiling and upper walls). The fixed contamination survey resulted in all readings less than h1DA for the ceiling and walls, and only 1 measurement above htDA on the floor (1900 i 518 dpm/100 cm?). All smears were less than h1DA with the Ludlum 2929 phoswich detector, while the highest 3 2 LSC smear, on the west wall, indicated 4918% dpm 11/100 cm. The twenty dose rate measurements at I meter above the floor ranged from 13 to 16prem/h and averaged 14.2 prem/h. Background dose rate was 13prem/h, Reactor Room Floor (Room 1000) The final release survey of the Reactor Room Floor consisted of 182 grids (im x im grids). The Reactor Room Floor was ci,asidered to have a high potential for radioactive contamination since the monolith demolition activities were conducted in the Reactor Room. The Reactor Room Floor was divided into 4 quadrants; nonhwest, northeast. 34

() i southwest, and southeast. Thit1y-two (32) elevated readings were identined on the reactor room 0oer during the final release survey. The elevated readings were defined as those " hot spots" exhibiting 2 averaged contamination levels exceeding approximately 3000 dpm pe//100 cm. An Eberline ESP-1 and AC-3 alpha scintillation probe were used to check the " hot spots" for 9-the presence of alpha contamination; no alpha contamination was detected. The " hot spots" were decontaminated and resurveyed. Post-decontamination results were documented for the appropriate grid location. The highest pre-decont,mination result was 2 43,600 2,500 dpm -y/100 cm, while the highest post-decontamination result was 2220 g 2 i 538 dpm -1100 cm, 2 The highest average fixed surface contamination level was 1680 i 226 dpm -1100 cm in grid A1 of the southeast quadrant. All smears were less than h1DA with the Ludlum 9 2929 phoswich detector while the highest LSC smear, on the northwest Door, indicated 516 i 56 dpm 'H/100 cm'. The 182 dose rate measurements at I meter above the floor ranged from 10 to 16 prem/h and averaged 13.6 prem/h. The background dose rate was 13 prem/h. Reactor Room Catwalk 9 The Onal re; ease survey for the Reactor Room Catwalk consisted of 45 grids. The Reactor Room Catwalk was considered to have a moderate potential for radioacdve contamination. The grid plan consisted of 25 2m x 2m grids on the top of the catwalk and 20 2m x 2m grids on the bottom of the catwalk (the me:al grating on the west side 9 was only surveyed from the top). The fixed contamination survey resulted in all readings less than MDA for the bottom of O the catwalk and only one measurement exceeded MDA on the top of the catwalk. All smears were less than MDA with the Ludlum 2929 phoswich detector while the highest LSC smear, taken on top of the catwalk on the south side, was 76 t 17 dpm 'H/100 cm. 2 9 35 9

i Reactor Room Menanine l The final release survey of the Reactor Room Menanine consisted of 76 grids. The Reactor Room Mezzanine was considered to have a moderate potential for radioactive j contamination. The grid plan for the Reactor Room Mezzanine consisted of 24 3m x 3m grids on the ceiling and 52 2m x 2m grids on the walls between the catwalk and the i ) ~ ceiling. A manlift was used to survey the Reactor Room upper walls and ceiling. 1 The fixed contamination survey resulted in all readings less than MDA for the ceiling and walls. All smears were less than MDA with the Ludlum 2929 and LSC. ) Reactor Room Ramp ) Die final release survey of the Reactor Room Ramp consisted of 57 grids. The Reactor Room Ramp was considered to have a moderate potential for radioactive contamination. The grid plan for the Reactor Room Ramp consisted of im x Im grids on the floor and lower walls,2m x 2m grids on the upper walls, and 3m x 3m on the ceiling. } The fixed contamination survey resulted in all readings less than MDA, except for 2 2 measurements at 1650 t 502 dpm/100 cm, All smears were less than MDA with the Ludlum 2929 phoswich detector, while the highest LSC smear, on the ramp floor, ) indicated 21 4.0 dpm "C/100 cm. The fourteen dose rate measurements at I meter 2 above the floor ranged from 12 to 16 prem/h and averaged 13.8 prem/h. The background dose rate was 13 prem/h. ) Reactor Room Walls The final release survey of the Reactor Room Walls consisted of 123 grids. The Reactor ) Room Walls were considered to have a moderate potential for radioactive contamination. The grid plan for the Reactor Room Walls consisted af im x im grids on the lower walls (up to 2m) and 2m x 2m grids on the upper walls. )= 36 )-

O. l l The fixed contamination suivey resulted in all readings less than MDA In addition, all smears were less than MDA with the Ludlum 2929 phoswich detector. The highest I.SC smear (75 dpm/100 cm on the east wall) was most likely from Co-60 and/or Eu 152, 8 2 since a significant fraction of counts were detected in the higher energy channels. No uncenainty in the LSC activity was calculated when a significant fraction of the counts t were present in the higher energy channels. The 98 dose rate measurements at I meter S fmm the lower walls ranged from 10 to 16 prem/h and averaged 13.1 prem/h. The background dose rate was 13 prem/It The ove head crane and ventilation duct on the cast wall were surveyed as separate items. g. The fixed and removable contamination measurements for both items were less than MDA. Process Pit S The f' al release survey for the Process Pit consisted of 81 grids. The Process Pit was m considered to have a high potential for radioactive contamination since it contained the reactor plumbing and sump. The grid plan for the Process Pit consisted of im x Im grids for all surfaces. Many of the lixed contamination grid locations were less ean MDA, however, there were locations on the south wall, process pit floor, and sump that exceeded the MDA. The highest averaged fixed suiface contamination levels were found in the precess pit sump, with readings of 2260 244,242G 249.2050 237,2230 244, and 2600 253 dpm 2 pq/100 cm, respectively, for the north, cast, south, w est and floor surfaces. All smears 8 were less than MDA with the Ludlum 2929 phoswich detector, while the LSC smears indicated minimal (all smears less than 50 dpm pq/100 cm ) removable surface 2 contamination within the process pit. The 18 dose rate measurements at i meter above O the floor ranged from 13 to 16 prem/h and averaged 14.2 prem/lt The background dose rate was 13 prem/lt O 37 9 _a

)- Fuel Storace Pits I - The final release survey package for the thirty Fuel Storage Pits was documented differently than for the other Boelter facility survey packages. Each Fuel Storage Pit was divided into upper and lower grids. Five 30 second direct contamination measurements were taken for each grid. The live survey locations per grid were detennined by scanning each pit with the Ludlum 2221 and 44 9 Gh1 pancake probe to identify the two highest measurements at the top (grid locations I and 2), the highest in the center of the grid (grid location 3), and the two highest measurements near the bottom of the grid (grid locations 4 and 5). The grid-locations were further identitled by north, cast, south or west positions. The thirty Fuel Storage Pit plugs were cach considered as one grid. Two cloth smears were taken per grid and one liquid scintillation smear was taken per pit. An dose rate measurement was obtained from each pit by lowering the Bicron hilcro-Rem meter into the pit an ami's length and recording the highest reading. The Bicron hiicm-Rem meter was also u~ sed to_take one dose rate measurement from each Fuel Storage Pit plug. The highest fixed contamination measurement was 1300 i 445 dpm -t/100 cm2 (Fuel Storage Pit C4), slightly greater than the h1DA (1270 dpm -y/100 cm ) for the Ludlum 2 2221 and 44 9 probe. All of the cloth smears for both the Fuel Storage Pits and plugs were less than htDA. The LSC smears were also less than h1DA. Floor Drains and Pioint inlet / Outlets A final release survey containing the results for the floor drains in the reactor room and the plumbing that exits the process pit was perfonned and documented as a separate task. A direct contamination measurement (1 minute count time) was obtained at the access I-point of each floor drain drain line or valve. All direct contamination measurements were 2 less than the N1DA (1210 dpm 7/100 cm ) for the Ludlum 2221 and 44 9 probe. A-cloth smear was taken at each access point all results were less than h1DA. hiinimal removable surface contamination was indicated by the LSC smears taken at each of the_. 38 D. ) l

O access locations. The interior surfaces of the four Door drains were surveyed by passing a hiasslin cloth through the drain lines. A direct scan of the Stasslin cloth with the Ludlum 2221 and 44-9 Gh1 pancake probe indicated no presence of contamination. Soil Sample Three soil samples were collected from beneath the concrete pedestal in the reactor room. The samples were located south of the process pit, Hole i being closest to the pit, followed by llole 2 and Hole 3. A concrete core of the pedestal was removed at each location and a sample of the undisturbed soil beneath the foundation was collected. The g soil samples were sent to a commercial laboratory for gamma spectrometry and analysis for H 3 and C-14. At the laboratory, the water in each sample was distilled and analy7ed for tritium. The 4 results for tritium were as follows: Hole 3 4.35 E3 pCi/l llole 2 3.63 E3 pCi/l Hole 1 1.60 E3 pCi/l For comparison, the EPA drinking water standard for tritium is 2 E4 pCi/1. Carbon-14 was not detected in any of the soil samples and only naturally occurring radioactivity (the other photon emitters reported are not likely to be present due to their short half-lives ofless than 32 days) was identified by gamma spectrometry of the soil. 3 Thus, the radionuclide concentrations in the soil indicate that the soil has not been contaminated. Verification of Surveys by UCLA Radiation Saferv Office The UCLA Radiation Safety Of fice perfonned numerous verification surveys for comparison with the contractor results. Direct exposure rate measurements with a Ludlum hiodel 19 survey meter O 39 9

i ) were in good agreement with NES results. Direct scans with a UCLA owned Ludlum 2200-ratemeter and a 44-9 probe were within approximately 107e of the NES results for fixed )~ contamination at hot spots in each room. Swipes were taken for removable contamination comparisons of the low energy beta emitters and analyzed with the University LSC. All 4 verification swipes were less than MDA which was in agreement with the NES results. ) As added verification, a Reuter-Stokes environmental monitor pressurized ion chamber RS-Ill was set up for long duration exposure rate measurements at various locations in the reactor room and the process pit. Results of the integrated measurements were all well within the release ) criteria. An unshielded high purity gennanium spectmsedpy system was also set up in the reactor room for qualitative detemiination of the most prevalent isotopes in the area. Approximately 50 photopeaks were observed dominated by naturally occurring isotopes of the uranium and thorium series. Based on gmss counts, no significant contribution from activation products was present ) in the room. ) ) l 0 ) ) 1 ) 40 )'

SUMMARY

The decontamination and dismantlement of the Boelter research reactor was successfully executed in a professional and timely manner. The resulting condition of the facilities is such that they are acceptable for release for unrestricted use. All remaining contamination levels are within the limits established prior to initiation of the decommissioning per the consent order between UCLA NRC, and the Committee to i Bridge the Gap. ) ) ) ) ) ) 41 )

REFERENCES i j-1. NES Procedure 82A8021. " Final Survey Procedure;" 1992. 2. NUREG/CR 2082, " Monitoring for Compliance with Decommissioning Tennination Suncy Criteria;" 1981. 3. Strom, Daniel J. and Stansbury, Paul S. " Minimum Detectable Activity When Background is Counted Longer Than The Sample." Health Physics 63(3):360-361; 1992. 4. NRC Reg.datory Guide 1.86. "Tennination of Operating Licenses for Nuclear Reactors;" 1974. 5. NES Operational Survey # 92-245 6. UCLA Decommissioning Phase 1 Report. i 7. NES Final Survey Report j 8. NES Final Decommissioning Repo61 for the Boeller Reactor Facility Dismantlement and Final 1 Decommissioning Project, November 1992, i i i i 4 b 4 L i 42

r APPENDIX A. PROJECT SCHEDULE 3 C

l August l Septrmber l October November ID Name Duration Scheduled Start 7/19 l 7/26 l 8/2 l 8/9 l 8/16 l 8/23 l 8/30 l 9/6 l 9/13 l 9/20 l 9/27 l 10/4 l10/11l10/18l10/25 11/1 l 11/8 l11/15]11/22 1 PROCEDURE DEVELOPMENT 6w 7/20/92 8:OOam 3- ,e 3 J 2 DESIGN SHIPPING CONTAINER 1.5 w 7/23/92 8:OOam ,q j 3 MOBILIZE STAFF / SITE MOBIL 12 7.38d 8!3/92 8:OOam 4 PREP SITE FACILITIES 7.5d 8/3/92 8:OOmm p 7-] 5 INITIAL SITE INSPECTION!SUR 1.5 w 8/3/92 8:OOmm p, g INITIAL SURVEY OF MONOLITH l1.25d 8/4/92 8:OOmm y 6 7 STAFF TRAINING /ORIENTATIO 1.25d 8/10/92 8;OOam Q 8 REMOVE SHIELD BLOCKS FOR Sd 8/11/92 8;OOam 9 SHIP CCNTAMINATED BLOCKS, 3.75d 8/14/9210:OOam i 10 CORE BORE BEAM PORTS 1.5 w 8/15/92 8:OOsm i 11 CUT BEAM PORTS INTO SECTI 6.25d 8/20/921:OOpm g l 12 PKG. BEAM PCRT SECTIONS F 2 5d 8/26/921:OOpm E i 13 SETUP MONOLITH CONTAINME 2.5d 8/10/92 7:OOam g 14 REMOVE ACT!VATED STEEL C 1.25d 8/12/92 7.OOam l l t 15 PKG. ACTIVATED CHANNELS F ! 1.25d 8/13/92 7:OOam l j

i I

j 16 REMOVE ACTIVATED MONOLIT 4.5 w 8/14/92 7:OOam { 17 PKG. ACTIVATED CONCRETE F 4.5w 8/17/92 720am l 18 INTERIM SURVEY INS DE MON 3.75d 9/9/9210:OOam l e 19 RMV CLEAN CONCRETE SO. Sl Iw 9/15/92 8:OOom 20 RMV CLEAN CONCRETE NO. Si Iw 9/15/92 8:OOam 1 l 21 RMV CONCRETE TO DIRT LEVE 1.75d 9/22/92 8:OOsm I 22 RMV CONTAMINATED DIRT 3d 9/23/92 3:OOpm g 23 RMV C05TAMINATED DRAIN L 2d 9/23/92 3:OOpm E i 24 SURVEY COUNTING, STOR AGE 6.25d 8/3/92 8:OOam ~ 25 DECON FUEL STORAGE PITS 1.5 w 9/14/92 8:OOsm i 26 UNAL RELEASE SURVEY STOR 6.25d 9/15/92 8:OOsm l Project: UCLA DECOMMISSIONtNG Cntical I2 Ll Progress E8HNMEluuMM'EE Summary Date: 11/2C/92 Nonentical M.lestone $ Rolled Up h Page 1 O e e 4 e e 9 6 4 e e

l ll 1I \\ v 2 2 / 1 1 r [5 e 1 b / m 1 e l1 vo 8 / N 1 1 y l 1 / 1 1 5 j iil4f 2 /0 g 1 I 8 1 / r 0 e w 1 b l o 1 t 1 j c / O 0 j 1 l 4 p /0 ) g 1 l l 7 2 / 9 y l 0 g 2 / r 9 e l b l 3 g 4 m 1 e / l9 O tp e S 6 y l g / 9 y p r U a 0 md n h u o mu 3 / l 8 [ S R l Q 3 2 / u l 8 l E t 1 2 s / W u 8 e g ga A 9 P / 8 l l 4 2 / 8 ll e 6 s n ~ 2 s o / e ts W l o l r 7 g e 9 r a P M 1 /7 1 m m m m m m m m m m m m m m m a a s s a m p a p a p a s c p t O O O O O O 0 O O O O O O 0 2 ra O O S O O 0 t O O O O 3 8 3 1: S 8: 7 7 7 7 7: 2 O O O O: 7: 8: 8 3 1 d 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 W lu e 9 9 9 9 9 9 9 9 9 9 9 9 9/ 8/ 6 9 9 / / / / / / / / / / / / / / 1 1 7 1 1 1 3 1 4 4 8 1 1 de 2 1 2 3 3 3 / / / 1 2 3 2 / / / / / / / 9 9 9 / / / / 0 0 h 9 8 8 8 8 8 9 9 8 9 1 1 c S l' lac i n d d d w w d w d d d d w w w w la it o S 5 5 5 5 5 5 5 r 2 7 5 5 7 7 7 4 3 1 it 3 1 2 8 3 ic c ita 1 1 n W r 1 3 3 3 3 o r 5 1 u C N D N R E P R W M i S E I G O R D P F E P Y TI O T A V A G M F O R M O V N P E M N K O E O A U Y U T I R S N O C R R L G E H S O F A F E S R I A R C A V D V P O S R i P S P E O Pl M M E R E U S P S S E IM E R S T U N T W T E E P L R A S L A N A N R S T S G TI A I E E L O L E M A S D N E K W P M S L M E I A O L N E U T E I C L E E H I C A C P R A N U M O S G N M A I E Z F E N A TI U T L F D A R S I T P S N E I S T IL E T I I / 2 E T P S PI N N N O R S E S B R A9 T N P E P C L L T L O A L 0

  • A O

U C O O O P C 2

m. L C

T O V C C C G A A S A M U/ O E E E R M E E E K Ni N A N E R

1 a

1 N SI D S P R D D D P F F I W I D P t F c : W j te e 0 7 8 9 0 1 2 3 4 5 6 7 8 9 0 1 o a 1 2 2 2 3 3 3 3 3 3 3 3 3 3 4 4 PD r ( f ilill !llllllt l lll !jll'lf

i APPENDIX B l l CALCULATION OF EFFICIENCY UNCERTAINTY ) k ) I ) N

O Efficiency I)eterininatlun for the L222! (# 73700) and 44 9 (#066761) Source Activity (A): 2060180 dpm Tc-99 8 Procedure: Ten (N = 10), I minute measurements of both background and source (T,=T,., = 1 min) <ere obtained with a reproducible source-detector geometry, e Data: Backcroond Counts (Cd Source Counts (C..d 82 519 84 $20 e 71 486 79 511 81 546 84 470 9' 77 5(M 76 522 i 86 499 75 474 g Mean background count (C ) = 79.5 counts Mean source count (C,..) = 505.1 counts The experimental standard deviation, o, was calculated with an llP42S calculator for botli distributions: ne O o,,, for the background count = 4.74 counts o for the source count = 23.6 counts ne The stand:.rd error in both the mean background count and source count was calculated from the 9 experimental standard deviation: 0E 0 (*

  • lR 4

Thus, C - 79.5 t 1.5 counts and C,,, = 505.117.46 counts. The respective counting rates are calculated as follows:

  • l R, u C, /li = 79.5 counts / min R., = C,.fr, 3 = 505. I counts / min s

el

--- -. - -. =. ) The standard error in either the background or source counting rates, og, is calculated by propagating the error in the number of counts. ) I o,'=fo'= -h.l a

  • c c

Since both the background and source were counted for 1 minute and no error was assumed in the ) measurement of time, the errors in the counting rates are the same as the error in the counts, as illustrated with the following calculation of the background counting raic error: Y' = 1.5 _Ct o,,. ) \\, T.,.$ 1, 1 The net counting rate. R,,is equal to R, - R,, or R, = 425.6 countsAnin. The ermr in the net counting rate is propagated as follows: ) o,, = /(o,,)3 + (o,,)I = /(1.5)' +(7.46)2 - 7.61 counu/ min Thus. R, = 425.6 i 7.61 counts / min. ) The counting efficiency is calculated by dividing the net counting rate by the source activity: epiciency, f, 425.6 counulmin, n_2, (p,,g,gg, A 2060 dis / min ) Propagating the error in both the net counting rate and the source activity results in the following equation: ' o,, %"I,*((-R,)(of* ( A2 ) Substituting values into the equation yields: ' 7.61 ): '(-425.6)(80)12 = 0.0088 counu/ dis ) N12060),, (2060)2 4 Tims, the counting efliciency is equal to 0.21 0.00S8 counts / dis. ) ) r.

APPENDIX C ) )

SUMMARY

OF FINAL FLOOR SURVEYS ) ) ? i ) ) ) ). )'

O Nonheast (NE) Souiheast (SE) i i i i i l l i l l d 9 El 2 3 4 5 6 7 F1 2 3 4 5 6 7 = E E D D C ...C.. B Hj 8 x A l d / l i G1 2 3 4~ 5 '6 l 7 G1 2 3 4 5 6 7 1 i r-e I p rc E d; D[ D x C C B B e bA A-I f I Northwest (NW) Southwest (SW) R Below Goor grade to foundation level [_,j Below foundation grade Figure 1: REACTOR ROOM FLOOR SURVEY GI el Cl el

k kcanot koom floor - - - ~ l NT. h% l AE %W la atum ST #m/l Am' premh 4 t Sm/IMkm' pmnh y @oulmm' pmwh 3y 4m/i% m' prenh at I mtt at I tatt at I n tf at I mit Mas Avg Mst Asg Mat Asp Man Asg Al IMO 377 16 64 447 14 3100 1684 16 140 615 16 2 667 326 1) 1200 449 12 1270 5 44 13 1480 It60 13 3 5 92 W2 14 444 100 14 1010 589 15 1440 1860 15 h 4 889 578 12 507 266 13 1140 7 35 13 1260 710 13 5 963 481 15 1520 7 56 14 161 Su6 14 1700 740 12 6 2mo 1110 15 1340 700 14 951 $$8 14 815 348 14 7 1630 74 14 los't 691 15 571 368 12 1830 814 14 Bl 415 563 13 627 190 13 3E0 1190 15 1040 6 38 15 2 %) 652 13 571 213 13 1520 60s 15 1560 801 14 3 1760 408 15 als 3N 13 1460 546 16 1630 977 15 4 5 92 252 13 951 450 15 1270 774 15 189 5 33 12 5 1410 A T, 11 951 441 16 1140 7M 13 1480 9 32 14 6 21e Alto 13 1010 887 13 6M 2M 15 1480 829 13 7 956 479 16 951 4% 14 1520 lli 43 IDO $92 13 Cl 815 489 15 6M 222 13 1840 166 14 INO 786 14 2 1110 815 14 507 273 12 1520 824 15 1560 904 15 3 1740 82 d 6M 349 14 1010 4 81 12 1110 $70 16 4 963 607 12 1010 557 12 1010 44 14 1260 815 13 5 1460 6 92 15 1080 413 13 IDO 697 13 1110 755 13 6 1920 977 16 1010 449 12 697 317 12 963 5 92 13 7 1520 877 15 1010 570 le 697 48' 15 667 385 13 DI 963 489 13 571 298 12 761 M9 13 1180 F25 15 2 INO 3M 12 888 514 13 14t>0 925 15 2220 1050 14 3 859 459 11 444 3M 43 1140 7M 14 1330 918 14 4 667 40 14 6M 406 IS 824 4 25 14 1NO 623 14 5 667 341 13 697 431 13 1520 729 13 1180 725 14 6 IDO 1020 13 951 SM 14 761 311 11s0 740 14 g 7 17ko 927 12 1010 697 12 444 241 15 667 363 14 El 1260 485 11 958 514 il 1390 575 13 1780 1150 14 2 741 385 16 380 165 15 IMO 8% 14 697 2 15 13 3 741 355 12 888 406 14 761 222 15 1140 7M 13 4 1260 h5 14 leo 628 14 1200 8% 13 9%) 799 14 5 22to 871 12 124 D5 13 sti 241 15 - 1840 6M 15 6 189 503 15 1010 5 05 14 1140 754 14 1330 615 12 1 Sil 229 14 JNO lure 'S 1650 799 - 13 1010 SW 13 g F1 741 296 16 824 323 li 1270 6N 15 1580 932 15 2 5 92 365 13 571 381 12 1390 607 13 188 463 13 3 IMO 44 14 IMO $12 12 IMO 7 22 14 1390 633 14 4 1560 593 12 1200 60s 14 1460 584 13 101D 595 13 5 1080 423 12 951 647 13 1010 785 1390 709 14 6 2300 845 !6 1580 809 15 1140 7 35 10 761 450 14 1 1840 524 IS 3040 13 4 14 507 273 11 691 444 13 01 3&O 222 13 911 6M 16 2 317 152 44 1140 374 12 3 317 121 11 1010 D5 15 4 1140 5 37 14 1520 1120 le 5 64 241 15 1520 1060 15 6 761 444 15 1080 337 15 7 1730 1090 15 IMO 551 13 MDA: 1160 & 1940 Jpm/100cm' . MDA. IMO dp/thm' MDA: 1570 dem/100cm' MDA: 1570 & 1780 dem/imm' Bkg: 13 prem,h Bkg: 13 pmnh lug; 13 pem h Bkg: 13 prem!h

.u-C2 D

l Oi l Oi 1 I s e: y,,,r r- ,s s., ./ s', e \\s N A B C N 1 1 f 2 a e. l [_- 3 l ) 4 x I s,'N /, 1, 6 ,,\\s / x, _7 7 l e i i - p O Below floor grade to foundation level Below foundation level g, 5 Sump Figure 2: Process Pit Floor and Sump Survey -e .g-C3

e;

.,.. ~

) m Process Pit Floor ) Location _ -y dpnV100cm prendi at I meter Maximum Average A1 517 146 14 2 507 133 15 ) 3 254 95 14 4 127 57 14 5 444 368 15 6 507 235 14 7 127 82 16 ) Bl' 317 96 15 2 634 374 13 3 697 298 13 4 571 235 14 5 888 482 15 6 317 273 15 ) 7 697 380 14 Cl 5135 2050 13 2 571 266 14 3 254 152 15 4 507 235 13 2 MDA: 1600 dpnV100cm

Background:

13prem/h I Sump-2 Location p y dpnV100cm prem/h at 1 meter Maximum Average N1 4250 2260 El 4370 2420 S1 3170 2050 -WI 5520 2230 F1 3870 2600 2 MDA: 1600 dpnV100cm

Background:

13prem/h - 4 C4 )--

O-I --i_j j F e: COUNTING ROOM E D 9: .1 C B 8 7 6 5 4 3 2 Al A B C D1 4 TRANSFORMER VAULT STOli AGE ROOM 2 e 3 4 1 5 O O C- 'u j 7 6 5 4 3 2 Al REACTOR RAMP AREA B 9l l !b Figure 3: Counting Room, Storage Room, and Ramp Area Floor Surveys (Not to Scale) e l l C5 9 ~.

. _ _. _ _ _. _ _ _ _ _ _. _. _ _ _ _.. _. _ _.. _..... _.. _.. _ _ _ _ ~ _. _. _ _ _ _. -.., 4 i 1 1 I Counting Roots: Shrage Room llamp Area r-. Locanon S.y dpWl(Kkm' prem/h p.y dpn/lotkm' pretitt h y dpWithm8 prenth 7 at I mtr at I mir si i mtr - j Mas Avg Mar Aig Avr hla -~ -.- - a aw ,I' Al 380 235 13 888 495 15 1350 976 14 l 2 1010 557 15 951 647 13 761 520 13 } 3 951 $$8 15 1270 799 14 1010 660 12 j. 4 951 $96 13 1080 546 15 1390 10S0 14 5 634-418 15 1900 10t0 14 697 494 l$ I 6 444 235 14 N/A N/A N/A 1140 $95 12 7 697 431 13 NM N/A N/A 824 533 13 8 507 260 14 N/A N/A N/A N/A N/A NM Ill 761-368 16 1080 622 13 951 514 14 2 444 247 15 1010 709 15 1010 671 16 3 697 469 14 888 $71 14 1080 774 13 i 4 571 380 14 1390 o87 16 1140 766 14 5 697 507 14 951 545 14 16M) 876 I4 6 824 545 14 NM N/A NM 1010 747 13 1 7 507 279 13 N/A N/A N/A 1650 1100 16 l 8 697 456 13 N/A N/A - N/A N/A N/A N/A l Cl 697 406 13 951 558 15 2 h38 494. 13 951 685 12 l 3 761 444 14 951 .602 15 1 4 697 456 13 14N3 565 15 5-N/A N/A NM 1010 646 14 4 6 N!A N/A N1A N/A NM N/A 7 507 380-la NM N/A N/A 8 697 456 13 N/A NM NM D1 1140 558 13 888 400 14 i 2 761 469-14 761 571 14 3 697 456-14 761 545 14 1 4 824 393 13 1270 660 14 I i 5 697 387 14 697 437 14 l 6 888 545 14 NM N'A NM 7 951 596 13 8 571 317 14 El 634 342 13 l 2 507-342 14 3 697 374 12 4 634 323 14 5 697-374 12 6 824 387 13-7 .1010 544 14 8 824 482 11 i = F1 697 387 '4 i 2 697 393 13 3 388 387 12 ~4 824-507 -12 i 5 _ 888 583 13' 6 697 -431 14 -7 951 482-14 8 l 634 311 12 - M DA: 1600 dyn/lakm'. MDA: 1550 dyn/100em' MDA: 1570 dpn/100cm' l BLg: 13 prem/h ' Ilkg: 13 prem/h BLg: 13 prenth i h r C6 ,w,.<..m e

j.. w.

...,wm.ww.-.,%,m. w. w .w w ,w .., -w w., # wm m.me a m e m m, w,m e ,,m y + m, .,.w m e. % .-mm,w,,. .,-w w w n e y.,w.-

- ~ ~. - ~ ~ ~ _.. - _.._.._._.___ __. ._.....m _.__,,__-m-.. ._m_m.... ___.---___..m ~ l i e r I 7 8 9 10 11 12 13 14 yl ( 6 15 ..-.7..... 5 / ,\\. I6 /j', k 's 3 / 4 17 9 J3 h 3 3 18 \\ / g / \\ /' 19 24-kiii IIkkkik .ii iii iiii liii' l E111111111111111 j g l l j jg l l 1,111111111111111 h* 25 2.] 22 21 20 I 1 7 r-~.- a a 9; Figure 4: Reactor Rdom Catwalk: Balcony-(Not to Scale) 9i I I - l hh C7 9: i v.- ~, -., -, - -, - - - - -, ,--.,--.-..-w-,- .. ~ -,., -.....,. -. - -..,, -. - - _ - -.... - -A

Reactor Room Catwalk: llalcony 2 Location p y dpnV100cm Maximum Average 1 1080 615 2 1460 514 3 1390 912 4 1580 1170 5 951 526 6 761 596 7 1080 470 8 1780 '888 9 1140 760 10 507 222 11 1270 774 12 1010 722. 13 1580 709 14 1200 747 15 1270 723 16 1140 810 17 951 438 18 824 209 19 127 57 20 127 51 21 190 76 22 127 51 23 127 51 24 127 57-25 254 76 2 MDA 1690 dpnV100cm 1 C8

O 1 l l i I I A 11 C1 e O CON 1ROL 3 ROOM O 4 4 5 il 5/ J Bl Al C I CHANGE J U 6 ROOM Al O 1 n d B2 A2 b ~ l l e I _j i T' Figure 5: Control Room and Change Room Floor Surveys (Not to Scale) O C9 0

) Control Room 3 Location p y dpnV100cm prenVh at I meter ) Avera.gej Maximum A1 63 38 14 2 190 64 14 ) 3 190 64 13 4 63 38 12 5 127 57 13 B1 254 76 12 2 127 51 13 )- 3 507 159 12 4 32 32 12 5 32 32 13 C1 444 114 13 2 254 83 11 3 824 209 12 3 4 32 32 13 5 127 51 -13 6 444 133 12 } MDA: 1570 dpm/100cm

Background:

13prenVh 2 ) Change Room 2 ' Location p y dpnV100cm prenVh at I meter 2 . Maximum Average Al 1460 849 12 ) A2-1200 760 14 81 1520 1100 13 B2 1780 .800-11 2 MDA: 1570 dpnV100cm

Background:

13prenVh ~ t C10 'i,-.-,-,,,;....-.-..,....---__.-...-.--.,,..'~..~,-.......-.,--..___...-.----.-}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20127K282&oldid=4266002"