ML20046B866

Final Decommissioning Rept for Boelter Reactor Facility Dismantlement & Final Decommissioning Project.

ML20046B866
Person / Time
Site:	05000142
Issue date:	12/31/1992
From:	CALIFORNIA, UNIV. OF, LOS ANGELES, CA
To:
References
NUDOCS 9308060344
Download: ML20046B866 (59)

Text

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• FINAL DECOAINIISSIONING REPORT FOR THE IlOELTER REACTOR FACILITY DISAIANTLEMENT AND

1. FINAL DECOAIAIISSIONING PROJECT D

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UNIVERSITY OF CALIFORNIA LOS ANGELES, CALIFORNIA DECEMBER 1992 o

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O' FINAL DECOSIMISSIONING REPORT FOR THE BOELTER REACTOR FACILITY DISMANTLEMENT

.O AND FINAL DECOMMISSIONING PROJECT

'O DOCKET NUMBER: 50-142 LICENSE: R-17 i O.

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Report Prepared By ,

Scott State. Dames & Moore Amir Huda, UCLA I O Joseph M. Takahashi, UCLA  :

. UNIVERSITY OF CALIFORNIA  :

LOS ANGELES, CALIFORNIA [

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i Table of Contents

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- INTRODUCTION . . .. . . ... . ... . . .. ... 1 BACKGROUND... . . . . .. . .. .. . .. .... . . ... .... 2 Reason For Decommissioning ... . .. . . .... .... ....... .... 2

,O Managemeat Approach . . . .. . . .. . . .. ....... 2 Schedule ... .. . ... .. . .. ... .. . ...... .. 3 -

SITE DESCRIPTION . ... . .. ... .. . .. . ... . . ....... .... 5 Type and Location of Facility . . .. . ... ... . .. .. ....... 5  ;

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g- Ownership . . . . . . . .. . . . . . .. ... ....... . .. ... .... 5 Facility Description .. ....... . ... . . . ...... . . ........ -5 Buildings . . . ... . . . . . .... .. .. ..... ..... ......... 5  ;

Grounds . . . . .. ... ....... .... ........ 9 OPERATING HISTORY. .. . .. ..... . . ..... .. ... . . .. . ..... 10 O Licensing and Operations . . ... . .. . .............. ........ .... . 10 .;

Processes Performed .. . ... . . ... ... .. ... .... . 10 ;

Waste-Disposal Practices . . . . . ........ ... . .... ...... 10 .

I DECOMMISSIONING ACTIVITIES . . ... . .. . ... .... ........ ...... 11 i Objectives .. . . . . . . . . . . .. .... ...... ....... 11 l O Results of Previous Surveys . . . . . . .. . . .. . ... .. . . . ...... 11 Project Health and Safety Program . ... ... ... .. .. ............. 16 Decontamination and Demolition Procedures . . . . . . , .. ... ........ 16  ;

Routine Activities . ... . . . . . . . ... .... 16 Operational Surveillance .. . . .. . . . . . . ... ...... 16 Air Sampling Surveillance . .. . . .. ... .. . ...... . 17 O Radiological Protection . 17.

Personnel Exposure Monitoring . . . .. .. ... . . . ... .. I8 l Waste Screening ... ....... .. .. . ... ... ... ..... .... ..... 18 l Specific Project Tasks . . . . .. .. . ... ..... .. . . ......... 18 Removable Shielding Blocks . . . . .... . . . .. . ......... 18  ;

O Decontamination & Removal of Shower & Sink ..... ..... .. ......... 19 Monolith Containment and HVAC Contamination Control .. ..... ..... .... 19 Steel Support Rail Removal .. . . . . ... .......... ..... 19 ,

Beam Port Removal . . . . . ... . ... ... .. . ... .....-20  ;

Activated Concrete Removal . . 20 Major Contaminants Identified .. . . ..... .. 21 O Activated Concrete Packaging ... .. .... . . .. .. .. ....... 22  :

Process Pit Containment . . . .. .. . . . . . ......... 22 Process Pit ifolding Tanks Removal .... . . .. ... . . ... 22 Process Pit Piping Removal . . . .... .. .... . . ..... 22 Removal of Sludge in Sump . ..... . . .. .. . . . . 23 Process Pit Decontamination . .. .. ..... . .. . . .. .... .. . 23 0 Clean Concrete Removal . ... .. . ....... . .. . .......... 23 Contaminated Drain Line Removal . . ........ .. . . ... ....... 24 O

0- - i LSA Shipments ... . .. . .. . .. ... .. ...... .... 24 Final Site Cleanup . . .. . ... . . . . ...... 25 -

O FINAL SURVEY PROCEDURES .. ... . . ... . . 26 +

Sampling Parameters . ... .. . . ... . . . .. . . . . . . 26 Background / Baseline Levels identified . . ... .. .. . . . . .... . 26 Guidelines Established . . . . . .. . . . . . . ... .... .. . . . . 27 Equipment and Procedures Selected . . .. ... . .. ....... . ..... 27 ,

Instruments and Equipment ... .. ... . .. ... ... ....... . .. 27

-O- Instrument Use Techniques . . ...... .. .. .. . . ......... 28 Procedures Followed ... . . . . . . ... ..... . ... .... 29 Surveying Organization . . .. . ... . . .. .. . . . . . . . . ...... 29 SURVEY FINDINGS . . . . .... . ... . . . .... .. ......... ........ ... 30 '

Techniques for Reducing / Evaluating Data .. . ... .. 30 0 Statistical Evaluation .... .. ...

.. ..... 32 Comparison of Findings with Guideline Values and Conditions ... ... ......... .. 32 Control Room (Room 2001) . . ......... . ..... ....... .. 33 f

Count Room (Room 1005) . .... . . . ..... ...... ...... 33 Storage Room (Room 1003) . . . .. . . .. .. . ...... ...... 34 O. Reactor Room Floor (Room 1000) . . . . . . . . ...... ... ..... ...., 34 Reactor Room Catwalk .... .. ...... .... ............ . 35 Reactor Room Mezzanine . . . . .... . .... . . . ...... .. 36 Reactor Room Ramp . . . . . . ............... 36 Reactor Room Walls . .. .. . .. ... .. ......... 36 Process Pit . . . . . . . .. . ..... ...... ...... 37 O Fuel Storage Pits . ... .. .. . . . .. . .. . ....... 38 l Floor Drains and Piping Inlet / Outlets . .... ..... .............. 38 Soil Sample . . ... .. ... . . . . . . ..... ... . ........ 39 i Verification of Surveys by UCLA Radiation Safety Office . . .... ........ ..... 39 .

SUMMARY

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REFERENCES ..... . . . . ... ..... . .. .. . .. . .. 42 APPENDICES  ;

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INTRODUCTION This report describes the final < isc of deconunissioning activities including release smveys that

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were performed at the UCLA Argonaut Reactor facility for its release for unrestricted use. l The Final Phase activities were undertaken in accordance with the requirements of Section 2.5 of O-We Atomic Safety and Licensing floard Sctilement Agreement dated September 30,1985 and the j NRC Phase II Order dated July 28,1989. j a

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D BACKGROUND Reason For Decommissioning D

UCLA announced its intent to decommission the Boelter Research Reactor on June 14, 1984.

Chancellor Charles Young noted that the decommissioning decision was reached solely as a result g 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 8

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, O

1986. A repon on Phase I decommissioning activities was submitted to the NRC on April 12, 1988. The first phase involved renoval of the following:

. reactor core-moderator graphite thennal column D . shield tank

. peripheral equipment

. fuel boxes

. control blade system components protruding paas of pipe and structures most of the concrete shield blocks D

The Phase Il Plan was submitted to the NRC on June 10,1988. The NRC requested additional infonnation in letters dated October 12,1988 and March 14,1989. Supplementary infomiation was provided by UCLA in letters dated December 7,1988 and March 31,1989. The NRC Order 3 for Phase II 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 a decontamination of all remaining facilities 3 . a tennination survey The final smvey was perfonned by Nuclear Energy Services (NES) with oversight provided by l the UCLA kadiation Safety Office. Figure i presents the overall organizational structure for the ;

l project and final survey. UCLA retained Dames & Moore to provide radiological engineering l

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Chancellor f i

O Executive Vice Chancellor Administrative Vice Chancellor Capital Programs I Vice Chancellor i i

Dean, School of Engineering Assistant Vice Chancellor Director of O. & ' pplied Sciences Business Operations Administrative Officer Radiation Safety Officer Senior Project Manager i Owner's Representative Capital Programs i O ____________________ij l

QA/QC Manager +------- Radiological Engineer, +--- NES, Site Supervisor Consultant O

Operations Health Physics Supervisor Supervisor  !

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Figure !: Boelter Reactor Final Decommissioning f Project Organization Chart

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IIACKGROUND Reason For Decommissioning ,

i UCLA announced its intent to decommission the Boclter Research Reactor on June 14, 1984.

Chancellor Charles Young noted that the decommissioning decision was reached solely as a result '

j of the changed circumstances affecting the academic benelits and escalating costs of continued operation of the reactor facility in a press release that day.  ;

i Afanagement Approach ,

9 The decommissioning of the UCLA Reactor facility was carried out in two phases. The f

Decommissioning Plan for Phase I was submitted to the Nuclear. Regulatory Commission (NRC)

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on October 29,1985. Phase I commenced after appmval 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, l 1988. The first phase involved removal of the following:

. reactor core-moderator graphite thennal column -

j shield tank

' peripheral equipment fuel boxes control blade system components i protruding parts of pipe and structures most of the concrete shield blocks b

t l The Phase 11 Plan was submitted to the NRC on June 10, 1988. The NRC requested additional i infonnation in letters dated October 12, 1988 and March 14,1989. Supplementary information  ;

j was provided by UCLA in letters dated December 7,1988 and March 31,1989. The NRC Order h for Phase !! was issued on July 28,1989. The second phase included the following:

demolition of the reactor monolith removal of the remaining shield blocks i removal of all process equipment decontamination of all remaining facilities 9 a tennination survey

( The final survey was performed by Nuclear Energy Services (NES) with oversight provided by the UCLA Radiation Safety Office. Figure i presents the overall organizational structure for the  ;

project and final survey. UCLA retained Dames & Moore to provide radiological engineering 2

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oversight and to ensure compliance with on site health and safety procedures. QA/QC acdvities I were perfonned independently within the UCLA Radiation Safety Olnce. Independent QA audits were also performed by both Dames & Moore and NES. 9 The management team possessed sufficient training and expertise to successfully complete the project and to ensure worker and general public health and safety. The UCLA Radiation Safety l Officer, Mr. Joseph Takahashi, served as the Owner's Representative. Mr. Takahashi is a certified health physicist and is responsible for campus wide radiation safety. The QA/QC manager was Mr. Amir fluda. Mr. Iluda possesses a Masters Degree in Nuclear Engineering and has worked in the UCLA Radiation Safety Office for over 4 years. The radiological engineer, Mr. Scott State, O

was provided to the project by Dames & Moore. Mr. State has a Master Degree in Nuclear Engineering, is a licensed pmfessional engineer, and also previously held a reactor operator's license on the Argonaut type reactor.

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The NES management team consisted of Mr. Lee Penney as the Site Supervisor, Mr. William Needrith as the Operations Supervisor, and Mr. Eric Abelquist as the 11calth Physics Supervisor.

Mr. Penney and Mr. Needrith have acquired extensive D&D experience over the past several years including the D&D of the Berkeley Research Reactor. Mr. Abelquist possess a Masters Degree O!

in licalth Physics and has previous D&D and radiological controls experience.

l Schedule O'

NES began mobilization on August 3,1992. IIcalth 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 g pmcess pit. The final release survey began on September 19,1992 and was completed on October 15, 1992. Appendix A details the project schedule.

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j Chancellor 3 Executive Vice Chancellor Administrative Vice Chancellor Capital Programs Vice Chancellor Dean, School of Engineering Assistant Vice Chancellor Director of O & Applied Sciences Business Operations Administrative Officer Radiation Safety Officer Senior Project Manager Owner's Representative Capital Programs O ____________________i]

l l QA/QC Manager +------- Radiological Engineer, +--- NES, Site Supervisor Consultant O

Operations Health Physics Supervisor Supervisor

,) Penhall, Subcontractor Supervisor

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Figure 1: Iloelter Reactor Final Decommissioning Project Organir.ation Chart l

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[3 SITE I)ESCRIPTION

' Type ami!nc.nion <f Facility The facility that was decommissioned was a 100 kW Argonaut type research reactor. General reactor research and teaching were perfonned at the facility including activation analysis and J

reactor physics instmetion.

The facility is located on the UCLA campus in Los Angeles, Califomia on the ground floor of 9 Boelter IIall. Figures 2 & 3 provide a general layout 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.

J Facility Description The facilities that were deconunissioned were fully housed within Boelter Hall. In the case of this

, project, there were no other grounds.

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Buildines O 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 bearing and are of 12-inch thick, reinforced concrete, f aced with Roman brick where exposed.

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0, The interior wall separating the reactor high bay from the remainder of the building is also load bearing and is of reinforced concrete,18 inches thick. All other interior walls are nonbearing,8-inch, cement block walls. O 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 Door to the west 01at contained two retention tanks. Directly cast 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. Above the step, the pit inside diameter is 10.25 inches. ,

The two adjacent rooms on the first D(x>r of the building numbered 1003 (storage mom) and 1005 (count room) were also part of the decommissioning project. These moms contain under 1000 square feet of combined floor space. Each room has a smooth e concrete floor. On the second Door of the building is the fonner control room and a locker room which were included in die scope of the decommissioning project. The control room is less than 500 square feet and contains a tiled Door. The adjacent locker room contains a shower and bathroom widi primarily ceramic surfaces. O None of the areas that were decommissioned contained expansion joints or floor penetrations that were dif6 cult to decontaminate. Piping used for drain lines is contained 9-within the floor of the reactor high bay. The openings of the piping were surveyed as described later in this report.

All areas except the reactor high bay were decontaminated in such a manner that wall and g!

Door surfaces were not damaged. In the high bay, it was necessary to demolish the reactor monolith which resulted in leaving a signiDcant ponion of the floor area in a pitted condition. After release of the facility for unrestricted use. the high bay Door will be restored to a condition suitable for using the facility for other research purposes. O 9:

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! Grounds 1 I O No Erounds outside Boelter 11all were within the scope of the decommissioning activities. [

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The area decontaminated is not subject to concems related to geology, hydrology, i' seismology, meteorology, or population with respect to the extent of contamination.

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E Southem California is seismically active but no major events have occurred which would

'O 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. [

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O OPERATING IIISTORY 3 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 research and teaching. Due to the academic mission of the reactor, it was operated g "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 NP C (license R-71) prio ;9 initial criticality with a limitation on 3 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.

O Processes Perfonned The processes perfomied 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-J product of the operations was activation of reactor core and structural material and radioactive contamination in the reactor process system.

O Waste-Disposal Practices Waste disposal practices were limited to those associated with the nonnal operations of a facility of this type. No waste disposal practices impacted the decommissioning status of the facility.

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Q I l DECOMMISSIONING ACTIVITIES

'O . Decommissioning activities were perfomied by NES and their subsidiary IES with oversight from the

! UCLA Radiation Safety Office. All significant activities associated with the project are described below.

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. Objectives

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i The objective of the decommissioning effort was to dismantle and remove the Boelter Research  :

Reactor, decontaminate the reactor room and ancillary facilities, and release the facilities for i unrestricted use.

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- Results of Previous Surveys i.

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! A walk-through survey of the Boelter Reactor facility was perfomied by NES prior to initiation ,

,0 i of the final decommissioning to confinn radiological conditions that UCLA had reported in the ,

i interim between Phase I and Phase 2 activities as shown in Figures 4 through 7.

1-i ig. Each shield block was surveyed for radiation and contamination levels and the results were

documented. No smearable contamination was detected on any of the shielding blocks.- Shielding 4

j block C-9 exhibited the highest radiation level with a contact exposure rate of approximately 40 .

mR/h. The five shield blocks that remained in the facility after Phase I efforts were roped off and 2 i IO the area posted as a " Radioactive Materials Area."

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l-t A general area removable contamination survey was conducted to detemiine the protective  ;

t I clothing requirements. A baseline exposure rate survey was also performed through out the O facilities to detemiine the proper radiation area posting requirements.

A detailed radiation sutvey was perfonned within the monolith contairunent to determine .

radiological conditions present during initial demolition in this area. A removable contamination ,

O suivey was perfonned to identify tiie nature and extent of removable contamination within the monolith. Baseline air samples were also obtained from within monolith containment. Both the -

smears and air samples were at background levels.

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O' A radiation survey was perfonned to identify contact exposure rates on four structural steel rails within the pedestal of the monolith. Using a teletector, the highest contact expose rate measured was 1100 mR/h. This reading was in agreement with results of previous suncys. 9 Project Heahh and Safety Program All NES personnel, including on-site subcontractors, were trained in accordance with the NES O Radiation Worker Training manual and received a general safety briefing. In addition, all site personnel were given an onentation on University specified guidelines by a University representative.

S' Health and Safety Site Specific Training was administered to both NES and Penhall personnel.

A fifty question test was given to aF personnel at the conclusion of the training.

9 Decontamination and Demolition Procedures The decontamination and demolition activities were perfonned by NES through their subsidiary IES. IES in turn hired subcontractors for concrete demolition (Penhall Company) and waste 9-transportation (Environmental Management & Control). NES/IES has successfully performed 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 9

Office with the aid of Dames & Moore whom the University retained to provide a radiological engineer to suppon the day-to-day D&D operational effons. Specific. 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 performed. This is also g shown in Appendix A.

Routine Activities 9-Or cratinnal Surveillance Routine radiation and contamination surveys were perfonned daily within the radiologically controlled area (RCA), as well as the count room, control room, storage room and reactor mom ramp area. These surveys were conducted to verify that 16 9

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'q contamination had not spread from the controlled surface contamination area (CSCA) during decommissioning operations. The surveys aided in the substantiation that

]. administrative and engineering controls implemented during operations were adequate.

Air Samoline Surveillance Baseline air monitoring was perfonned for the reactor room ventilation exhaust duct on the third floor of Boeller Hall. Both pre- and post- air sample results exhibited no activity above background levels.

D-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 pCi/mi for beta / gamma emitters were recounted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to allow for decay of shon-lived naturally-occurring radionuclides. The air samples collected during activated concrete demolition were allless than the maximum permissible concentration (MPC) for each radionuclide present. The dominant radionuclides detected were Eu-152 and Co-60.

C- Eleven air samples that were collected during concrete demolition within the monolith containment were also analyzed with a liquid scintillation counter (LSC) to determine the H-3 and C-14 concentrations in air. The LSC analysis resulted in overestimated quantities of H-3 and C-14 because of the presence of Co-60 and Eu-152 in the air samples. Even C with the gross overestimate, the concentrations of H-3 and C-14 were significantly below their respective MPC's.

Radiological Protection 9

All decommissioning work within the radiologically controlled area (RCA) was performed under

, the issuance of Radiation Work Pennit (RWP). The RWPs provided the radiation workers with h 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).

D The RWP was prepared by the Health Physics Supervisor (HPS) based on expected and surveyed u

h. 17

()-

conditions. Each radiation worker making an entry to the RCA was provided with both a film badge and a self-reading dosimeter (SRD). The entry and exit SRD readings were documented and reviewed by the HPS on a daily basis. In addition, pre- and post whole tudy counts were 9-administered to all personnel.

An NES/IES administrative limit of 800 mrem / person was established for this project. Appmval t rom the Radiological Services Department Manager was required before project personnel could exceed this limit.

Personnel Exposure Monitorine 9:

The University issued dosimetry and administered entry and exit whole body counts for all site personnel. The total n.an-rem for the project was 1.17 rem, as detemlined from commercial film badge results. 3-The whole body counts were negative for all personnel from the decommissioning project.

Waste Screening

• The liquid radwaste generated during decontamination and demolitjon of the stmetures, equipment, and floors was collected, monitored, and then released to the sanitary sewer, if under the limits 9

of 10CFR20. The major bulk of solid weste, such as concrete rubble, metallic embedments, etc.,

was surveyed according to the release cr;teria of Regulatory Guide 1.86 and the 5 prem/h over background at I meter. Based on the results, these wem either packaged and shipped to a licensed burial site or disposed as clean waste. g Specific Project links Removable Shieldine Blocks 9 The five removable shield blocks u ere removed from their positions on the monolith and placed in the southeast comer of the reactor room. They were subsequently shipped off-9 site for disposal as radioactise waste.

18 9

O' Decontamination & Removal of Shower & Sink O A shower stall and sink were removed from the reactor room (refer to Figure 5). - These two items were surveyed 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.

O-Monolith Containment and liVAC Contamination Contml The monolith contamination control envelope was constructed with 2"x 4" framing and g

10 mil plastic sheeting. The plastic was fastened to the monolith with spray adhesive and

-i duct tape. Two high efficiency particulate air (IIEPA) units were connected to the I containment. Adequate ventilation was ensured by measuring for negative pressure within O the containment. After the activated concrete was removed, all materials used for contamination contml were disposed of as clean or contaminated waste as appropriate.

The reactor room heating, ventilation, air conditioning (HVAC) exhaust duct on the east O wall was scaled with two HEPA filters. The four 11VAC ceiling inlets were covered with cloth media to protect the system from contamination.

Steel Support Rail Removal (O

A detailed radiological survey was perfonned inside the monolith prior to the removal of the four steel rails imbedded in the pedestal below the location the core was previously O positioned. An ALARA review and mock-up training were conducted for personnel perfonning the removal of the steel rails. A remotely operated demolition hammer was used to break up the concrete surrounding the embedded rails. The portion of the rails not being worked on was shielded with lead bricks.

O The steel ralls were loosened and eventually freed from the concrete by the demolition hammer. The demolition hammer was used to dismantle the four separate pieces at each comer. An oxy-acetylene cutting torch was then used to cut the rail into manageable O

sections.

O- 19

O.

The rail pieces were double-bagged and taped. Workers handling the steel rails wore 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 this section.

Beam Port Removal g

The six experimental beam pons that penetrated thmugh the monolith were 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 3 appropriately. The coring equipment cooling water was recirculated to the extent hiossible.

Following removal, the beam ports were surveyed to estimate the depth of concrete activation inside the monolith. This activation depth guided subsequent activated concrete 0-removal.

The beam pon cores were disposed of as LSA waste.

O Activated Concrete Removal As detennined from the beam port characterization, a predetennined amount of concrete g from the monolith surface was removed. The amount of concrete removed was based on an assumed spherically symmetrical activation pattem. The concrete was removed using a Mini-Max hydraulic ram. Water was continuously sprayed on the concrete beiag 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 removed, the monolith was O

surveyed with a Ludlum 2221 and 44-40 shielded Geiger-Mueller (GM) pancake probe to better characterize remaining activation locations. With the hot spots identified and 20 0

l O.

marked, further activated concrete removal was perfonned. This iteration of surveying the monolith and decontaminating hot spots was continued until the release criteria were

'O achieved. l i

Maior Contaminants Identified

'O 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 i 1

concentrations. A concrete chip representative of the interior monolith walls was collected .)

i and analyzed. The analysis identified the following radionuclide content: 1

.g Isotope Activity (DCi/c) l 4

l

'll 1,494.00 l O- "C 502.00 l "Co 2.575.00 j

'"Cs 3.82 l

'"Eu. 5,481.00

'"Eu 293.00

""Eu 45.00 i

'O .

A sample of the steel rebar from the monolith was also analyzed and resulted in 19,600  ;

pCi/g of Co-60.

1 i

O A concrete sample was also taken from one of the removable blocks (block C-9) and similar results were obtained: ,

Isotope Activity (pCi/c)

O

'H 1,750.00 "C 348.00 "Co 3.809.00

'"Cs 20.80

"Eu 23.090.00

'"Eu 1.263.00

'"Eu. 62.00 l

I LO 21 I

l0

0-Activated Concrete Packacing The activated concrete rubble was loaded into LSA containers within the containment. e 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.

O Process Pit Containment A contamination control enclosure was constructed that covered the process pit. HEPA 9:

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 g 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 overflow of water from the sump, the water in the sump was 9 simultaneously pumped into 55 gallon drums. The water was sampled and disposed of appropriately.

O Once de-watered, the holding tanks were disconnected from the piping and mounting brackets. The tanks were then cut within the pmcess pit containment and disposed of as clean waste following a detailed survey.

O 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 4 was placed in a drum, transponed from the containment, and surveyed. All piping was classified and disposed as either clean or radioactive waste as appropriate.

O 22 9

4 Removal of Sludee in Sumr>

$ Water contained in die sump was pumped into a 55 gallon dmm and surveyed to detemiine the correct disposal option. The sump pump and associated piping were then removed. The equipment was packaged and disposcJ of in an LSA container.

O The sludge at the sump bottom was mixed with water and vacuumed into a 55 gallon drum. The sludge was then allowed to settle to the txittom of the drum. The wa.cr in the drum was pumped off and surveyed. The sludge was then solidified and disposed of as LSA waste.

Process Pit Decontamination 3 The process pit floor and walls were surveyed after removal of all hardware. Hot 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.

3 The sump tloor and walls required 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.

.;J Cican Concrete Removal 3 The interior of the monolith was cleaned of all mbbled concrete after preliminary surveys indicated the release criteria had been met. The monolith was then gridded and surveyed in detail for free release. The survey consisted of an initial 100% scan of the monolith interior walls wah the Ludlum 2221 and 44-40 shielded GM pancake probe. The hiehest 3 direct reading was reponed within each gnd. It ranged from 196 to 4549 dpm/100cm 2 2

on the south face and 392 to 4941 dpm/100cm on the nonh face. Smears were then taken on all monolith surfaces and ~259 of the smears taken inside of the monolith were counted on a liquid scintillation counter Smears that were counted with the Ludlum 2929 9

23 l

l 1

q

,)

O.

were reponed as less than minimum detectab!c activity (MDA) except for an area on .op of the monolith which read 447 dpm. Smears counted on the LSC showed a reaximum reading of 12.3 dpm for 'H. After confirmation that all activated concrete had Leen G removed, the monolith walls were demolished and the concrete pedestal was removed.

A larger, more pow crful, demolition hammer was used to efficiently dismantle the " clean" monolith. Water was again used for dust mitigation.

O, The concrete rubble 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 detennined in the laydown area prior to removal of any concrete from the reactor room and then the highest Bicron Micro-Rem meter iluctuation 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 environmental monitoring pressurized ion chamber to ensure acceptable exposure rates, and disposed of at a concrete recycling facility. The O.

average background exposure rate was reponed as 11 pR/h with readings between 10 to 13 pR/h for the concrete surveyed. The steel rebar embedded within the monolith was surveyed and disposed of as clean waste.

O Contaminated Drain l_ine 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 g in a hole that penetrated the building foundation. Shoring was inserted into the hole created in the process of removing the drain lines to prevent the foundation fill material f rom croding below the reactor room floor and sliding down into the hole.

O LSA Shipments Five LSA shipments were made during the Boelter Reactor D&D project. Waste shipments #1 ano #2 consisted of the removable concrete shield blocks. The five shield 0 blocks w cre each contained within a strong, light wooden boK and surveyed to ensure that no loose contamination existed on the outside of package. The overhead crane and forklift were used to remove the blocks from the reactor room and place them on the 0:

24 O

i transport vehicle.

)L Waste shipments #3 - #5 consisted of primarily rubbled concrete packaged in LSA containers. Once filled, each LSA container was weighed, surveyed and shipped to the Beatty, NV burial site. Fcur containers constituted a typical LSA waste shipment. The radionuclide content data necessary for preparing the shipping manifests was calculated

) using isotopic analysis perfomied by a commercial laboratory. I 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 .

f container and to ensure minimal movement during transponation thus minimizing personnel exposure. The remaining ponion of the container was filled with activated concrete. The LSA container containing the rails was part of waste shipment #3. The .

l beam ports were insened into LSA containers during routine filling operations. A 1

)

ponable band saw was used to cut beam pons into appropriate lengths to fit in the i

containers.

The dates of the LSA waste shipments were as follows: .

)

1. 1 08/18 S 2

1. 2 08/25 S 2

) #3 09/02S2

1. 4 09/14 S 2

1. 5 09/23 S 2

) 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.

) Final Site Cleanup Upon conclusion of the decontamination and final survey activitics, the site was cleaned to a condition that met or exceeded the as-found housekeeping of the site.

25 l

) i

. . l

o_-

FINAL SURVEY PROCEDURES ,

O The Boeller Reactor facility was surveyed to demonstrate compliance with the established release criteria.

Several hot spots were identified during the reactor mom 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 further decontamination.

O' Sampling Parameters ,

The general approach consisted of dividing die surfaces (i.e., floor, walls, ceiling) into a O ,

predetennined grid pattem. Typically, the floors and lower walls (up to 2 meters) were divided b

into 1 meter by 1 meter grids, while the upper walls were 2 meter by 2 meter grids and the ceilings 3 meter by 3 meter grids.

O The water in the process pit sump was assayeci with the liquid scintillation counter. The water was disposed of in the sanitary sewer system if the radionuclide concentration was below NRC cffluent release guidelines for each radionucli'Je present. Oniy tritium was identified in the sample O and at levels which permitted release into the sanitary sewer.

Soil samples collected beneath the pedestal were sent to a commerrial laboratory for gamma spectrometry and analysis for 11-3 and C-14. The radionuclide concentrations in the soil indicated O ~

that the soil had not been contaminated by reactor opcrations. ,

Background l Baseline Levels idennfied

'O  ;

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 construction materials to diat of Boelter llall. The background dose rate was calculated using the 0 equation in Section 3.4 of " Final Survey Procedure" (Ref.1) and was equal to 13 prem/h.

Background detennination for lhed and removable contamination surveys is discussed in the j section on techniques for reducing /cvaluating data later in this repon. ,

1 l'

!o I 26

!O: '

O' Guidelines Established The Boelter reactor deconunissioning was perfonned within the guidelines established in the O consent order between UCLA, the NRC. and the Committee to Bridge the Gap. The radioactive 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 G.

Operating Licenses for Nuclear Reactors" (Ref. 4). For beta-gamma emitters, the average fixed plus removable contamination levels could not be greater than 5000 dpm q/100 cm2 and ren;ovable contamination levels could not be greater than 1000 dpm -y/100 cm2 . Furthennore, the maximum surface contamination level, applied to an area of not more th:m 100 cm2 , could not g exceed 15,000 dpm pq/100 cm 2.

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 c 'lected Survey measurements for fixed cont:unination, removable cont amination and dose rates (floor grids only) were obtained for each survey grid using the instruments and techniques described below.

Instr.iments and Equipment Two LuJlum 2221 meters with associated 44-9 GM pancake probes were used during the final release survey. A Tc-99 electroplated beta source (2060 80 dpm) was used to detennine the efficiency of all GM pancake probes and the Ludlum 2929 smear counter 4 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 ciliciency for the Ludlum 2221 (serial # 73700) and 44-9 (serial # 066761) was 0.21 0.00SS counts per disintegration, while the ef0ciency for the O Ludlum 2221 (serial # 73683) and 44-9 (serial # 057871) was 0.18 0.0077 counts per disintegration. The ef ficiency for the Ludlum 2929 smear counter using the Tc-99 source was 0.15 0.0076 counts per disintegration.

4 27 6

r

!O l

Appendix B illustrates an example of the efficiency uncertainty calculation using a l propagation of errors.

O l The Tc-99 source was chosen because it emits radiation of the level and type diat was 4- expected at the Boelter Reactor facility. Tc-99 provided a conservative estimate of the a

.i GM pancake probe efficiency for the two prevalent radionuclides at the Boelter reactor

'o .

site, Co-60 and Eu-152 (based on the isotopic analysis of the activated monolith concrete).

. The conseivatism resulted from the fact that the beta endpoint energy of Tc-99 (292 kev) t is less than that of Co-60 (318 kev) and Eu-152 (696 kev). The higher beta energies f C -60 and Eu-152 resulted in higher actual efficiencies than that detennined with Tc-O i 99.

I i

l LSC counting vials were analyzed on a Packard Model 2500 TR liquid scintillation 1

.O counter with a dual-label protocol for 11-3 and C-14 (named "3H 14C-OPEN WINDOW"). +

~

1

• j Three regions were established for the smear analysis, Region A (0 - 12 kev) for H-3.

I 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.

{0 Instrument ilse Techniques f

Initially, a 1007c 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 O pattern of a five 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 O of which was taken at die location of the highest direct contamination reading. The l

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 of low energy beta emitters with a liquid scintillation counter (LSC). The LSC t

28 O

O analysis functioned primarily as a screening tool to ensure that the surface contamination levels of 11-3 and C-14 were below the release criteria.

O The liquid scintillation smears were placed in 20 ml 11 ass vials and prepared for counting.

One milliliter of a 50%50G water and alcohol solution was dispensed on IJE smear paper to facilitate clution of the collected activity. Approximately 10 ml of Ultima Gold XR 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 acuvity results as long as only H-3 and/or ,

C-14 were present in the sample. However, when higher energy beta emitters were present w.g., Co-60, Eu 152) the activities reported were overestimated due to the spilldown of counts into the lower energy regions. For this case, conservative activities wem reponed for 11-3 and C-14. 4 Dose rate measurements were taken for each floor grid (and the lower walls of the Reactor Room) with a Bieron Micro-Rem meter. The dose rate measurements were taken at approximately 1 cm and I m from the surface. O Procedures Followed 9

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-20S2 (Ref. 2).

4' Surveying Organi:ation The surveying organization consisted of Mr. Abelquist in the role of lead weyor assisted by Mr Needrith, one senior radiation technician, Mr. Patrick llorkman, at junior radiation O technicians. Mr Mark Wachowski and Mr. James Castle. All final surveyi> g w as overscen by the UCLA Radiation Safety Office and Mr. Scott State of Dames & Moore, the radiological engineer retained by UCLA for the decommissioning project.

9 29 1

O i

1 0

SURVEY FINDINGS O Techniquesfor Reducing! Evaluating Data The minimum detectable activity (MDA) was calculated for both the fixed contamination survey instrumentation (i.e., Ludlum 2221 and 44-9 GM pancake probes) and the smear counter (i.e.,

O Ludlum 2929). 'Ihe MDA was calculated by the following equation (Ref. 3):

2.71 R 3 R3

+ 3.29 -+-

T, g T3 T, (1) gg3 ,

0 (eDiciency) (E'*

100 cm 2

"'")

where,

)

O R3= Background counting rate (cpm),

t T, = Background count time (min), and O T, = Sample count time (min).

The MDA for the Ludlum 2221 and 44-9 GM pancake proix was calculated in the same units as 2

the fixed contamination reaults (dpm/100 cm ). As an example, the MDA 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 mm, 50 cpm

+ 3.29 0.5 min MDA = 0.5 min

% 1 mty (2)

(0.18 c/ dis)( 15 ca.' ,

100 cm 2 r,

O MDA = 1690 D* . (3) i

. 100 cm2

'O 4

- 30 4

4 0 .

O Thus, the h1DA is approximately one-third of the average surface contamination level (5000 dpm ef/100 cm2 ) in NRC Regulatory Guide 1.86 (Ref. 4).

O The 51DA was calculated in a similar manner for the Ludlum 2929 smear counter. Ilowever, no 2

correction for probe area was necessary to convert to units of dpm/100 cm since the smeared i surface area was approximately 100 cm2 . Also, the background counting rate for the Ludlum 2929 l

eI was detennined by a thiny (30) minute count (typical background was about 80 cpm). The SIDA for the smear counter was approximately 320 dpm/100 cm2 , about one-third of the removable surface contamination level (1000 dpm p //100 cm2 ) in NRC Regulatog Guide 1.86.

1 9l The h1DA for the liquid scintillation counter was extremely low due to the low background counting rates (e.g.,12 cpm for 11-3) and high counting efficiencies. Each LSC smear, as well as the background smear, was counted for ten minutes. The ef6ciencies for 11-3 and C-14 of 67%

and 96%, respectively, resulted in extremely low minimum detectable activities (about 8 dpm/100 $l 2

cm for 11-3).

l The survey data were recorded onto the appropriate survey fonn. The direct contamination l readings were converted to dpm/100 cm 2 using the following fommla (probe area for the Ludlum O

l 44-9 GN1 pancake probes was 15 cm2 ):

l

&m/100 cm2 = gr s cpm - background cpm O.

(egicienn-) (probe area)

(4) 100 cm2 f

The average background count rate for each Ludlum 2221 and 44-9 GN1 pancake probe was O l detemlined by a series of three 1 minute counts. Each direct measurement of fixed contamination was thirty seconds in duration. If the direct measurement (correctedfor a 1 minute count time) resulted in a value less than or equal to the background count rate, the net count rate was given 8

a value ofI cpm. For e.tample, if the Ludlum 2221 and 4J-9 probe had a background count rate of 50 cpm and a 30 second direct measurement resuitcJ in 23 counts (gn ss count rate equals 46 cpm). the net count rate would />e given a value of I cpm (as opposed to the actual net < Junt ra:e of-4 cpm) and then converted to d;>m.'100 cm2 by dividing by the efficiency and correc dngfor the e 3i O

D probe area. Ahhough this practice biases the average surface contamination results, the effect is the documentation of conservative survey results. The average of the live direct contamination C measurements was calculated for each grid and compared to the average surface contamination level (50(X) dpm -1100 cm )2 in NRC Regulatory Guide 1.86.

Statistical Evaluation O

Limited statistical evaluation was performed to analyze the survey findings. The standard errors in each of the average surface contamination levels were calculated by propagating the error in the variables of Equation (4). Specifically, the errors in the gmss count rate and background count rate were obtained from the application of Poisson statistics and the method of detemiining the error in the counting efliciency 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 O 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.

l 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 ) n NRC Regulatory Guide 1.86. The highest average i

surface contamination levels were found in the process pit sump, with readings of 2260 i 244, C

2420 1 249, 2050 237,2230 244, and 2600 253 dpm -1100 cm ,2 respectively, for the north, east, south, west and floor surfaces. Overall, there were very few fixed contamination l readings that exceeded the MDA for the instmmentation. A summary of the fixed contamination 3 readings for the floors are listed in Appendix C. The maximum and average readings for each l grid are listed.

l t

The removable cont.unination survey consisting of the two cloth smears per grid resulted in all

'3 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. However, 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 3

O the highest direct reading prior to decontamination was not always the highest direct reading following the decontamination cifort.

9 The LSC smear results were all below the removable surface contamination levels (l(00 dpm p-7/100 cm2 ) in NRC Regulatory Guide !.S6. The highest tritium surface contamination level identified was 516 56 dpm/ LOO cm 2 The standard error in the tritium activity was calculated by assuming a conservative error in the LSC cfticiency detennination of 109 and for calculational 2

purposes assuming no error in a smeared area of 100 cm ,

The dose rate survey resulted in all measurements being less than 5 prem/h above background (13 ,

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 2001)

The final release survey of the Control Room consisted of 97 grids. The Control Room O'

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 O

locations in the bathroom. Gamma spectrometry analysis attributed the readings on the floor and wall tiles to naturally occurring radioactivity. The removable contamination survey resulted in all measuremems less than MDA with the Ludlum 2929 and LSC. The twenty dose rate measurements at I meter ranged from i1 to 15 prem/h and averaged g; 12.6 prem/h. Background dose rate was 13 prem/h.

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: floor and lower walls O

33

)

i O!

O (below 2m) were im x Im grids; upper walls and ceiling were 2m x 2m grids.

O The fixed contamination sun ey resulied in all measurements less than MDA for the floor grids. While several grid locations on the walls were slightly above MDA, the averaged fixed contamination was below MDA. All smears were less than MDA with the Ludlum 2929 and LSC. The 48 dose rate measurements at I meter ranged from 12 to 16 prem/h O

and averaged 13.5 prem/h. Background dose rated was 13 prem/h.

The laboratory bench in the center of the Count Room was sun' eyed as a separate item.

Both the fixed and removable contamination measurements werc less than MDA, g

Storace Room (Room 1003) lO 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 ( 1m x Im grids on the floor and lower walls; 2m x 2m grids on the ceiling and upper walls).

lO The fixed cont:unination sun cy resulted in all readings less than MDA for the ceiling and walls, and only 1 measurement above MDA on the floor (1900 518 dpm/100 cm2 ). All smears were less than MDA with the Ludlum 2929 phoswich detector, while the highest LSC smear, on the west wall, indicated 49 8.8 dpm 'H/100 cm2 . The twenty dose rate measurements at 1 meter above the floor ranged from 13 to 16prem/h and averaged 14.2 O prem/h. Background dose rate was 13prem/h.

Reactor Room Floor (Room HH4

,0 The final release sun'ey of the Reactor Romu Floor consisted of 182 grids (lm x Im grids). The Reactor Room Floor was considered to have a high potential for radioactive contamination since the monolith demolition activities were conducted in the Reactor I Room. The Reactor Room Floor was divided into 4 quadrants; nonhwest, northeast.

34 O

}

o.

southwest, and southeast.

Thirty-two (32) elevated readings were identiRed on the reactor niom Hoor during the 9-final release sun cy. The elevated readings w ere defined as those " hot spots" exhibiting 2

averaged contamination levels exceeding approximately 3000 dpm p-yl00 cm . An Eberline ESP-1 and AC-3 alpha scintillation probe were used to check the " hot spots" for the presence of alpha contamination; no alpha contamination was detceted. The " hot spots" were decontaminated and resurveyed. Post-decontamination results were documented for the appropriate grid location. The highest ore-decontamination result was 43,60012,500 dpm q/100 cm ,2 while the highest po; 'c tamination result was 2220 ,

2 538 dpm -1100 cm .

The highest averace Exed surface contamination level was 1680 i 226 dpm p-yl00 cm2 in grid A1 of the southeast quadrant. All smears were less than MDA with the Ludlum e; 2929 phoswich detector while the highest LSC smear, on the northwest Door, indicated 516 56 dpm 'H/100 cm . 2The 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. O Reactor Room Catwalk

. 9-The Onal release survey for the Reactor Room Catwalk consisted of 45 grids. The Reactor Room Catwalk was considered to have a mcde' ate potential for radioactive contamination. The grid plan consisted of 25 2m x 2m grds on the top of the catwalk and 20 2m x 2m grids on the bottom of the catwalk (the metal grating on the west side g; was only surveyed from the top).

The fixed contamination survey resulted in all readings less than MDA for the bottom of the catwalk and only one measurement exceeded MDA on the top of the catwalk. All #

smears were less than MDA with the Ludium 2929 phoswich detector while the highest LSC smear, taken on top of the catwalk on the south side, was 76117 dpm 'H/100 cm2 .

35 1

I I

O i

l J

]

Reactor Room Menanine

} The final release survey of the Reactor Room Menanine consisted of 76 grids. The Reactor Room Menanine was considered m have a moderate potential for radioactive 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

) ceiling. A mantift was used to survey the Reactor Room upper walls and ceiling.

I 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.

D-Reactor Room Ramp The final release survey of the Reactor Room Ramp consisted of 57 grids. The Reactor 3

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.

D The fixed contamination survey resulted in all readings less than MDA. except for 2 measurements at 1650 502 dpm/100 cm2 . All smears were less than MDA with the I Ludlum 2929 phoswich detector, while the highest LSC smear, on the ramp floor.

D indicated 21 i 4.0 dpm "C/100 cm .2 The fourteen dose rate measurcments at I meter above the floor ranged from 12 to 16 prem/h and averaged 13.8 prem/h. The background dose rate was 13 prem/h.

J Reactor Room Walls 1

The final release survey of the Reactor Room Walls consisted of 123 grids. The Reactor l 3 Room Walls were considered to have a moderate potential for radioactive contamination. l The grid plan for the Reactor Room Walls consisted of im x im grids on the lower walls (up to 2m) and 2m x 2m grids on the upper walls.

3 36 s

D

O The fixed contamination survey resulted in all readings less than MDA. In addition, all smears were less than MDA with the Ludlum 2929 phoswich detector. The highest LSC smear (75 dpm/100 cm2 on the cast wall) was most likely from Co40 and/or Eu-152. O since a significant fraction of counts were detected in the higher energy channels. No uncenainty in the LSC actisity was calculated when a significant fraction of the counts were present in the higher energy channels. The 98 dose rate measurements at I meter from the lower walls ranged from 10 to 16 prem/h and averaged 13.1 prem/h. The background dose rate was 13 prem/h.

The overhead crane and ventilation duct on the east wall were surveyed as separate items. g The fixed and removable contamination measurements for both items were less than MDA.

Process Pit 9 The final release survey for the Process Pit consisted of 81 grids. The Process Pit was 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 O

for all surfaces.

Many of the fixed contamination grid locations were less than MDA, however, there were 9-locations on the south wall, process pit floor, and sump that exceeded the MDA. The highest aseraged fixed surface contamination levels were found in the process pit sump, with readings of 2260 244,2420 249,2050 237,2230 244. and 2600 1253 dpm

-y/100 cm2 , respectively, for the nonh. cast. south, west and floor surfaces. All smears g were less than MDA with the Ludlum 2929 phoswich detector, while the LSC smears indicated minimal (all smears less than 50 dpm -y/100 cm 2) removable surface contamination within the process pit. The 18 dose rate measurements at 1 meter above the noor ranged from 13 to 16 prem/h and averaged 14.2 prem/h. The background dose O' rate was 13 prem/h.

9:

37 9

O Fuel Storace Pits 9 The final release survey package for the thiny Fuel Storage Pits was documented differently than for the other Boelter f acility 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 Dve survey locations per grid were detennined by scanning O

each pit with the Ludlum 2221 and 44-9 GS1 pancake probe to identify the two highest measurements at the top (grid locations 1 and 2). the highest in the center of the grid (grid location 3), and the two highest measurements near the bottom of the grid (gnd locations 4 and 5). The grid locations were further identified by north, cast, south, or west 3

positions. The thiny Fuel Storage Pit plugs were each considered as one grid.

Two cloth smears were taken per grid and one liquid scintillation smear was taken per pit.

O An dose rate measurement was obtained from each pit by lowering the Bicron hiicro-Rem meter into the pit an ann's length and recording the highest reading. The BicrE hiicro-Rem meter was also used to take one dose rate measurement from each Fuel Storage Pit plug.

O The highest fixed contamination measurement was 1300 445 dpm -y/100 cm2 (Fuel Storage Pit C4), slightly greater than the 51DA (1270 dpm -y/100 cm2 ) for the Ludlum 2221 and 44-9 probe All of die cloth smears for both the Fuel Storage Pits and plugs J

were less than 51DA. The LSC smears were also less than N1DA.

Floor Drains and Pir>ine inle!! Outlets O

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 (l minute count time) was obtained at the access O point of each Door drain, drain line or valve. All direct contamination measurements were 2

less than the N1DA (1210 dpm p-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 51DA. Niinimal removable surface contamination was indicated by the LSC smears taken at each of the

,e 38 0

f

O' access locations. The interior surfaces of the four floor drains were surveyed by passing a Masslin cloth through the drain lines. A direct scan of the Masslin cloth with the Ludlum 2221 and 44-9 GM pancake probe indicated no presence of contamination. 9-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 liole 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 ,

soit samples were sent to a commercial laboratory for gamma spectrometry and analysis for 11-3 and C-14.

At the laboratory, the water in each sample was distilled and analyzed for tritium. The 9i results for tritium were as follows:

Hole 3 4.35 E3 pCi/l Hole 2 3.63 E3 pCi/l O-Hole 1 1.60 E3 pCi/l For comparison, the EPA drinking water standard for tritium is 2 E4 pCiil. .

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 Office performed numerous verification surveys for comparison with the contractor results. Direct exposure rate measurements with a Ludlum Model 19 survey meter O

39 O

P l- >

t t were in good agreement with NES results. Direct scans with a UCLA owned Ludlum 2200 ratemeter and a 44-9 probe were within approximately 107c of the NES results for fixed j k

i contamination at hot spots in each room. Swipes were taken for removable contamination f; L comparisons of the low energy beta emitters and analyzed with the Univer.sity LSC. All  !

I l verification swipes were less than MDA which was in agreement with the NES results. l l

}

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 }

j criteria. An unshielded high purity gennanium spectroscopy system was also set up in the reactor room for qualitative detennination of the most prevalent isotopes in the area. Approximately 50 photopeaks were observed dominated by naturally occurring isotopes of the uranium and thorium l series. Based on gross counts, no significant contribution from activation products was present

) in the room. j e

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- . . _ . , . _ _ . _ . . , _ . , _ _ . . . . . _ . - ,_ ..._ . _ _ ....-.....,._.,...-.._~....--.-..~.,.m.,,,m,~,,,,.-..

h i

O -j

SUMMARY

-)

O- The decontamination and dismantlement of the Boelter research reactor was successfully executed in a l pmfessional and timely manner. The resulting condition of tne facilities is such that they arc acceptable I 1

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  ?

'O Bridge the Gap. j

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REFERENCES t

O 1. NES Procedure 82A8021. " Final Survey Procedure;" 1992. I

2. NUREG/CR-2082. " Monitoring for Compliance with Decommissioning Tennination Survey  ;

Criteria;" 1981.

O

3. Strom. Daniel J. and Stansbury, Paul S. " Minimum Detectable Activity When Background is Counted Longer Than The Sample." llealth Physics 63(3):360-361; 1992.

O

4. NRC Regulatory Guide 1.86. " Termination of Operating Licenses for Nuclear Reactors;" 1974.

I

5. NES Operational Survey # 92-245 l t

0 i'

6. UCLA Decommissioning Phase i Report.

7. NES Final Survey Report O -

8. NES Final Decommissioning Repon for the Boeller Reactor Facility Dismantlement and Final ,

Decommissioning Project, November 1992.

O i

, Y 9

O.

O.

t O ,

I 42 i

,O

.I ,

APPENDIX A 4

4 PROJECT SCHEDULE O

O O -

O O

1 l

l b

l August l hp11mbst l October Novsmber ID Name Durenon 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 l11/15l11/22 1 PROCEDURE DEVELOPMENT 6w 7/20/92 8:OOsm [ pr , q 2 DESIGN L'1PPING CONTAINER 1.5 w 7/23/92 8:OOam p- g 3 MO%I2E STAFF / SITE MOBILIZ 7.38d 8/3/92 8:OOam 4 PREP SITE FACILITIES 7.5d 8/3/92 8:OOam J ;, j 5 INITIAL SITE INSPECTION /SUR 1.5 w 8/3/92 8:OOam u

6 INITIAL SURVEY OF MONOLITH 1.25d 8/4/92 8:OOsm y 7 STAFF TR AINING/ORIENTATIO 1.25d 8/10/92 8:OOam y 8 REMOVE WELD BLOCKS FOR Sd 8/11/92 8:OOsm l 1

9 SHIP COPJAMINATED BLOCKS 3.75d 8/14/9210:OOsm g l 10 CORE BORC BEAM PORTS 1.5 w 8/15/92 8:OOam  :

11 CUT BEAM PORTS INTO SECTI f 6.25d 8/20/921:OOpm .

~

12 PKG. BEirIPORT SECTIONS F 2.5d 8/26/92 1:OOpm g 13 SETUP MONOLITH CONTAIN 2.5d 8/10/92 7;OOam I I

14 1.25d 8/12/92 7:OOam l l REMOVE ACTIVATED STEEL C l 4 -

+

15 PKG. ACTIVATED CHANNELS F l 1.25d 8/13/92 7;OOam l

1 16 REMOVE ACTIVATED MONOLIT 4.5 w 8/14/92 7:OOam i i I 17 PKG. ACTIVATED CONCRETE F 4.5 w 8/17/92 7:OOam l

18 INTERIM SURVEY INSIDE MON 3.75d 9/9/9210:OOsm 19 RMV CLEAN CONCRETE SO. SI Iw 9/15/92 8:OOsm l 20 RMV CLEAN CONCRETE NO. SI 1w 9/15/92 8:OOam 21 RMV CONCRETE T'O DIRT LEVE 1,75d 9/22/92 8;OOam g {

22 RMV CONTAMINATED DIRT 3d 9/23/92 3:OOpm ,

i 23 RMV CONTAMINATED DRAIN L 2d 9/23/92 3:OOpm g j 24 SURVEY COUNTING, STORAGE 6.25d 8/3/92 8:OOam 4 1

25 DECON FUEL STOR AGE PITS 1.5w 9/14/92 8:OOam 26 FINAL RELEASE SURVEY STOR 6.25d 9/1'J2 8:OOsm .  !

Critical 13

• Progress ENEMEM M M EMI'EME Summary Project: UCLA DECOMMISSIONING tl Dats: 11/20/92 Noncritical Milestone $ Rolled Up O Page 1 0 _

e _

e e e e , _

e e .

• e e

'O O O 'O O O O O O- 0 O JJ l August ~l Sept:mber l Octobcr Newsmbsr ID Nome Duration Scheduled Start 7/19 l 7/26 l 8/2 l 8/9 l 8/16 l 8/23 l 8f30 l 9/6 l 9/13 l 9/20 l 9/27 l 10/4 l10/11l10/18l10/25 11/1 l 11/8 l11/15j11/22 27 ISOLATE FUEL STORAGE PITS Sd 9/21/92 8:00am j 28 DECONTAMINATE & REMOVE 1.25d 8/11/92 7:00am l 29 3.75d 8/27/92 7:00am SETUP PROCESS PIT CONTAIN

]

30 PROCESS SLUDGE, PACKAGE 1.5 w 8!31/92 7:00a;.. g i

f I

31 RMV PIPING & ITEMS FROM PR 1.5w 8/31/92 7:OOsm g '

32 DECON TANK & PREPARE FOR 3.75d 8/31/92 7:Comm h ,

33 DECON PIT WALLS & FLOOR 3w 9/3/921:00pm 34 DECON SUMP & RMV HARDW 3,75d 9/1/92 8;OOam 35 PKG CONTAMIN ATED SUMP M 3.75d 9/4/92 3;OOpm

> 36 FINAL RELEASE SURVEY OF PI 5d 9/14/92 8:OOam 37 FINAL SITE CLE ANUP 5d 9/28/92 3:OOpm g 38 WASTE SHIPMENTS 5.41 w 8/31/92 7:00am 1.; - , .,l  !

39 FINAL SITE RELEASE SURVEY 3.2w 9/21/92 8:OOsm p-  ;;  !

40 DEMOBILIZATION 1.18 w 10/8/92 8:OOsm l _j i

41 PREPARE FINAL REPORTS 3w 10/6/92 3:12pm s"]

l- l Project: UCLA DECOMMISSIONING Critical in +1 Progress - Summary Date: 11/20/92 Noncritical Milestone $ Rolled Up D Page 2

a .

1 APPENDIX B )

o j l

a CALCULATION OF EFFICIENCY a UNCERTAINTY a

O e- ,

O:

Efficiency Determination fur the L2221 (# 73700) and 44-9 (#066761)

Source Activity (A): 2060 80 dpm Tc-99 #:

Procedure: Ten (N = 10), I minute measurements of both background and source (T,,=T,.3 = 1 min) were obtained with a reproducible source-detector geometry.

O Data: Backcround Counts (C,J Source Counts (C. ,J 82 519 84 520 04 71 486 79 511 81 546 84 470 0

77 SN 76 522 86 499 ,

75 474 g.

Mean background count (C3 ) = 79.5 counts Mean source count (C,. ) = 505.1 counts The experimental standard deviation, o,,, was calculated with an HP42S calculator for both distributions:

O o,y for the background count = 4.74 counts o,, for the source count = 23.6 counts The standard error in both the mean background count and source count was calculated fmm the 8-experimental standard deviation:

0 O,"

c i VN g-Thus, C, = 79.5 i 1.5 counts and C,.3 = 505.117.46 counts. The respective counting rates are calculated as follows:

O R3= C,, /Tn = 79.5 counts / min R,,,= C,.,,ffm = 505.I counts / min O.

3 The standard error in either the background or source counting rates, ca , is calculated by propagating the error in the number of counts:

3 'aR 2 <3 2

of o,"

c ~~.

O,c i

SC 1 7

Since both the background and source were counted for 1 minute and no erro was assumed in the 3 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 rate error:

1 \1

' c. -

' 1.5 '2 = 1.5 o ,* = - -

g  %,T, T,1, The net counting rate, R, , is equal to Rs ., - R, , or R, = 425.6 counts / min. The error in the net counting rate is propagated as follows:

O o,, = /(og)2 + (o,,)2 = /(1.5)2 +(7.46)2 = 7.61 counts / min Tims, R, = 425.6 i 7.61 counts / min.

O The counting efficiency is calculated by dividing the net counting rate by the source activity:

=

0' '*"""

eDiciency = A = 0.21 counuldis 2060 disimin O

Pmpagating the error in both the net counting rate and the source activity results in the following equation:

' os, '(-R,)(aa)'2

"**"n ' y J, * , 32 ,

O Substituting values into the equation yields:

, '7.61,2 . '(-425.6)(80)'*

,' #"" = 0.0088 counts /du 9 \i2060; s (2060)2 ,

Thus, the counting ef ficiency is equal to 0.21 0.0088 counts / dis.

!D 0

O-APPENDIX C O

'O

SUMMARY

OF FINAL FLOOR SURVEYS O

O O

O O

O 6

O

l O'

Northeast (NE) Southeast (SE)

! i i l l l l l l l l I ! #'

F1 2 3 4 5 6 7 F1 2 3 4 5 6 7 E E 4-D D

=

C C...

B Bl 8' A

- f

/

^

G1 2 3 4~ -! 5 L 6 7: .'G1 2 3 4 5 6 7 _

t __

F F_

E s E1

. O D D!

= e a se a C C B B . 9-

'h A I I

i A '

i i i f Northwest (NW) Southwest (SW) l l Below floor grade to foundation level ld Below foundation grade Figure 1: REACTOR ROOM FLOOR SURVEY O

9:

Cl 9

O kmtor Room floor NE N% sw NL f

e In noon 3 ydpm/ih0cm i p r m ,/h p y @m/; Am ge ' y ir @ m; A m'

. ,mA f }y J;m r. Am rtsn/h at I mu at 1 ma at i Wr at 1 mir Mas Asg Mn Asg NLu Asg N1n Asg Al 1330 377 16 6M 4%7 14 H 00 I t,M O 16 1480 615 16 2 667 33 13 120a 444 12 1270 54 13 14h0 ID60 13 3 592 372 14 44 141 14 ID!D 569 15 1440 1160 15 g 4 hK9 578 12 507 :66 13 1140 7 35 13 1260 710 13 5 963 481 15 1520 7% 14 761 546 14 1700 740 12 6 2000 1110 15 1330 7N 14 451 558 14 815 48 14 7 1630 796 14 lux 0 641 15 571 368 12 1850 814 14 HI 815 563 13 627 14 13 340 1140 15 ING 638 15 2 963 652 13 571 273 13 1520 608 15 1560 B01 14 3 1780 405 15 MbH 3M 13 !460 5 46 16 1630 9M 15 4 592 252 13 451 450 15 1270 774 15 889 533 12 g 5 1410 637 11 951 44 16  !!40 760 13 1480 9 32 14 6 2300 llRO 13 1010 KK7 13 6 34 266 15 1480 829 13 7 951 479 16 951 4% 14 1520 811 13 1330 5 92 13 Cl 815 484  !$ bM 222 i '4  !!40 7h6 14 INO 7k6 14 2 1110 815 14 507 273 12 1520 824 15 1560 904 15 3 1780 860 12 6M uis 14 10!0 451 12 1110 570 16 4 063 607 12 1010  ?$7 12 1010 664 14 1260 815 13 5 1460 692 15 1080 413 13 1310 697 13 1110 755 13 g 6 1920 977 16 1010 444 12 697 317 12 963 592 13 7 1520 #77 15 1010 570 14 697 482 15 667 385 13 D1 963 469 13 571 294 12 761 349 13 1180 725 15 2 1040 34 12 8x8 514 13 1460 4 25 15 2220 1050 14 3 8h9 459 11 44 %1 13 1140 7fo 14 1330 918 14 4 667 489 14 6M 496 15 824 4 25 14 1040 6 13 14 5 667 341 13 647 4 31 13 1520 729 13 1180 725 14 6 1330 1020 13 c51 59 14 761 311 ll 1180 740 14 H7 444 667 363 14

] 7 fl 1780 1260 927 4K5 12 11 1010 451 514 12 11 IND 241 H5 li 13 1780 1150 14 2 741 3h5 16 3x0 165 15 1340 k 36 14 697 285 13 3 741 355 12 shR av 14 761 222 15 1140 7M 13 4 1260 865 14 14t>0 628 14 1200 8% 13 951 799 14 5 2280 871 12 824 355 13 888 241 15 1140 6M 15 6 EN9 503 15 Itil o 595 14 1140 754 14 1330 615 12 7 518 224 14 Ofnc 1000 15 1650 799 13 l'11 0 54 13

] fl 2

741 592 296 385 16 13 824 571 323 181 12 12

, ??O iNO 609 607 15 13 1580 888 932 463 15 13 3 1330 4% 14 !340  %: 12 1330 7:2 14 1390 633 14 4 1560 593 12 1:00 60x 14 1460 $89 13 1010 595 13 5 1080 423 12 c5t 647 13 1010 755 11 IMO 709 14 6 2300 845 16 1580 AN 15  !!40 7M 10 76) 450 14 7 1140 524 15 INO tho 14 507 273 11 647 444 13 G1 ha 2.2 13 951 6M 16

] 2 3

317 317 l .' l 14 11 1140 1010 374 335 12 15 4 1140 kP 14 1520  !!:0 14 5 64 246 1% 1520 1060 15 6 Mi na 15 10R0 337 15

< 17h:1 Nu 15 1340 551 13

__ ) -

] MDA: IE60 & 1940 dpm/louem' MD A.- I MD , - l Nd MD A; 15'O @milWm' MDA: 1570 & 1780 dpm/100ana l m, o , ma - m,m -

Jm r o- i e

C2 3

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1 r- 0 3 1-4 5 .

6; ......

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O i i F O Below floor grade to foundation level O Below foundation level O E] Sump Figure 2: Process Pit Floor and Sump Survey 9

O C3 4

a Process Pit Floor 3 Location g dpm/100cm 2 prem/h at 1 meter Maximum Average Al 517 146 14 2 507 133 15 D 3 254 95 14 4 127 57 14 5 444 368 15 6 507 235 14 7 127 82 16 3 B1 317 96 15 2 634 374 13 3 697 298 13 4 571 235 14 5 888 482 15 6 317 273 15 3

7 697 380 14 Cl 5135 2050 13 2 571 266 14 3 254 152 15

, 4 507 235 13 s

MDA: 1600 dpm/100cm 2

Background:

13prem/h D

Sump Location q dpm/100cm 2 prem/h at I meter Maximum Averace J

N1 4250 2260 -

El 4370 2420 -

S1 3170 2050 -

Wl 5520 2230 -

3 F1 3870 2600 -

MDA: 1600 dpm/100cm 2

Background:

13prem/h O

C4 0

O l l l 1 f f

F 9

COUNTING ROOM E D e C

B 8 7 6 5 4 3 2 A1 el A B C D1 i

TRANSFORMER VAULT STORAGE ROOM 2 el 1

3 I

• l 4 l t 5

O O

- . r - $;

7 6 5 4 3 2 Al REACTOR RAMP AREA B 4 l  !

!  ! !h Figure 3: Coun:ing Room, Storage Room, and Ramp Area Floor Surveys (Not to Scale) e.

C5 e

r S

Counting Room Storage Room Ramp Area laatwn D Y dpm/100em prenth {i-y dpm/liJkm' prenth D 't d pnt'liN k m2 prem,h O at I mtr at 1 mir at I mtr Max Asp Mas Asg Max Asg At 380 235 13 888 495 15 1350 976 14 2 1010 557 15 951 647 13 761 520 13 3 951 558 15 1270 799 14 1010 860 12 4 951 596 13 1080 546 15 1390 1050 14 O 5 634 418 15 1900 1000 14 697 494 15 6 444 235 14 N'A N/A N/A 1140 595 12 7 697 431 13 N/A N'A NrA 824 533 13 8 507 260 14 N:A N'A NA N!A N!A N/A Ill 761 368 16 1080 622 13 951 514 14 2 444 247 15 1010 709 15 1010 671 16 3 697 469 14 888 571 14 1080 774 13 4 571 380 14 1390 887 16 1140 766 14 5 697 507 14 951 545 14 1650 876 14 6 824 545 14 N/A N/A N/A 1010 747 13 7 507 279 13 N'A N'A N/A 1650 1100 16 8 697 456 13 N:A N:A NlA N;A N/A NIA Cf 697 406 13 951 558 15 2 888 494 13 951 685 12 0 3 761 444 14 951 602 15 4 697 456 13 1460 565 15 5 N/A N/A N/A 1010 646 14 6 NA N!A N:A N!A NA N%

7 507 380 13 /"A N/A N/4 8 697 456 13 N'A N 'A N/A Di 1140 558 13 888 400 14 2 761 469 14 761 571 14 3 697 456 14 761 545 14 4 824 393 13 1270 660 14 5 697 387 14 697 437 14 6 888 545 14 NA NA VA 7 951 596 13 g 8 El 571 634 317 342 14 13 2 507 342 14 3 697 374 12 4 634 323 14 5 697 374 12 6 824 387 13 g 7 1010 544 14 8 824 482 14 F1 697 387 14 2 697 393 13 3 888 387 12 4 824 507 12 5 888 583 13 O 6 7

697 951 431 482 14 14 8 634 311 12 MDA: 1600 dpm!!Okm 2 M DA.1550 dpm/lofk m MDA 1570 dpm/l(Ocm2 Ilkg: 13 prentt likg.13 prenwh likg: 13 prenth O

C6 O

O 1 l-r \ / 3 7  !) 9 10 11 12 13 14 r  %

• 6 15 l

9 5

16 4 .

4  ;

. 17 9 i

.- ] [.)

3 18 e

2

'h / 19

'4 iiiii iiiiiiiiiii iiiiiij iiii . .

I 25 23 " 71 20 i i i  :::

r i

i  !

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.] I i i- i i-e Figure 4: Reactor Room Catwalk: Balcony (Not to Scale)

O S

O C7 e

. _ _ _ _ _ _ _ _ _ _ _ E'

P D

Reactor Room Catwalk: Balcony 3 Location p / dpm/100cm 2 Maximum Average 1 1080 615 2 1460 514 D 3 1390 912 4 1580 1170 5 951 526 6 761 596 7 1080 470 8 1780 888 3

9 1140 760 10 507 222 11 1270 774 12 1010 722 13 1580 709 g

14 1200 747 15 1270 723 16 1140 810 17 951 438 18 824 209 O 19 127 57 20 127 51 21 190 76 22 127 51 23 127 51 8 24 127 57 25 254 76 MDA: 1690 dpm/100cm 2 O

e O

C8 0

b O

iI l l 1 I A B C1 - 9

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S 2

O CONTROL 3 ROOM O

4 __

4 5

il V J B1 Al C _

CHANGE J U 6 ROOM Al 1 n 8 B2 A2 I

i i i Y

Figure 5: Control Room and Change Room Floor Surveys (Not to Scale) g C9 9;

O Control Room Location -/ dpm/100cm 2 prem/h at 1 meter 3

Maximum Average ,

A1 63 38 14 2 190 64 14 3 3 190 64 13 4 63 38 12 5 127 57 13 B1 254 75 12 2 127 51 13 3 507 159 12 3 4 32 32 12 5 32 32 13 C1 444 114 13, 2 254 83 11 3 824 209 12 0 4 32 32 13 5 127 51 13 6 444 133 12 MDA: 1570 dpm/100cm 2

Background:

13prem/h O

Change Room O Location -y dpm/100cm 2 prem/h at I meter Maximum Average A1 1460 849 12

, A2 1200 760 14 B1 1520 1100 13 B2 1780 800 11 MDA: 1570 dpm/100cm 2

Background:

13prem/h 6

O C10 9