Text

Decommissioning Plan for Triga Mark III
	ML20147D769
Person / Time
Site:	Berkeley Research Reactor
Issue date:	01/31/1988
From:	; KAISER ENGINEERING (FORMERLY KAISER ENGINEERS), U.S. ECOLOGY, INC. (FORMERLY NUCLEAR ENGINEERING
To:	;
Shared Package
ML20147D750	List: ML20147D744; ML20147D760; ML20147D769;
References
	87-039-R, 87-39-R, NUDOCS 8801200293
	Download: ML20147D769 (144)
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DECOMMISSIONING PLAN FOR THE TRIGA MARK i

BER<ELEY RESEARCH REACTOR

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prepared for University of Cali"ornia, Ber<e ey prepared by Kaiser Engineers (California) Corporation In association with US Ecology, Inc.

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,P DCD Facility License No. R 101 Report No. 87 039 R Docket No. 50 224 January 1988

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CONTENTS O

CHAPTER 1. BACKGROUND AND MANAGEMENT PLAN 1-1

1.0 INTRODUCTION

1-1 1.1 SUWARY DESCRIPTION 1-1 1.1.1 Reactor Facility Description 1-1 1.1.2 Reactor Description 1-7 1.1.3 Operating License History 1-13 1.1.4 The Decommissioning Approach 1-13 1.1.5 Decommissioning Alternatives 1-14 1.1.6 Availability of Funds 1-14 1.1.7 Major Tasks and Schedules 1-14 1.1.8 Quality Assurance Plan 1-15 1.1.9 Executive Engineer / Contractor Participation 1-16 1.1.10 Termination Radiation Survey Plan 1-17 1.1.11 Estimated Personnel Dose Equivalent 1-17 1.2 OPERATING HISTORY 1-18 1.2.1 Reactor Energy Production 1-18 1.2.2 Radioactivity Release / Discharge 1-18 1.2.3 Experiments and Related Unusual Events 1-18 1.3 CURRENT RADIOLOGICAL STATUS OF FACILITY 1-18 1.3.1 General 1-18 1.3.2 Neutron Flux 1-28 1.3.3 Neutron Activation Analysis 1-30 1.3.3.1 Irradiation Times 1-30 1.3.3.2 Neutron Energy Groups 1-30 O

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CONTENTS (Cont) 1.3.3.3 Material Compositions 1-31 1.3.3.4 Significant Neutron Activation Reactions 1-31 1.3.4 Material Activation Strength and Surface Dose-Rate Calculations 1-31 1.3.4.1 Core Support Assembly and Rotary Speciman Rack (Lazy Susan) 1-35 1.3.4.2 Reactor Position at Exposure Room 1-36 1.3.4.3 Reactor Position at Pool Center 1-37 1.3.4.4 Reactor Position at Thermal Column 1-38 1.3.5 Radiation Dose Rates Estimate 1-39 1.3.5.1 Core Support Structure 1-40 1.3.5.2 Inside Exposure Room 1-40 1.3.5.3 Inside Reactor Tank 1-41 1.3.6 Estimate of Volume of Activated Material to Be Removed 1-41 1.3.6.1 Exposure Room 1-42 1.3.6.2 Reactor Tank, Wall, and Floor 1-42 1.3.6.3 Thermal Column 1-42 1.4 DECOMMISSIONING ALTERNATIVE 1-42 1.5 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES 1-43 1.5.1 Decommissioning Steering Committee 1-43 1.5.2 Decommissioning Program Committee 1-43 1.5.3 Reactor Hazards Committee 1-45 1.5.4 Office of Environmental Health and Safety 1-45 1.5.5 Radiation Safety Committee 1-46 1.5.6 Physical Security 1-46 1.5.7 Planning and facilities Management, Design and Construction Services; Executive Engineer 1-46 iv

CONTENTS (Cont) 1.5.8 Decomissioning Contractors 1-47 1.5.9 UC Decomissioning Project Engineer 1-47 1.5.10 Deputy UC Decomissioning Project Engineer 1-47 j

1.5.11 Reactor Health Physicist 1-48 1.5.12 UC Quality Assurance Supervisor 1-48 C

1.6 REGULATIONS, REGULATORY GUIDES, AND STANDARDS 1-48 1.6.1 Applicable Regulations 1-49 1.6.1.1 State of California 1-49 1.6.1.2 Code of Federal Regulations 1-49 1.6.2 Regulatory Guides 1-50 1.6.2.1 NRC Regulatory Guides 1-50 1.6.3 Standards 1-51 1.6.3.1 ANSI Standards 1-51 1.6.4 Informal Guidance ard Technical Reports 1-51 1.6.5 Permits / Licenses Covering 1140 Etcheverry Hall Activities 1-52 1.7 TRAINING AND QUALIFICATIONS 1-52 l

1.7.1 Training Program Descriptions 1-52 1.7.2 Administration and Recordkeeping 1-53 i

CHAPTER 2 OCCUPATIONAL AND RADIATION PROTECTION PROGRAMS 2-1 l

2.0 INTRODUCTION

2-1 2.1 RADIATION PROTECTION PROGRAM 2-1 2.1.1 Personnel 2-1 2.1.2 Training 2-2 2.1.3 Administrative and Radiological Controls 2-3 y

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CONTENTS (Cont) 2.1.3.1 Exposure Limits 2-3 2.1.3.2 Radiation / Hazardous Work Permits 2-6 2.1.3.3 Controlled Surface Contamination Area 2-9 2.1.4 Radiation Protection Facilities, Instrumentation, and Personal Protective Equipment 2-10 2.1.4.1 Facilities 2-10 2.1.4.2 Instrumentation 2-11 2.1.4.3 Personal Frotective Equipment 2-12 2.2 INDUSTRIAL SAFETY AND HYGIENE PROGRAM 2-15, 2.2.1 Personnel 2-15 2.2.2 Training 2-16 2.2.3 Administrative and Work Practice Controls 2-16 2.2.3.1 Exposure Limits 2-16 2.2.3.2 Inspection and Audit Programs 2-17 2.2.3.3 Medical Surveillance Program 2-18 2.2.3.4 Hearing Conservation Program 2-19 2.2.4 Operational Activities 2-19 2.2.4.1 Fire Protection and Prevention 2-19 2.2.4.2 Hand and Power Tools and Cutting Equipment 2-20 2.2.4.3 Lifting Equipment 2-20 2.2.5 Personal Protective Measures 2-20 2.2.6 Excavations 2-22 2.3 EXECUTIVE ENGINEER / CONTRACTOR ASSISTANCE 2-22 2.4 COST ESTIMATE 2-23 2.4.1 Cost Estimate Elements 2-23 2.4.2 Assumptions 2-25 vi

CONTENTS (Cont)

CHAPTER 3 DISMANTLING AND DECONTAMINATION TASKS AND SCHEDULES 3-1

3.0 INTRODUCTION

3-1 3.1 TASKS AND ACTIVITIES 3-1 3.1.1 Prior to Dismantling 3-1 3.1.1.1 Removal and Relocation of Classroom Equipment 3-1 3.1.1.2 Removal and Shipment of Fresh Reactor Fuel 3-1 3.1.1.3 Removal and Shipment of Irradiated Reactor Fuel 3-3 3.1.2 During Dismantling 3-3 3.1.3 Task Descriptions 3-7 3.1.3.1 Task 1: Contractor Move-in 3-7 3.1.3.2 Task 2:

Initial Radiation Survey 3-7 3.1.3.3 Task 3:

Installation of Confinement Barriers 3-8 3.1.3.4 Task 4:

Removal of Reactor Components and Pool Liner (Groups 1 and Part of 3) 3-11 3.1.3.5 Task 5:

Removal of Material With Potential Surface Contamination and Other Activated Materials (Group 2 and Part of Group 3) 3-11 3.1.3.6 Task 6: Cleanup and Removal of Tools and Equipment 3-11 3.1.3.7 Tash 7:

Packaging and Shipment of Radioactive Waste 3-12 O

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CONTENTS (Cont)

O 3.1.3.8 Task 8:

Perform Termination Radiation Survey 3-12 3.1.3.9 Task 9: Demolition of Non-Activated /

Contaminated Portion of Ractor Installation 3-12 3.2 SCHEDULE 3-12 3.2.1 Milestones 3-15 3.3 TASK ANALYSES 3-15 3.3.1 Task 2:

Initial Radiation Survey 3-15 3.3.1.1 Reactor Room Initici Conditions 3-17 3.3.1.2 Reactor and Biological Shield Assembly 3 17 3.3.1.3 Etcheverry Hall Interior Conditions 3-18 3.3.1.4 Etcheverry Hall Exterior Conditions 3-18 3.3.1.5 Evaluation of Initial Survey Results 3-19 3.3.2 Task 3:

Installation of Confinement Barriers 3-19 3.3.3 Task 4:

Removal of Reactor Components and Pool Liner 3-20 3.3.4 Task 5:

Removal of Components With Potential Surface Contamination and Other Activated Materials 3-21 3.3.5 Task 6:

Cleanup and Removal of Tools and Equipment 3-23 3.3.6 Task 7:

Packaging and Shipment of Radioactive Waste 3-24 3.3.7 Task 8:

Perform Termination Radiation Survey 3-24 3.3.8 Task 9: Demolition of Non-Activated /

Contaminated Portion of Reactor Installation 3-25 CHAPTER 4 SAFEGUARDS AND PHYSICAL SECURITY 4-1

4.0 INTRODUCTION

4-1 4.1 DECON AREA 4-1 viii

CONTENTS (Cont)

CHAPTER 5 RADIOLOGICAL ACCIDENT ANALYSIS 5-1 CHAPTER 6 RADIOACTIVE MATERIALS AND WASTE MANAGEMENT 6-1

6.0 INTRODUCTION

6-1 6.1 FUEL DISPOSAL 6-1 6.2 LIQUID RADWASTE 6-1 6.3 SOLIO RADWASTE 6-1 6.3.1 Packaging 6-2 6.3.2 Container Handling 6-3 6.4 VENTILATION SYSTEM 6-3 6.5 WASTE CLASSIFICATION 6-3 6.6 SHIPPING OF RADIOACTIVE WASTES 6-4 CHAPTER 7 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS 7-1

7.0 INTRODUCTION

7-1 7.1 HEALTH AND SAFETY LIMITS 7-1 7.1.1 External Exposure 7-1 7.1.2 Internal Exposure 7-1 7.1.3 Concentration of Airborne Radioactive Material in Restricted Areas 7-1 7.1.4 Concentration of Airborne Radioactive Material in Unrestricted Areas 7-2 7.1.5 Concentration of Non-Radioactive Substances in Restricted Areas 7-2 7.1.6 Concentration of Non-Radioactive Substances in Unrestricted Areas 7-2 7.1.7 Noisc Levels 7-2 7.1.8 ALARA 7-2 7.1.9 Health and Safety Limits for Unrestricted Use 7-2 ix

CONT ENTS (Cont) 7.2 SURVEILLANCE REQUIREMENTS 7-3 7.2.1 The Environmental Dosimeter Program 7.2.2 The Routine Swipe Program 7-3 7.2.3 The Routine Instrument Survey Program 7-3 7.2.4 The Air Sampling Monitoring Program 7-3 7.3 ADMINISTRATIVE CONTROLS 7-3 7.3.1 Administrative Controls During DECON 7-3 7.3.2 Responsibility 7-4 7.3.3 Organization 7-4 7.3.4 Records and Reports 7-4 7.4 ENGINEERING CONTROLS 7-5 CHAPTER 8 - PROPOSED TERMINATION RADIATION SURVEY PLAN 8-1

8.0 INTRODUCTION

8-1 C.1 PRELIMINARY SURVEY 8-1 8.2 TERMINATION SURVEY PROCEDURES 8-1 8.2.1 Indoor Survey 8-1 8.2.2 Outdoor Survey Procedures 8-6 8.2.3 Instrumentation and Methods for Contaminated Surface Surveys 8-7 8.2.3.1 Instrument Selection 8-7 8.2.4 Documentation 8-12 9

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J FIGURES Figure No.

Title Page 1-1 Location Map - San Francisco Bay Area 1-2 1-2 Location Map - City of Berkeley 1-3 1-3 Location Map - University of California, Berkeley Campus 1-4 1-4 Section Through Etcheverry Hall 1-5 1-5 Vertical Section - Berkeley Research Reactor 1-8 1-6 Plan View - Berkeley Research Reactor 1-9 1-7 Core Grid Array 1-11 1-8 University Organization for Decommissioning 1-44 s

3-1 Plan View of Etcheverry Hall Showing DECON J

Area 3-2 3-2 Localized Contamination Control at Reactor Pool 3-9 3-3 Exhaust and Supply Systen for Room 1140 3-10 3-4 8erkeley Rescarch Reactor Decommissioning Schedule 3-13 8-1 Measurements Made in a Typical Survey Block 8-3 8-2 Example of Maximum Observed Beta-Gamma Dose Rate and Direct Alpha Beta-Gamma Points in a Survey Unit 8-4 l

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O TABLES Table No.

Title Page 1-1 Principal Characteristics 1-1G 1-2 Estimated Personnel Dose Equivalent 1 - l'. '

1-3 BRR Operation History - Energy Production in Megawatt-Day (mwd) 1-27 1-4 Berkeley Research Reactvr - Approximate Neutron Flux at 1 MW 1-29 1-5 Material Compositions and Densities Used in Neutron Activation Calculation 1-32 1-6 Significant Neutron-Activation Gamma Emitting Radienuclides Six Months After Shutdown 1-33 1-7 Surface Activation Strength and Surface Dose Rate of Core Shroud as a function of Decay Time After Irradiation 1-35 1-8 Reactor Position at Exposure Room Surface Activation Strengths and Surface Dose Rates of Aluminum Tank Liner and a Typical Concrete Shield Wall 1-37 1-9 Reactor Position at Tank Center Surface Activation Strengths and Surface Dose Rates of Tank Side Walls 1-37 1-10 Reactor Position at Pool Center Surface Activation Strengths and Surface Dose Rates of Reactor Tank Floor 1-38 1-11 Reactor Posi' ion at Thermal Column Surface Activation Strengths and Surface Dose Rates of Carbon Steel Shield 1-39 1-12 Reactor Position at Thermal Column Surface Activation Strengths and Surface Dose Rates of Tank Liner and Concrete Shield Wall 1-39 1-13 Significant Gamma Emitters in Neutron-Activated Core Support Structure at Six Months Ater Shutdown 1-40 xii

%J TABLES (Cont)

Table No.

Title Page 1-14 Significant Gamma Emitters in Neutron-Activated Concrete in Exposure Room at Six Months Af t.er Shutdown 1-40 1-15 Significant Gamma Emitters in Neutron-Activated Concrete Floor in Reactor Tank at Six Months After Shutdown 1-41 2-1 Administrative Guidelines for Radiation Whole-Body Doses During Decommissioning 2-5 2-2 Regulatory Limits for Radiation Doses During Decommissioning for a Calendar Quarter (mrem) 2-5 2-3 Radiation Survey and Monitoring Instrumentation and Equipment Available for Decommissioning Activities 2 '. i 2-4 Mitigation and Monitoring of Hazards During Decommissioning 2-21 2-5 Estimated Ccst for Decommissioning the Berkeley Re: search Reactor 2-24 3-1 Reactor Components With Induced Activity -

Group 1 3-4 3-2 Miscellaneous Components With Potential Surface Contamination - Group 2 3-5

'3-3 Reactor Tank Structure With Induced Activity -

Group 3 3-6 3-4 Contaminated Tools and Equipment - Group 4 3-6 3-5 Decommissioning Milestones 3-15 8-1 Alpha, Beta-Gamma, and External Gamma Radiation '.evels in Room 1140 Etcheverry Hall, including Floor and Lower Wall Surfaces 8-5 8-2 Isotopes Potentially Present at the Berkeley Research Reactor and Their Principal Decay Characteristics 8-8 l

8-3 Typical Minimum Oetection Capabilities for Various Survey Instruments 8-10 xiii

ACRONYMS ACGIH American Conference of Governmental Industrial Hygienists ALARA As Low as Reasonably Achievable ANSI Americal National Standards Institute BRR Berkeley Research Reactor CCP Contamination Control Point CFR Code of Federal Regulations CIH Certified Industrial Hygienist CSCA Controlled Surface Contamination Area DECON Decommissioning DOE Department of Energy OP Decommissioning Plan 00T Department of Transportation EFPH Effective Full-Power Hours Office of EH&S Office of Environmental Health and Safety EPA Environmental Protection Agency HEPA High Efficiency Particulate Arrestor ICRP International Commission on Radiological Protection INP0 Institute for Nuclear Power Operations ISHP Industrial Safety and Hygiene Program LLD Lower Limit of Detection LSA low Specific Activity 0

xiv

ACRONYMS (Cont)

MPC Maximum Permissible Concentrations MSHA Mine Safety & Health Administration NAC Neutron Activatior. Code NCRP National Committee on Radiation Protection and Measurement NIOSH National Institute for Occupational Safety and Health NRC Nuclear Regulatory Commission ORPPS Occupational and Radiation Protection Programs Ca10SHA California /0ccupational Safety and Health Administration RAM Radioactive Material RC&SO Radiological Control & Safety Officer t

RHP Reactor Health Physicist RHWP Radiation / Hazardous Work Permit RPP Radiation Protection Program RSM Radioactive Shipment Manifest TLV Threshold Limit Values TRIGA Training Research Isotope Production, General Atomic Company l

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CHAPTER 1 Q

BACKGROUND AND MANAGEMENT PLAN

1.0 INTRODUCTION

The University of California Regents plan to cease operation of, and de-commission, the TRIGA Mark III Berkeley Research Reactor (BRR) located on the Berkeley campus in Etcheverry Hall, Room 1140.

The facility in which the reactor is located is to be released by the NRC for unre-stricted use. Use of radioactive materials in this decommissioned laboratory is to be under a State of California Radioactive Material License.

This decommissioning plan (DP) is directed towards removal of radioactive materialc down to levels of release for unrestricted use as defined in Regulatory Guide 1.86 Part C, Paragraph 2c, "Removal of Radioactive Com-ponents and Dismantling," and does not cover any activities relating to subsequent reuse. The University may elect to exclude particular equip-ment and facilities, such as the chemical hood in Room 1140, from decon-tamination to the levels for unrestricted use, with the provision that such equipment and facilities be released by the NRC to be under the State of California Radioactive Materials License. This OP is formatted as suggested by the U.S. NRC Standardization and Special Projects Branch paper entitled "Guidance and Discussion of Requirements for an Applica-

^

tion to Terminate a Non-Power Reactor Facility Operating License,"

h Revision 1, dated September 15, 1984.

1.1

SUMMARY

DESCRIPTION This section provides a brief description of the reactor facility, the history of the operating license, the decommissioning approach, costs and schedule summaries, QA requirements, contractor participation, and estimated man-rem exposures.

1.1.1 Reactor Facility Description The BRR is located on the Berkeley campus of the University of Califurnia in the City of Berkeley, Alameda County, California.

Figure 1-1 provides a general location map identifying the location of the City of Berkeley in the San Francisco Bay Area.

Figure 1-2 locates the campus within the City of Berkeley, and Figure 1-3 locates Etcheverry Hall on the campus.

The location of Room 1140 within Etcheverry Hall is shown on Figure 1-4.

The reactor site is approximately 4 km (2-1/2 mi) east of the San Francisco Bay, Interstate Highway 80, and the Southern Pacific Railroad.

The nearest n'ajor airport, Metropolitan Oakland International Airport, is approximately 7 km (11 mi) south of the site. The site is on the basement floor of Etcheverry Hall which is on a hillside approximately 97.5 m (320 ft) above sea level and sloping in a westerly direction p

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O EXHAUST FRou RMW 1140 l

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Figure 14 Section through Etcheverry Hall 1-5

The facilities that are under the NRC licensing jurisdiction, for the purposes of this decommissioning plan, include:

o Reactor Room 1140 in its entirety o

The room ventilation system including the emergency ventila-tion system o

The personnel airlock, room entrance cna emergency exit o

The truck access door o

The cooling water system o

Room 1110 B (previous pneumatic rabbit terminu;)

o Personnel decontamination room and adjacent office Room 1140 is a high bay room serviced by a 4.5-tonne (5-ton) bridge crane, containing the reactor et the east end and various research and teaching facilities distributet throughout the remainder of the room.

Additional rooms are located on the north side and on the west side.

The room is below grade, and the east wall is a retaining wall.

There is a patio on the roof above Room 1140.

Controlled truck / equipment access to Room 1140 is provided by a 3.6-m (12-ft)-wide truck access aisle leading through doors to a loading dock on the exterior of the west side of the building. Controlled personnel access is provided by an aisle equipped with an airlock and opening into a hallway area. An emergency personnel exit is provided in the northeast corner of Room 1140 leading to the patio area above.

This door is locked to prevent entry into Room 1140. All doors are controlled, and entry is under the direct supervision of authorized personnel.

The western half of the room contains the following equipment:

o Chemical research facility with fume hoods o

Computer room, enclosed o

Nuclear engineering class laboratory and lecture area o

Microwave experiment o

Radiation effects equipment and class area o

Fusion experiments o

Gamma-ray counting laboratory These facilities and equipment are currently operating under the NRC license and a State of California Radioactive Material License.

The University intends to continue their operation.

1-6

{Q s

This plan shows how the room will be partioned to isolate areas from the portion where decommissioning activities are being performed.

Room 1140 contains an independent ventilation exhaust system equipped with high efficiency particulate (HEPA) filters.

There is also an emergency ventilation system.

Room 1140 and the rest of Etcheverry Hall is of reinforced concrete construction.

1.1.2 Reactor Description The BRR is a standard TRIGA Mark III manufactured by GA Technologies of San Diego, California.

It is a heterogeneous, pool-type, 1-MW reactor employing solid uranium-zirconium hydride fuel elements of less than 20%

U-235 onrichtrant.

Fuel cladding is stainless steel.

The reactor core is suspended from a movable bridge in a large open tank of demineralized water which provides cooling, shielding, and moderation.

Control of the reactor is ac.11eved by insertion and withdrwal of 4 neutror.-absorbing control rods.

Transient pulses are achieved by the pneumatic ejection of a single trsnsient rod.

The core may be moved laterally in the tank to accommodate experiments at the thermal :olumn or beam ports at the west end of the tank, the exposure room a: the east end, or any position in between. An overall view of the reactor is shown on Figures 1-5 and 1-6.

Principal charac-terisi.ics of the TRIGA Mark III are summarized on Table 1-1.

The reactor core currently consists of a lattice of 104 standard cylin-drical U-ZrH, fuel elements, one instrumented fuel element,15 graphite dummy elements, three fuel followed control rods, and one air followed transient control rod.

The active (or fueled) region of the reactor core forms a right circular cylinder 53.3 cm (-21.0 in.) in diameter and 38.1 cm (15 in.) high and contains -3.9 kg (8.6 lb.) of U-235. Water coolant occupies approximately one third of the core volume.

Top and bottom axial reflection is provided by 8.9-cm (-3.5-in.)-long graphite plugs incorporated into the individual fuel elements.

Figure 1 7 illustrates the current core grid array.

The Gel elements are positioned laterally at the top and bottom by two tyre 6061-T6 aluminum grid plates 1.6 cm (bs in.) thick and 1.9 cm (%

in ) thick respectively.

The lower grid plate tupports the weight of the fuel elements. Both grid plates are supported by pads welded to the core support assembly, which, in turn, is suspended from the movable bridge over the top of the reactor pool. The reactor can be pulsed or operated at full steady state power at any position in the pool; how-ever, it has principally been operated in the three positions marked on Figure 1-5.

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O 1-9

Table 1-1 Principal Characteristics Characteristic Description Reactor type TRIGA Mark III Maximum steady-state power level 1 MW Peak pulse power 2800 MW, licensed 1300 MW, actual Maximum energy release, pulse 15 MW-sec Fuel element design Fuel-moderator material U-ZrH 1,7 Uranium content 8.5 wt % (0.195 Kg/ element)

Uranium enrichment

<20% U-235 Shape Cylindrical Length of fuel 38 cm (15 in.) overall Diameter of fuel 3.63 cm (1.43 in.) outside Cladding material Type 304 stainless steel Cladding thickness 0.051 cm (0.020 in.)

Number of fuel elements 105 Excess reactivity, maximum 4.9%A k/k (cold, clean)

Control Rod Descriptions Safety Boron / graphite - fuel followed Regulating Boron / graphite - fuel followed Shim Boron / graphite - fuel followed Transient Boron / graphite - air followed Total reactivity worth of rods 7.6%A k/k Reactor cooling Natural convection of pool water O

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The operation of the BRR is monitored by instrumenta' ion channels that measure fuel element temperature and reactor power.

n.crmocouples in an instrumented fuel assembly provide information on fuel material tempera-ture during both steady-state and pulse operation. This signal is dis-played at the control console.

The reactor tank is a welded aluminum vessel located at floor level and surrounded by a reinforced-concrete shield structure. The tank is in the form of an elongated cylinder 0.63 cm (0.25 in.) thick, 7.6 m (25 ft) long, 3 m (10 ft) wide, and 6.1 m (20 ft) deep, with a capacity of approximately 135,000 1 (35,700 gal) of light water.

The outside of the tank is coated with a bituminous coating for corrosion protection.

The tank assembly rests on a 1.3 m (4 ft, 4 in.) thick concrete slab.

The reactor core is positioned near the bottom of the tank under 4.6 m (15 ft) of demineralized water, which serves as a radiation shield, neutron moderator, and reactor coolant. A 12.7-cm (5-in.)-thick steel plate is incorporated in the bridge above the tank to provide additional vertical shielding.

Natural thermal convection of the water adequately disperses the heat generated in the core during both steady-state and pulse operations.

When necessary, the coolant water may be pumped through an external heat exchanger system that ultimately disposes of the heat to the atmosphere.

Suspended fuel storage racks in the reactor tank are available for routine storage of fuel elements and/or reactor components.

In addition one 1.5-m (5-f t)-diameter, 4.6-m (15-f t)-deep storage well outside of the reac'.or tank provides isolated storage for fuel elements or radiation sources.

The tank has a capacity of 8,300 1 (2,200 gal).

All storage racks have been designed to be criticality safe for TRIGA fuel elements immersed in water.

The core is shielded horizontally by approximately 2.1 m (7 ft) of ordin-ary concrete and 1.2 m (4 ft) of water. Vertical shielding is provided by 4.6 m (15 ft) of pool water above the core.

At the beam port end of the facility in the radial direction nominal shielding is 0.9 m (3 ft) of water and 2.9 m (9 ft, 8 in.) of concrete.

At the exposure room end of the facility, the horizontal shielding is 2.2 m (7 ft, 4 in.) of heavy concrete with a nominal density of 3

3.5 g/cm. All other concrete in the shield is ordinary concrete with a nominal density of 2.3 g/cm'.

Shielding arrangements are shown on Figures 1-5 and 1-6.

At the west end of the pool structure, an aluminum-enclosed graphite thermal column extends from the eastern most core position through the concrete shield structure, as shown in Figures 1-5 and 1-6.

Horizontal access and shielding for this thermal column are provided by a track-mounted, motor-driven, high-density-concrete door. Also at the thermal-column end of the structure are four 15.2-cm (6-in.)-diameter radial beam ports, which extend from the core periphery through the water and concrete to the outer face of the shield structure.

Two 20.3-cm (8-in.). diameter straight-through beam ports intersect in the thermal column immediately adjacent to the core.

Another radial port can be 1-12

i I

i

(,

provided by removing the graphite and lead stringers in line with the access port through the thermal column door.

Two other 15.2-cm (6-in.)-diaineter beam ports intersect the boral, polyethylene, and graphite-lined cavity (hohlraum) in the outer portion of the horizontal thermal column; the hohlraum is about 91 by 91 by 102 cm (36 by 36 by 40 in.).

A region of graphite, about 81 by 81 cm (32 by 32 in.),

located above the hohlraum, provides a vertical thermal column; a large, stepped concrete shield plug provides vertical shielding. Any or all of the facilities at the thermal column end of the pool may be used simultaneously.

At the opposite end of the structure is a borated-concrete-lined expo-sure rcom about 2.7 m (9 ft) high, 3 m (10 ft) wide, and 3.7 m (12 ft) long in which large engineering and biological experiments may be irradi-ated. A boron-lined segment of the aluminum pool liner projects into the exposure room; the size of this segment permits maximum transmission of radiation into the room and provides 180-degree access to the core perimeter for experiments.

The north side wall of the room is high-density concrete; the other walls are normal-density concrete.

The ceiling, walls, and floor of the exposure room are lined with 30.5 cm (12 in.) of borated concrete to minimize neutron-induced activation.

Access to the room is by means of a track-mounted, motor-driven, high-density-concrete plug door.

Electrical power leads and other services used with experimental appara-tus are routed through conduits in the shield.

Cooling water pipes em-O bedded in the concrete remove the heat deposited by absorbed radiation and prevent excessive thermal Stresses in the concrete.

1.1.3 Operating License History The BRR was installed in Etcheverry Hall and licensed for operation at 1 MW on August 10, 1966 (License No. R-101, docket no. 50-224).

Subsequent license amendments are as follows:

Amendment No. 2 September 28, 1979 Amendment No. 3 November 19, 1980 The cumuistive energy production from August 1966 to March 1987 was 287.4 MW-days (89.2% at pool center, 8.4% at thermal column, 2.4% at exposure room). The total cumulative energy production from August 1966 through final shutdown was approximately 293 MW-days.

1.1.4 The Decommissioning Approach The University of California Regents plan to decommission the BRR located in the eastern portion of Room 1140 of Etcheverry Hall and release the facility for unrestricted use as defined by Regulatory Guide 1.86.

The selected decommissioning alternative is "DECON."

The University plans to remove and ship the fuel off-site under its cur-rent operating license.

1-13

The University ceased use of the reactor as a teaching facility in December 1987. However, due to continued academic activities in the laboratory, which also occupies Room 1140, the DP is structured to achieve minimum interruption to these activities by concentrating the bulk of the decommissioning activities during one of the academic recesses. The scheduled dates are May 15 through September 1, 1988.

No reuse of the reactor and its biological shielding is contemplated.

After a 90-day cooldown period, fuel and neutron sources will be removed from the reactor.

Fuel will be shipped to the U.S. Department of Energy (00E).

The Am-Be neutron source and six Pu-Be neutron sources will be retained by the University.

The original Po-8e neutron source will be shipped off site.

The reactor pool water heat exchanger and associated pumps and valves may be retained by the University or donated to the University of California, Irvine. All other components will be dis-mantled and disposed of in accordance with (as low as reasonably achiev-able) ALARA principles and the OP as described herein.

The analytical results of the estimates of low-level waste quantities and worker exposure given in this DP will be confirmed by field measure-ment.

In general, decontamination will proceed from the areas exhibit-ing maximum radiation levels to those areas with minimum radiation levels.

1.1.5 Decommissioning Alternatives As stated in section 1.1.4, the University of California's intent that the reactor site and the remainder of the reactor facility be available for unrestricted use has led to the decis, ion to select the immediate dismantling and decontamination (DECON) decommissioning alternative.

Other alternatives such as SAFSTOR and ENTOMB would not result in imme-diate release for unrestricted use of the site.

In addition the Uni-versity wants the option of unrestricted use to allow for future build-ing abwe this site. On this basis, these latter two alternatives were considered inappropriate to the University's needs.

1.1.6 Availability of Funds Funds for the decommissioning of the BRR are being allocated by the Uni-versity. Funds to start the decommissioning project were appropriated by the Regents of the University of California at their January 1987 meeting. These funds will cover the preparation of the DP, defueling, and some construction.

Additional University funds will be allocated to complete reactor dismantlement.

1.1.7 Major Tasks and Schedules After removal of fuel, nine major tasks are planned for the DECON pro-gram.

To lessen interference with the University's academic activities, the plan proposes to initiate and complete decommissioning in a three-month period starting mid-May 1988 and ending by mid-September 1988.

1-14

C)

Tasks and related planned milestone dates are summarized below:

%)

Milestone Task No.

Description Completion Date 1

Contractor Move-in May 20, 1988 2

Initial Radiation Survey June 3, 1988 3

Installation of Confinement Barriers June 3, 1988 4

Removal of Reactor Components and June 17, 1988 Pool liner 5

Removal of Material with Potential July 22, 1988 Surface Contamination and Other Activated Materials 6

Cleanup and Removal of Tools and August 19, 1988 Equipment 7

Packaging and Shipment of Radioactive August 26, 1988 Waste 8

Perform Termination Radiation Survey September 30, 1988 9

Demolition of Non-Radioactive Portion December 30, 1988 of Reactor Installation Because of the short duration of the summer recess (75 days), the termina-tion radiation survey (Task 8) will not start until after classes resume.

The milestone completion date for Task 9 planned for December 30, 1988, depends on two preconditions:

o Duration of the NRC final inspection prior to granting approval for termination o

The University's need for the site and integration into the overall future facilities plan Task 9 could be postponed until the summer recess of 1989 with no impact on this DP.

1.1.8 Quality Assurance Plan A quality assurance plan will be initiated to provide for:

(a) confor-mance to the DP procedures, and (b) procurement of equipment and ser-Vices that affect public and operational health and safety.

O 1-15

The Quality Assurance Plan will include the following items:

o Review of operating personnel health and safety training procedures calibration and maintenance practice @srgs and instrumentation Review of radiation monitoring proce o

o Review and monitoring of the decommissioning procedures for adequacy in regard to public health and safety, security, maintenance of ALARA conditions, choice of methods and equip-ment, and conformance to. Department of Transportation (00T) regulations for low level waste shipment o

Revies and comment regarding proposed changes or deviations to the DP o

Review of procurement documents for equipment and/or services that affect public health and safety o

Monitoring of document control system with regard to work instructions and procedures, drawing and information manage-ment, radiation survey results, and field changes o

Review of documents released for NRC review / approval 1.1.9 Executive Enginecr! Contractor Participation The University plans to have the majority of the work as listed in Section 1.1.7 performed by one or more outside contractors.

Some sup-plemental services / facilities, such as health physics monitoring and radiation surveys, will be provided by the University.

The Executive Engineer /centractor(s) participation is planned to include:

o Provision of organization and management to implement the DP o

Preparation of work procedures, QA procedures, and training plans o

Training of appropriate personnel as required o

Preparation of progress reports, cost and schedule reports, deviation and/or field change reports, and radiation survey reports o

Supervision of day-to-day decommissioning activities including direction of craft supervisors o

Procurement of services and equipment o

Administration of subcontracts and subcontractors 1-16

i f

o Provision of crafts and labor to perform all dismantling, demolition, shipment, and disposal of radioactive and non-s radioactive materials o

Assistance and cooperation with the University staff in any supplementary related activities that the University may elect to conduct 1.1.10 Termination Radiation Survey Plan After the completion of Tasks 1 through 7, as outlined in Section 1.1.7, a final comprehensive termination survey will be conducted by the Execu-tive Engineer to establish that contamination is within the limits specified in Table 1, "Acceptable Surface Contamination Levels," of Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors." The survey will cover that portion of the facility, with the reactor structure and its supporting equipment, that are specified to be within the boundary of the NRC licensed facility.

Details of the plan are provided in Chapter 8.

A final report will be prepared and submitted to the NRC at least 30 days prior to the scheduled date for demolition of the non-activated portions of the reactor shield structure. The report will include:

o Identification of the decommissioned premises o

Demonstration that reasonable effort has been made to reduce residual contamination to as low as practicable levels o

A description of the scope of the termination survey and the general procedures followed o

A statement of the findings in units specified in Table 1 of Regulatory Guide 1.86 1.1.11 Estimated Personnel Dose Equivalent During cecommissioning, one of the major goals of the ALARA program is to maintain the collective dose equivalent (man-rem) for all persons involved in the BRR decommissioning to ALARA.

The collective dose equivalent is a summation of the following variables:

o Number of tasks performed o

Average dose rate at the location where the task is performed o

Task duration o

Crew size required for task performance o

Measures taken to implement ALARA practices 1-17

Table 1-2, "Estimated Personnel Dose Equivalent," provides the estimated collective dose equivalent based on the tasks defined in section 1.1.7, the estimated man-hours required to perform each task, and the average dose rate associated with each task, taking into account ALARA practices.

The approximate exposure for DECON is 7.7 man-rem based on preliminary calculations using available data. The value cannot be accurately deter-mined until samples of activated materials in the walls of the pool and exposure room can be taken and analyzed.

Due to the uncertainties it is estimated that the man-rem dose will be within the range of 1-15 man-rem.

1.2 OPERATING HISTORY The BRP received a notice of issuance of a facility license, No. R-101, authorizing operation of the BRR on August 10, 1966.

The reactor was in continuous use until December 1987 when it was shut down in anticipation of the dismantlement.

1.2.1 Reactor Energy Production The reactor energy production history as a function of time and location in the reactor pool is shown on Table 1-3.

1.2.2 Radioactivity Release / Discharge The reactor experienced a leaking fuel assembly during 1985.

Three sus-pect fuel assemblies were removed and replaced with fresh new assem-blies.

The suspect leaking assemblies were stored in the fuel rack within the reactor pool. After removal from the reactor, no leakage was detected.

1.2.3 Experiments and Related Unusual Events Experimental work has been performed using laboratory hoods in the west end of Room 1140.

This work has been performed under the State of California Radioactive Material License.

Measurements of filter media with a windowless gas flow GM indicates presence of radioactive material slightly above background.

1.3 CURRENT RADIOLOGICAL STATUS OF FACILITY 1.3.1 General The current radiological status of the facility has been estimated by performing neutron-induced activation calculations for the core support, reactor tank (pool), exposure room, and their associated structural materials, and calculating the resulting radiation levels in the sur-rounding area.

It was assumed that the irradiated fuel elements and fuel-followed control rods have been removed under the existing operat-ing license.

O 1-18

i

.l Tab e 1-2 j

Estimated Personnel Dose Equivalent I

MANPOWER Duration Supervision Craft Laborer H.P.

Total Exposure

Dose Rate Total b

{

Task Number / Activity (Days) Men Days Men Days Men Days Men Days MD Hour Man-Hours mR/ Hour Man-REMS

)

1.

Contractor Move-In I

o Set uo tratier 1

1 1

2 2

5 40 30 0.2 6 x 10-8

)

o Install / connect utilities 2

1 2

2 1

1 1

5 40 30 0.2 6 x 10-'

i i

o Clear and secure work areas 1

1 1

2 1

4 32 24 0.2 4.8 x 10-:

3 insta11 barriers o Complete health and safety training, medical exam, respirator, etc.

3 1

3 3

3 8

3 11,2 1

37.5 300 0

BKG

~*

j Subtotal Task 1 7

51.5 412 84 0.017 i

i I

i i

a - Exposure based on 6-hour day D - Based on UCS survey data and NUREG/CR - 1756 1

4

Es2C M 3633722 I.

m e

Table 1-2 (Cont)

Estimated Personnel Dose Eculvalent MANPOWER a

b Duration Supervision Craft Laborer H.P.

Total Exposure Dose Rate Total Task Number / Activity (Days) i4en Days Men Days Men Days Men Days M0 Hour Man-Hours mR/ Hour Man-REMS 2.

Initial Radiation Survey A.

Gescter Accm 1140 o Grcss gamma survey 1

2 1

2 16 12 0.2 2.4 x 10-2 o Cetailed surface s.1;e test:

2 2

2 4

32 24 0.2 4.8 x 10-8 cetailed beta. gamma o Airborne and pool water 1

2 1

2 16 12 0.2 2.4 x 10-8 o Frepare radiation zone map 1

1(CAF 0

1 8

0 oper) 8.

Pesctor & Shield

/2 1, 2 4

3 0.2 6 x 10-'

/2 1

1 a Gross gama survey 1

o Core samples (assume 10) 1 1

1 1

1 3

24 18 10 0.18 (10-20)

/2 1, 2 4

3 0.2 6 x 10-'

o Interior pool 1/2 1

1 C.

Etcheverry Hall (Interior)

/2

1. 2 4

3 0.2 6 x 10" o Gross gama (5 areas)

In 1

1 o Air samp'e (5 areas) in 1

1,2 1; 2 4

3 0.2 6 x 10-'

O.

Etcheverry Hall (Exterior) o Gross gama 1, 2 1

12 2

4 3

BKG 6x 10-5 (0.02) o Air sa. pie a, 2 1

12

1. 2 4

3 BKO 6 x 10-5 (0.02)

E.

Esaluation of Resu'.ts o Plotting 1

1 1

8 0

o Recording / reporting 1

2 2

16 0

Subt<

18 144 84 0.192

- Ecosure based on D - Eased on UCS surv'

.756. BKG fcr background 353633723

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)

(

Nd Table 1-2 (Cont)

Estimated Personnel Dose Equivalent MANP(b1R a

b Duration Supervision Craft laborer H.P.

Total Exposure Dose Rate Total Task Number / Activity (Days) iten Days Men Days Men Days Men Days M0 Hour Man-Hours mR/ Hour Man-REMS 3.

Installation of Confinement Barriers o Erect barriers Raom 1140 and 4

1 4

4 4

2 4

28 224 168 0.2 0.0336 over reactor pit o Inspect / test existing HEPA n

2 16 12 0.1 1.2 x 10'8 ventilation in 1

in 2

1/2 1

2 o Install / test temporary venti-lation system 2

1 2

2 2

1 2n loin 84 63 0.2 0.0126 o Test air-monitoring system in 1

1 1

In 1

In 2

16 12 0.2 2.4 x 10

n 1

2n 1

In 2 In 2

16 12 0.2 2.4 x 10

o Inspect / test crane and hoist i

o Install laydown areas. step-offs, signs plastic sheet.

7 etc.

I 1

1 2

1 8

1 In 1 111n 92 69 0.2 0.0138 fu o Remove temporary shield mall n 4 in 3

24 18 0.1 1.8 x 10-s at truck door in 1

In 1

i Subtotal Task 3 9

59 472 354 0.0678 l

8 - Esposure based on 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months day D - Besed on UCB sevey data ar.d NUREG/CR - 1756

E920853633724 l

Table 1-2 (Cont)

Estimated Personnel Dose Equivalent MNPOWER a

b Duration Supervision Craft Laborer H.P.

Total Exposure Dose Rate Total Task Number / Activity (Days)

Men Days Men Days Men Days Men Days MD Hour Man-Hours mR/ Hour Man-REMS 4.

Removal of Reactor Components and Pool Liner o Octary specimen rack 1

4 1

/2 1, 2 1

12

,in

,1/2 1 /2 61/2 52 39 4

0.156 o Con +.rol rod guides, j

chambers, detecto-s 2

J 4

in o Graphite dumy elements 1

1 1

1/2 v2

'in in

'1 /2 n

41n 36 27 4

0.208 i

o Rabbit 1

1 4

in o Grid plates 4

1 1 /2

>1f2 1

2/2 In

'1/2 In

' Sin 44 33 4

0.132 o Suppcrts assembly, shroud 1

j 4

1, 2 i

o Cut, = ash down, and package part-2 1

2 3

2 8

2 1

2 26 208 156 3.5 0.546

$)

o Pool lirer; ash da.n in 1

In 0

4 in 1

in 3

24 18 2

0.036 ro o Treat ster (See Section 3.3.3) in 1

in 2

in 1

In 1

In 22n 20 15 2

0.030 o Co-e samples; floor 1

1 1

2 1

1 1

5 40 30 2

0.060 o Pool liner; remove cut and 2

1 2

2 2

8 2

1 2

24 192 144 2

0.288 package o Metal emoedments; remove, cut, 3

1 3

2 3

4 3

1 3

24 192 144 2

0.288 packa;e Subtotal Task 4 101 2 101 808 606 1.644 a - E=posare based on 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months day b - Essed en UCE survey data a M NUFEG/CR - 1756

E' 08725

~~-.

-. ~.

~.

.. - - - ~.

~-

n O

. Table 1-2 b wt)

Estimated Personnel Dose Equivalent I

)

MMPOWR s

h Duration. Supervision G aft taborer H.P.

Total Exposure Dose Rate

. Tota!

Task Number /Activitf (Days) Men Days Men Days Men Days Men Days, le Hour Man-Hours mR/ Hour Man-REMS 5.

Removal of Material with Potential Surface Contamination at:d Other Activated Materials o Esposure roos conc nte

- Cetling/=sils/ floor

- Using po.er hammer f 23 n 23 2871/2 2300 1725 2

3.45 1

23 3

23 8

23 2

remotely o Pool ccncrete: malls, floor o Rebar n

1 3tn 9

21 2

0.042 l

o Thermal coluwt 1

1 1

2 1

i t

o Contaminated carbon steel shield incl. above i

n 1

51/2 44 33 2

0.066 o Washdown 1

1 1

4 1

2 o Package materials incl. above w

i Suetotal Task 5 25 2961n 2372 1779 3.558 L

5 6

[

L

.t

+

l

- ENDos Fe based on 6-hour day b - Based on UC3 s,rvey cata and NUREG/CR - 1756 (E920853633?26 L

Table 1-2 (Cont)

Estimated Personnel Dose Equiv-lent MANPOWER a

D Duration Sup geision Craft laborer H.P.

Total Exposure Dose Rate Tota'.

Task Number / Activity (Days)

Men I)ays Men Days Men Rs Men Days MD Hour Man-Hours mR/ Hour Man-REMS 6.

Cleanup and Removal of Tools and Equipment

' acuum inside pool and 1

1 1

4 1

1 in 51n a4 33 1

0.033 o

espesure room o femove temporary reactor i

n 3

12 1

2n 4

32 24 1

0.024 n

1 1

2 3

i enclosure o OECC4-stored waste packages 5

1 5

4 5

1 5

30 240 180 2

0.360 (est. 33 bones concrete (3200 cu ft); 5 bones met-(90 cu ft) o Clean ductwork; remove /

replace HEPA filters 1

1 1

4 1

1 1

7 56 42 0.I' 4.2 x 10-s o DEC0h tools in 1

in 4

2n 1

In 3

24 18 1

0.018 o CECON storage well 3

1 3

3 3

1 1

16 128 96 1

0.096 o Dewineralize water 2n 1

In 2

in 1

in 2

16 12 1

0.012 o Dismantle aater cleanu; ys m 3

1 3

2 3

4 3

1 1

22 176 132 1

0.132 o Femove confinement bar

-s 2

1 2

2 2

4 2

1 2n 14 n 116 87 2

0.087 o Package all solid writes 3

1 3

2 3

6 3

1 lin 28:n 228 171 2

0.324 e Lead out all solid -aste to truck 1 1

1 8

1 1

1 10 80 60 2

0.120 Sabtotal Task 6 20; 2 1421 2 1140 855 1.228 8 - Exposure based on 6-h3ur day l-EasedenUCBsurveydataandNUREG/CR-1756 Mezzarire area 5362c727

l 1

j i

Table 1-2 (Cont)-

Estimated Personnel Dose Equivalent t

)

MMePOWER a

b Duration Supervision Craft Laborer H.P.

Total Exposure Dose Rate Total Task Number / Activity (Days) Men Days Men Days Men Days Men Days 16 Hour Man-Hours aft /Nour. Non-REMS 7.

Packaging and Shipment of Radioactive Weste i

o Close and seal containers lin 1

lin 1

11n 6 lin 12 96 72 2

0.144 i

o Radiation survey all containers 2

1 2

4 32 24 2

0.048 j

o Clean and labei 3

1 3

4 3

15 120 90 2

0.180 3

o Load / secure 2

1 2

4 2

12 72 2

0.144 o Final ved. survey packages, 22n 1

21n 1

21/2 5

40 30 2

0.060 vehicles, etc.

I r

Assume: 40 containers 320 cu ft material

't i

j Subtotal Task 7 11 48 384 288 0.576 a

a N

ui i

T h

I I

I 4

i 1

4 5

a

- Exposuee based on 6-hour day b - Based on UCB survey data and NUREC/CR - 1756 j

F.E920853638728 I

l l

i I.....

Table 1-2 (Cont)

Estimated Personnel Dose [quivalent MANPOWE R a

Duration Supervision Craft Laborer H.P.

Total Exposure Oose Rate Total Task Number / Activity (Days)

Men Days Men Days Men Days Men Days MD Hour Man-Hours mR/14our Man REMS 8.

Perform Te m ination Radiation Survey o Survey Room 1140 4

2-4 8

64 64 BKG (0.02) 1.28 x 10-'

o Survey adjacent rooms 4

2 4

8 64 64 BKG (0.02) 1.28 x 10**

o Esterior survey 3

2 3

6 48 48 BKG (0.02) 9.6 X 10

o Soil / ater sampling 3

1 3

3 24 24 BKG (0.02) 4.8 x 10

o Write final report 10 2

10 20 80 0

Subtotal Task S 24 45 350 290 0.004 7

UC8 Decommissioning Staff o Proj. e9gineer & QA CM neer 20 40 320 240 0.5 0.12 b

i o Peactor HP & ceputy project 20 40 320 240 1.0 0.24 D

engineer Subtotal UCB Staff 60 640 480 0.36 GFsMO TOTAL 841.5 6732 4810 7.66 a _ g,s,3 gn,JJB survey esta and NUR[G/CR - 1756 D - Based on sporon. 2C% of the total working days P

'3633729

Table 13 BRR Operation History Energy Production in Megawatt Day (mwd)

Year Pos.1 Pos.3 Pos.6 Total Exposure Pool Thermal Room Center Column 1966 0.003 1.269 0.436 1.709 1967 0.043 15.006 2.434 17.484 1968 0.310 28.617 0.351 29.278 1969 0.363 27.463 3.689 31.515 1970 0.976 28.539 0.642 30.157 1971 1.090 26.899 2.799 30.788 1972 0.526 18.217 2.841 21.554 1973 0.378 14.648 5.937 20.963 1974 0.393 20.237 1.174 21.804 1975 0.437 13.086 0.388 13.911 1976 0.149 7.294 0.285 7.729 1977 0.198 6.899 0.021 7.118 1978 0.256 7.938 0.418 8.611 1979 0.239 4.345 0.007 4.591 0

1980 0.279 5.352 0.995 6.625 1981 0.324 4.270 0.312 4.905 1982 0.252 4.297 0.361 4.909 1983 0.176 4.775 0.299 5.251 1984 0.062 5.613 0.097 5.772 1985 0.180 5.113 0.296 5.589 1986 0.174 5.876 0.320 6.369 1987 0.176 5.548 0.100 5.824 Total 7.0 261.3 24.2 292.5 Percent 2.4 89.3 8.3 100.0 0

1-27

The neutron-induced activation calculation for the core support struc-ture will be limited to those components located near the core:

the core shroud, and the top and bottom grid plates.

They are all fabri-cated of aluminum alloy and have a relatively low neutron activation.

However, because of continuous exposure to high neutron flux, the core shroud and both grid plates will be radioactive and require underwater storage and dismantling, prior to packaginq and shipment to a waste storage facility.

The pool structure materials consist of aluminum liner and a thick con-crcte biological shield surrounding the pool liner.

The beam ports, thermal column, and the carbon steel gamma shield are mostly embedded in the concrete shield structure; however, some activation of these com-ponents is expected and will require a separate radiation survey.

l As stated in the previous Section 1.1, the reactor has principally been operated in three positions for various lengths of time. Radionuclide concentration in the surrounding pool liner and concrete biological shield of each reactor position were evaluated as a function of effec-tive neutron flux, exposure time, and time after irradiation to estimate the dose rate within the reactor pool and the exposure room.

l From this information it is possible to:

l i

o Determine to what extent the reactor pool structural com-ponents (mainly consisting of the pool liner and concrete) will have to be removed in order to comply with the limit of 5 pR/h above background at one meter from the exposed surface and other limits imposed by Regulatory Guide 1.86 o

Specify the necessary safety measures and procedures for various dismantling, removal, decontamination, storage, and disposal operations so that exposure to personnel is main-tained ALARA 1.3.2 Neutron Flux The three reactor positions -- position 1, Exposure Room; position 3, Pool Center; position 6 Thermal Column -- provide unique neutron fluxes to the surrounding materials caused mainly by the difference in the thickness of water around the reactor and separation distance between the core and exposed structural material.

Steady-state neutron flux measurements at the power level of 1000 kW for different locations are tabulated in Table 1-4.

The thermal neutron fluxes were determined using bare and cadmium-covered gold foils.

The flux values for neutrons with energies greater than 3 MeV were deter-mined using the Fe-54(n,p) Mn-54 reactions.

Recently the neutron flux measurements were made by the reactor opera-tion staff in order to provide the data necessary to perform the pool liner and concrete wall and floor activations calculations.

1-28

i Table 14 Berkeley Research Reactor Approximate Neutron Flux at 1 MW a

Flux n/cm sec Location Thermal

= 3 MeV 28 3.3 x 10 t t

Central Thimble 3.6 x 10 (Maximum)

Lazy Susan 5.0 x 1012 1.5 x 101 Rabbit 1.3 x 1018 8.8 x 102!

Exposure Room 2.5 x 10' 6.4 x 10 '

2 (10 cm from pool wall, 80 cm from floor)

Thermal Irradiation Facility 1.2 x 10' (Hohlraum)

Distance from South Side Tank Wall l

Liner Along Core Mid-plane, cm (in.)

]

0 (0) 5.5 x 10' l

22.8 (9) 8.8 x 105 45.7 (18) 1.3 x 10' Distance from Vertical Core Center-l line at Tank Floor Liner, cm (in.)

j 0(directlybelow)

(0) 7.0 x 107 l

25.4 (10) 4.6 x 10' 50.8 (20) 1.6 x 10 7

l l

T T

I I

I t

l 1-29 l

1.3.3 Neutron Activation Analysis Neutron-induced activation of materials considered herein was calculated by the Neutron Activation Code (NAC) computer code, originally developed by the NASA Lewis Research Center. The code is available from the ORNL Radiation Shielding Informa' ion Center as RSIC Code Package CCC-164.

The NaC is designed to predict the neutron induced gamma-ray radioactiv-ity for a wide variety of composite materials.

The NAC compiles a prod-uct isotope inventory containing the isotope name, the disintegration rate, the gamma source strength, and the absorbed dose rate at one meter from an unshielded point source.

The induced activity was calculated for aluminum, concrete, and carbon steel as a function of irradiation and decay times.

For this study, each reactor position has its irradi-a t. ivn Ume and neutron flux. A decay time from shutdown to six months with monthly increments was assumed for each calculation.

1.3.3.1 Irradiation Times At the time the reactor was shutdown for aecommissioning, the BRR accu-mulated about 293 mwd of total energy production.

This amounts to a total of 7,020 effective full-power hours (EFPH) at a power level of 1,000 kW for the entire lifetime of the reactor.

The EFPH is divided into three reactor positions, i.e. 168, 6,270, and 586 hours0.00678 days 0.163 hours 9.689153e-4 weeks 2.22973e-4 months for Expo-sure Room, Pool Center, and Thermal Column respectively.

They consti-tute 2.4%, 89.3%, and 8.3%, respectively, of the total EFPH.

To sim-plify the activation calculation the reactor was assumed to have oper-ated continuously for the applicable EFPH just prior to shutdown.

Therefore, the induced activity level represents the maximum activity which is appreciably higher than the actual actisity based on the inter-mittent reactor operating history shown in Table 1-3.

1.3.3.2 Neutron Energy Groups The neutron flux in the four energy groups can be specified in the NAC code as listed below.

However, for the calculations, groups 2 and 3 are not used since the activation contributions are expected to be small compared to the other two groups.

Group 1 0.82 MeV

<E (fast)

Group 2 5.5 kev

< E < 0.82 MeV (intermediate)

Group 3 1.1 eV

< E < 5.5 kev (epithermal)

Group o E < 1.1 eV (thermal)

A' seen in Table 1-4, the pool liner and concrete walls and floor at the exposure room will be exposed to a relatively high fast neutron flux compared to thermal neutron flux due to the lack of moderating material.

Therefore, the fast neutron induced activation, in addition to the ther-mal neutron activation, was evaluated. On the other hand, the reactor at the pool center and the thermal column positions are surrounded by more than 0.9 to 1.2 m (3 to 4 ft) of the pool water and the fast neutron-induced activation contribution is expected to be negligibly small compared to that of thermal neutrons.

1-30

1.3.3.3 Material Compositions The three materials considered are Aluminum 6061-T6 alloy for the core support and the pool liner; ordinary concrete for the biological shield, walls, and floor slab; and A-36 carbon steel for the laminated steel shield above the tnermal column. Their material compositions, used in the activation calculations, are shown in Table 1-5.

Only those radio-nuclides whose half-lives and/or concentration at shutdown result in a significant contribution to the total activity after three months of decay will be considered. A rare earth element, europium, was included in the concrete composition based on a concentration of 0.1 pg of Eu per gram of concrete since definitive data on its pro'eable initial concen-tration is not available. After the reactor shutdown and during the actual decommissioning operation, it will be necessary to obtain and analyze samples from the concrete biological shield in order to give a definitive estimate of the radionuclide inventory and its dose-rate contribution.

1.3.3.4 Significant Neutron Activation Reactions All three materials evaluated, Aluminum 6061-T6 Alloy, Ordinary Cencrete, and A-36 Carbon Steel, contain isotopes which upon absorption of neutrons are transformed into radioactive isotopes.

Since the defueling will be scheduled at least three months after the reactor shutdown and the actual decommissioning of the facility will follow another three months sub-O sequent to defueling, those isotopes having a half-life of less than a few weeks will not contribute to the post-shutdown radiation level.

A list of the most significant neutron activation reactions contributing to the gamma radiation level during decommissioning operations and their radiation characteristics are shown in Table 1-6.

In addition, signifi-cance of carbon-14 production was briefly evaluated based on the data provided in the decommissioning reference research reactor report of NUREG/CR-1756. Since the BRR is not equipped with a graphite reflector of the type used in the reference research reactor, carbon-14 production is expected to be very small. However, a potential of contamination during the decommissioning of the thermal column needs to be considered in the waste disposal process.

1.3.4 Material Activation Strength and Surface Dose Rate Calculations Material activation strengths and the surface dose rates as functions of irradiation and decay times were calculated by the NAC code for the core support assembly, the tank liner, and concrete shield materials appli-cable to the three reactor operating positions:

exposure room, pool center, and thermal column.

As noted in Section 1.3.3.1, the neutron activation calculation based on the continuous reactor operation using EFPH value immediately prior to shutdown overestimates the source strength. Separate calculations indi-cate this overestimation is approximately a factor of three or more O

depending upon the half-life of radionuclide. On the other hand, the 1-31

Table 15 Material Compositions and Densities Used in Neutron Activation Caiculation Elemental Weight Percent Chemical Atomic A 3C 1

Element No.

A16061 T6 Ord. Concreto 2 8

Carbon Steel H

1 0.53 C

6 0.03 0.27 N

7 0.0013 0

8 47.0 Na 11 1.6 Mg 12 1.20 0.24 Al 13 95.85 4.40 Si 14 0.80 30.0 0.40 S

16 0.12 0.05 K

19 1.80 Ca 20 7.90 Ti 22 0.15 Cr 24 0.35 0.068 Mn 25 0.15 0.20 1.2 Fe 26 0.70 5.80 98.00 (Include _s rebar)

Co 27 0.0008 Ni 28 0.094 Cn 29 0.40 0.014 Zn 30 0.25 Nb 41 Mo 42 0.083 Eu 63 0.00001' Others 0.150 0.120 0.08 Sum 100.000 100.000 100.00 Density 2.70 2.35 7.80 1 Table E.1-1 of NUREG/CR-1756 (See Section 1.6.4) and chemical composition limits of wrought aluminum alloys by Aluminum Association 1966 2 Table C.1-2 of NUREG/CR-0130 "Technology, Safety and Costs of Oecommissioning a Reference Pressurized Water Reactor Power Station" 8 Table 2 of Volume 01.04, Section 1, Iron and Steel Products 1987 Annual Book of ASTM Standa.-ds

' Table E.1-5 of NUREG/Cn-0672 "Technology, Safety and Costs of Decom-missioning A Reference Boiling Water Reactor Power Staticn," rounded to the nearest unity 1-32 1

t O

Table 16 Significant Neutron Activation Gamma Emitting Radionuclides Six Months After Shutdown Radio.

Half.

Production

Major Gamma Radiation Material **

nuclide life Reaction (Energy in MeV, Intensity)

Irradiated Cr-51 27.8 y Cr-50 (n,y) 0.32 (9%)

A1. and Conc.

Mn-54 291 d Fe-54(n.p) 0.835(100%)

A1., Conc., & Steel Mn-55(n,2n)

A1., Conc., & Steet Fe-59 45 d Fe-58 (n,y) 1.09 (56%), 1.29 (44%)

A1., Conc., & Steel Co-57 270 d Nr-58 (n.np) 0.122 (87%), 0.136 (11%) Conc.

0.692 (0.14%)

Co-60 5.27 y Co-59 (n,y) 1.17 (100%), 1.33 (100%) Conc.

Zn-65 245 d Zn-64 (n,y) 0.51 (3.4%), 1.11 (49%)

A1.

Eu-152 12.7 y Eu-151 (n,y) 0.122 (37%), 0.245 (8%)

Conc.

0.344 27%), 0.779 (14%)

0.965 15%), 1.087 (12%)

1.113 14%), 1.408 (22%)

Note: Sources for half-life and major gamma radiations:

Lederer, C.M.

et al. Table of Isotopes (7th edition)

Production Reaction:

n - neutron, y - gamma, p - proton

- Materials Irradiated:

A1. -- Aluminum 6061-T6 alloy Conc. -- Ordinary concrete Steel -- Carbon Steel A-36 O

1-33

potential large impact of uncertainty associated with a minor trace ele-ment in concrete, or the effective neutron flux estimate, make an accu-rate prediction of the activation source strength to within a factor of five unlikely.

The surface dose rate of a relatively thin activated material such as the tank liner can be estimated by assuming that the effective neutron flux within the material stays constant and the self-absorption of activ-ation gamma is negligible. The surface dose rate of the tank liner, DRs, is expressed by:

(1) ors (Liner) = [ Svjt (mR/h) 2y Volumetric source strength of gamma energy i where Svi

=

expressed in MeV/cc-sec and function of irradia-tion and decay times as estimated by the NAC code Thickness of a thin material in cm t

=

Dose-rate conversion factor of gamma energy i kj

=

expressed in (MeV/cm -sec)/(mR/h)

[

Summation for all activation gamma energy groups

=

O i

'he other hand, for conservative estimates, we calculate the surface u e rate of a thick semi-infinite slab of activated material, such as a c3ncrete shield, assuming that the volumetric source strength of the activated species is spatially uniform throughout the slab at a value equal to that at the inner surface:

B Sv 3

(2) DRs (Uniform source = [ 2D

[1-E2(pi t)l (mR/h) in thick slab) j ii B Sv 3

=[2 f r uit 1 and E2(pi t)e o where Bj = symbolic buildup factor Svi and kj are the same as defined before pi Gamma attenuation coefficient of energy i in

=

the activating material E2(pi t)

Exponential integral

=

O 1-34

Forthickactivatedmategalwithavolumetricsourcestrengthdecreas-ing with thickness by e'

, the dose rate can be expressed by:

BS ors (Exponentialsource=[2\\vo fe Et(uz) dz t

(3) in a thick slab) j i

o B Svo j

(ut,})(mR/h)

[ 2pi j Ft k

where Bj, Svoj, pj, and kj are the same as defined before neutron attenuation coefficient m

=

( t, *) 3 { in (1 + *), if _ ut n 1 F1 For a thick concrete slab over 0.3 m (1 ft) thick, it was found that the surface dose rate can be expressed by the thick slab equation (2) with about 25% over-estimate compared to the exponential source equation (3).

Therefore, the thick slab equation (2) was used to estimate the surface dose rate of the concrete slab for conservatism.

To calculate the dose rate at a given distance away from an activated surface, the above described surface dose rate is converted into an applicable surface geometry, such as disc, ring, or cylindrical surface O

source as described in Section 1.3.5 Radiation Oose Rates Estimate.

1.3.4.f Core Support Assembly and Rotary Specimen Rack (Lazy Susan)

The core support assembly consists of the top and bottom grid plates and the core shroud.

The core shroud is exposed to the highest neutron flux compared to the top and bottom grid plates.

It was assumed that the Lazy Susan and the core shroud are exposed to the same neutron flux because of their close proximity to each other. The irradiation time was based on the total effective full power operating hours (EFPH) which was about 7,000 hours0 days 0 hours 0 weeks 0 months .

Table 1-7 shows the surface activation strengths in u Ci/cc and the surface dose rates in mR/h as a function of decay time after irradiation of the core shroud.

Table 17 Surface Activation Strength and Surface Dose Rate of Core Shroud As a Function of Decay Time After Irradiation i

Time After Core Shroud Material Irradiation 1.3 cm (1/2 in.) At 6061 T6 Alloy Surface Activity Surface Dose Rate (month)

(p Ci/cc)

(mR/h)

At Shutdown 1.6 x 10 1.1 x 108 5

O 3

2.2 x 108 3.4 x 10' 8

4 1.4 x 10 2.5 x 10' 5

9.7 x 102 2.0 x 10' 6

7.3 x 102 1.6 x 10' 1-35

The top and bottom grid plates are also fabricated from the same alumi-num alloy; but, their thicknesses are 1.6 cm (5/8-in.) and 1.9 cm (3/4-in.),respectively.

The effective neutron flux to the grid plates is somewhat lower than that of the shroud; therefore, the neutron activa-tion will be slightly lower than the shroud.

The surface activity and surface dose rate of the Lazy Susan at ex-pected to be nearly similar to that of the core shroud, becaust aoth componerts are fabricated of the same aluminum alloy and exposed to the similar neutron flux environment.

The only exception will be that the actual surface dose rate of thick components of the Lazy Susan will be proportionally higher than that of the shroud, which was based on 1.3 cm (1/2 in.) thick plate.

1.3.4.2 Reactor Position at Expc Sure Room Neutron-induced activation in the exposure rcom will be affected by a high level of both fast and thermal neutron fluxes incident upon the surrounding concrete and the pool liner materials.

The neutron flux at full power operation in the exposure room given in Table 1-4 was modi-led to calculate the applicable neutron flux incident on various sur-e.es as described below. An effective full power exposure time of o J hours was used in the calculation.

The exposure room cavity in the concrete shield is approximately 3 m (10 ft) wide, 3.7 m (12 ft) long, and 2.7 m (9 ft) high.

Shielding on three 8

sides consists of concrete with a density of 2.35 g/cm. The north wall of the reactor shield in the area around the exposure room is made of high density concrete (3.5 g/cc).

Boron is added to the inner 30.5 cm (12 in.) of the concrete walls, ceiling, and floor of the exposure room to minimize activation of the concrete. About 1.5 m (5 ft) of concrete is provided in the floor as shielding against soil activation.

In order to estimate the effective neutron fluxes at the surrounding concrete surfaces, the inverse distance square correction is applied to the known flux values adjacent to the core.

The nearest concrete sur-face is located at about 0.9 m (3 ft) and the furthest at about 3.5 m (11-1/2 ft) from the core center; thus, the effective flux varies by approximately a factor of 15.

The activation of average concrete sur-face was based on 1.5 m (5 ft) from the core center to estimate a typi-cal concrete neutron activation strength.

Table 1-8 shows the surface activation strengths in u Ci/cc and the sur-face dose rates in mR/h of the tank aluminum liner and the ave"age con-crete biological shield as a function of time after irradiatir.n.

O 1-36 t

l

A Table 18 Reactor Position at Exposure Room Surface Activation Strengths and Surface Dose Rates of Aluminum Tank Liner and a Typicai Concrete Shield Wall Time After Aluminum Tank Liner 0.6 cm (1/4 in.) Ordinary Concrete Shield Irradiation Surface Surface Surface Surface Activity Dose Rate Activity Dose Rate (month)

(p Ci/cc)

(mR/h)

(u Ci/cc)

(mR/h)

At Shutdown 1900.

76000.

32.

13000.

3 0.18 1.5 0.033 8.2 4

0.11 1.2 0.026 6.4 5

0.076 1.0 0.021 5.2 6

0.057 0.83 0.018 4.4 1.3.4.3 Reactor Position at Pool Center The neutron activation of the tank side wall liner nd the concrete shield walls for this reactor position will be low because the tank walls are shielded by more than 1.2 m (4 ft) of tank water.

The thermal neutron flux measured at the tank wall was about 5.5 x 10' n/cm -sec 8

(see Table 1-4).

O Table 1-9 shows the surface activation strengths and the surface dose V

rates of the tank aluminum liner and concrete biological shield as a function of time after irradiation based on the EFPH of 6,270 hours0.00313 days 0.075 hours 4.464286e-4 weeks 1.02735e-4 months , which represents nearly 90% of the time the reactor was in the pool cen-ter position. As seen from the surface dose rate columns, the radiation level at 0.9 m (3 ft) from the surface of both the tank liner and con-crete shield will be less than 5 u R/h (5 x 10-8 mR/h) at six months after irradiation; therefore the pool side wall liner and concrete need not be removed due to Ectivation considerations.

Table 19 Reactor Position at Tank Center Surface Activation Strengths and Surface Dose Rates of Tank Side Walls Time After Tank Liner 1.9 cm (3/4 in.)

Concrete Shield Wall Irradiation Surface Surface Surface Surface Activity Dose Rate Activity Dose Rate l

(month)

(u Ci/cc)

(mR/h)

(p Ci/cc)

(mR/h)

At Shutdown 2.1 x 10-2 2.2 x 10' 2.7 x 10-3 9.6 x 10-1 3

2.8 x 10-5 6.7 x 10-*

3.6 x 10-5 1.0 x 10-2 4

1.8 x 10-5 4.9 x 10-'

2.2 x 10-5 7.1 x 10-8 5

1.2 x 10-5 3.8 x 10-'

1.4 x 10-5 4.5 x 10-8 l

6 9.1 x 10-'

3.0 x !G-'

9.2 x 10-'

3.0 x 10-8 O

1 l

1-37

The tank floor is located about 0.9 m (3 ft) from the core center.

The 7

2 thermal flux measured at the tank floor was about 7.0 x 10 n/cm -sec (see Table 1-4.)

The tank floor activation due to thermal neutrons will 7

be h'.gher than the pool wall by approximately 7.0 x 10 /5.5 x 10' i 1300.

Since the thermal neutron flux decreases at the floor away from the core vertical center line, the floor surface activation strength and cor-responding surface dose rate can be calculated as a function of the incident thermal neutron fluxes. After the reactor removal, the dose rate at 0.9 m (3 ft) from the floor along the vertical core center line can be expressed by the sum of dose-rate contributions from a disc and a series of ring sources.

Table 1-10 shows the floor activation source strengths and corresponding surface dose rate of the aluminum tank floor and concrete floor based on the ef fective thermal flux of 6.0 x 10' n/cm:-;ec.

Table 1 10 Reactor Position at Pool Center Surface Activation Strengths and Surface Dose Rates of Rear. tor Tank Floor Aluminum Floor Liner Concrete Floor Slab Time After 1.3 cm (1/2 in.)

Irradiation Surface Surface Surface Surface Activity Dose Rate Activity Dose Rate (months)

(u Ci/cc)

(mR/h)

(p C1/cc)

(mR/h)

At Shutdown 19.0 1300.

2.50 870.

3 0.026 0.39 0.032 9.6 4

0.016 0.29 0.020 6.5 5

0.011 0.22 0.013 4.1 6

0.0083 0.18 0.0083 2.8 1.3.4.4 Reactor Position at Thermal Column The neutron activation of the pool liner and concrete shield wall for this reactor position is also governed by the thermal neutron flux; how-ever, the activation of the carbon steel shield dbove the thermal column and the beam ports near the reactor need a separate activation evalua-tion.

The riearest tank wall is about 1.2 m (4 f t) from the core center compared to the tank wall 1.5 m (5 ft) from the reactor at the pool cen-ter case; therefore, the increase in the neutron activations of the tank liner and concrete wall is proportional to the ratio of thermal neutron flux incident on each wall, namely 2 x 10'/6 x 10' = 33 or about 30.

Since the 30.5-cm (12-in.)-thick carbon steel shir:ld is 1.5 m (5 ft) measured vertically and the shortest distance from the core center is as close as about 137 cm (54 in.), a conservative thermal neutron flux value of 1 x 10' n/cm -sec, (at 127 cm l50 in.] from the core center) a will be used to estimate the steel activation.

The irradiation time will be based an EFPH of 590 hours0.00683 days 0.164 hours 9.755291e-4 weeks 2.24495e-4 months .

Table 1-11 shows the surface acti-vation strengths and surface dose rates of carbon steel, and Table 1-12 shows the same for the aluminum liner and concrete wall.

1-38

Table 1 11 O.

Reactor Position at Thermal Column i

Surface Activation Strengths and Surface Dose Rates of Carbon Steel Shield Time After Carbon Steel Shield Shutdown Surface Activity Surface Dose Rate (months)

(u Ci/cc)

(mR/h)

L At Shutdown 3.4 x 10-4.8 x 10' I

3 8.7 x 10-'

8.7 x 10-2 4

5.5 x 10-'

5.5 x 10-2 5

3.5 x 10-'

3.4 x 10-2 6

2.2 x 10-*

2.2 x 10-2 4

Table 1 12 Reactor Position at Thermal Column Surface Activation Strengths and Surface Dose Rates of Tank Liner and Concrete Shield Wall Aluminum Liner Ordinary Concreto Time After Surface Surface Surface Surface Shutdown Activity Dose Rate Activity Dose Rate s

(month)

(u Ci/cc)

(mR/h)

(p Ci/cc)

(mR/h)

{

At Shutdown 6.2 x 10-1 6.5 x 10 8 7.9 x 10-8 2.8 x 10 3

6.2 x 10-'

5.1 x 10'8 3.5 x 10-'

9.8 x 10-8 4

1.6 x 10-'

3.4 x 10-8 2.2 x 10-*

6.3 x 10-2 5

9.2 x 10-8 2.4 x 10-8 1.3 x 10-'

4.0 x 10-8 i

6 5.8 x 10-5 1.7 x 10-8 8.5 x 10-5 2.6 x 10-8 l

1.3.5 Radiation Dose Rates Estimate I

J t

The expected external dose rates from the core support assembly, inside the exposure room and inside the reactor tank under dry conditions, were calculated based upon the neutron activation calculation described previously, The surface dose rates of various activat. materials were used to esti-l mate a dose rate at 0.9 m (3 ft) from the various activated surfaces and

[

their appropriate geometry.

For example, the core support assembly can 1

be expressed as a cylindrical surface source and the grid plate by a i

disc source. Tank liner and concrete slab are expressed by the sum of a i

disc source and a series of partial ring sources.

l

?

The radiation characteristics of the major radionuclides considered in the activation analyses are summarized in Table 1-6.

For the dose rate l

calculation, six months decay after the reactor shutdown was assumed.

I t

l t

1-39 l

1.3.5.1 Core Support Structure The estimated surface (contact) dose rate of the core shroud is calcu-lated to be 16 R/h six months after the shutdown, based on the effective 18 2

thermal flux level of 5.0 x 10 n/cm -sec, which is applicable to the Lazy Susan located just outside of the core shroud, and an EFPH of 6,270 hours0.00313 days 0.075 hours 4.464286e-4 weeks 1.02735e-4 months .

Because of this high estimated contact dose rate for the core support assembly, all dismantling operations must be performed under at least 1.5 m (5 ft) of water in order to limit the dose rate to less than 1 mR/h.

Significant gamma-ray contributions from the aluminum alloy were found to be from Zn-65 and Fe-59 at 6 months after shutdown as indicated in Table 1-13.

Table 1 13 Significant Gamma Emitters in Neutron Activated Core Support Structure at Six Months After Shutdown Surface Activity Surf ace Volumetric Concentration Gamma Strength Radionuclido (p Ci'cc)

(MeV/cc sec)

Cr-51 1.2 x 102 1.8 x 108 8

Co-60 1.9 x 10-2 1.3 x 10 Mn-54 2.0 x 10' 6.3 x 10

Fe-59 8.5 x 108 3.8 x 10' 6

2 9.5 x 10 Zn-65 5.2 x 10 r

Based on f ast neutron flux of 1.5 x 10 ' n/cm -sec and thermal neutron 22 2

flux of 5.0 x 10 n/cm -sec.

1.3.S.2 Inside Exposure Room The estimated peak dose rate inside the exposure room is approximately 9 mR/h at six months after the EFPH of 170 hours0.00197 days 0.0472 hours 2.810847e-4 weeks 6.4685e-5 months . This dose rate esti-mate was based on the maximum activity in concrete floor, ceiling, and walls.

The tank liner contribution was found to be insignificant com-pared with the activated concrete.

Significant gamma-ray contributions from the activated concrete are from Mn-54 and Fe-59 at six months after shutdown, as indicated in Table 1-14.

Table 1 14 Significant Gamma Emitters in Neutron Activated Concrete in Exposure Room at Six Months After Shutdown Surface Activity Surf ace Volumetric Concentration Gamma Strength Radionuclide (p CI'cc) _

(McV/cc sec) 2 Mn-54 1.2 x 10-2 3.8 x 10 2

Fe-59 2.3 x 10-8 1.0 x 10 Co-60 7.2 x 10-5 6.7 x 10 Eu-152 1.2 x 10-8 5.0 x 10-'

l-40

/O Based on fast neutron flux of 4.0 x 10' n/cm -sec and thermal neutron 2

O flux of 1.9 x 10e n/cm -sec.

a 1.3,5,3 Inside Reactor Tank The highest dose rate in the reactor tank is expected to occur at the center of the tank floor where the irradiation time is the longest and neutron flux is also the highest because the tank floor is only 0.9 m (3 ft) from the core center. The estimated dose rate at 0.9 m (3 ft) from the tank floor center is about 1.5 mR/h at three months and about 0.5 mR/h six months after the EFPH of 6.270 hours0.00313 days 0.075 hours 4.464286e-4 weeks 1.02735e-4 months .

Because over 1.2 m (4 ft) of water shield separates the side walls from the reactor, the activation contribution of the tank liner and concrete side wall to the selected dose point near the tank inner surface is negligibly small as reflected in the surface dose rates shown in Table 1-9.

The activation strengths of the tank floor corresponding to the three reactor positions were investigated as a function of irradiation time with a constant thermal neutron flux incident on the floor.

The irradi-ation time at the exposure room, pool center, and thermal column was 2.4%, 89.3%, and 8.3% respectively, of the total EFPH for three reactor positions.

It was found that the activation strengths of the concrete floor and also the dose rates of three positions varies proportionally to 1:10:3 respectively.

Since the dose rate at the pool center is esti-mated to be 0.5 mR/h, the dose rate at the reactor location near the 7

exposure room due to floor activation only is estimated to be about 0.05 (V

mR/h six months after shutdown.

Significant gamma-ray contributions from the activated concrete were found to be from Fe-59 and Co-60 at six months after shutdown, as indicated in Table 1-15.

Table 1 15 Significant Gamma Emitters in Neutron Activated Concrete Floor in Reactor Tank at Six Months After Shutdown Surface Activity Surf ace Volumetric Concentration Gamma Strength Radionuclide (p Cl/cc)

(MeVicc sec)

Cr-51 2.5 x 10-'

3.6 x 10-2 2

Fe-59 7.3 x 10-8 3.2 x 10 Co-60 7.4 x 10-'

6.9 x 10 Eu-152 1.4 x 10-'

5.9 x 10' Based on thermal neutron flux of 6.0 x 10' n/cm -sec.

2 1.3.6 Estimate of Volume of Activated Material To Be Removed Estimated volume of concrete requiring demolition to freet the dose rate of 5 pR/h was estimated to be approximately 82 m3 (2,900 ft' or 107 O

yd').

The final volume determination must be made based on concrete

(/

core sample measurements as previously described.

1-41

1.3.6.1 Exposure Room Because of boron addition to the inner 30.5-cm (12-in.) concrete surface, the thermal neutron activation is assumed to be completely suppressed, but builds up in the non-borated concrete due to thermalization of fast neutrons.

Since the activation source strength is proportional to neu-tron flux distribution, the surface dose rate can be estimated as a func-tion of neutron flux.

To meet the dose rate criterion of 5 uR/h at 0.9 m (3 ft) from the activated concrete surface, about 76 cm (30 in.)

cf the east wall facing the reactor to 91.4 cm (36 in.) of other concrete surfaces closer to the reactor need to be removed.

1.3.6.2 Reactor Tank, Wall, and Floor The north and south sides of the reactor tank walls are shielded by over 1.2 m (4 ft) of water. Calculations indicate that no demolition of the concrete is required to meet the dose rete criteria.

However, the tank floor is found to be somewhat activated ';o the extent that the dose rate at 0.9 m (3 ft) is approximately 0.5 mR/h six months after shutdown.

An estimated radius of about 105 cm (42 in.) from the core center will be affected by the neutron activation.

The depth of concrete floor requir-ing removal to meet the dose rate criteria is estimated to be about 50 cm (20 in.) for the Pool Center, 46 cm (18 in.) for thermal column, and 30 cm (12 in.) for the exposure room positions.

1.3.6.3 Thermal Column The carbon steel shield above the thermal column is also somewhat activ-ated giving a contact dose rate of -0.02 mR/h at six months.

The dose rate at 0.9 m (3 ft) from the carbon steel surface based on a disc source having a radius of 0.8 m (2.5 ft) is estimated to be 0.01 mR/h six months after shutdown.

Therefore, it is marginal and may require the removal of the steel shields at least partially.

The tank concrete wall below the carbon steel shield is estimated to be 0.04 mR/h six months after shutdown. Therefore, approximately 15-30 cm (6 to 12 in.)

of concrete wall thic< ness, including the tank liner up to the thermal column height, will need to be removed.

1.4 DECOMMISSIONING ALTERNATIVE The facility in which the BRR is housed, Room 1140 in Etcheverry Hall, is used concurrently as a nuclear research laboratory and for classroom activities.

These activities operate under a State of California Radio-active Material License.

The University plans to continue use of these portions of Room 1140, and after the dismantling of the reactor, the affected portion of the room will be refurbished for reuse as part of the ongoing research program, to be operated under the state license.

At some future date the University intends to construct a computer sci-ences building on this site.

The decommissioning alternative chosen that fits these requirements is DECON, immediate removal of radioactive material down to residual levels which permit release for unrestricted access.

Other alternatives were considered inapplicable.

1-42

)

I l

l O

The University requires minimum disruption to academic activities.

Therefore DECON-related activities are planned for the summer recess period. Some of the laboratory activities will continue during the dis-mantling period. This plan is structured to permit these activities to l

continue in a limited manner by providing for:

o Temporary contamination control barriers around existing oper-l ating experimental equipment.

i o

Separation and control of entry of University staff and students.

i o

Provision of temporary storage and contamination protection for inactive areas and equipment Portions of these areas may contain radioactive materials under the con-trol of the State of California radioactive Materials License, r

]

With regard to the reactor and its auxiliaries, this OP includes meas-ures to reduce radiation to levels that will permit termination of the license and unrestricted use of the area occupied by the reactor.

i 1.5 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES l

The Chancellor of the Berkeley Campus of the University of California is 2

responsible for the decommissioning of the BRR. An organization has O

been developed to oversee the decommissioning program, procure outside services and materials, administer contracts, review and approve docu-a ments, obtain permits and licenses, provide quality assurance, and pro-1 vide personnel experienced in TRIGA reactor operations, health physics, i

l and security.

The decommissioning organization is shown in Figure 1-8.

6 l

Most of the work of detailed planning, decontamination and demolition I

will be performed by outside contractors.

)

i 1.5.1 Decommissioning Steering Committee i

This Decommissioning Steering Committee is responsible for overall Uni-

]

versity policy regarding decommissioning of the BRR.

It guides the Pro-1 gram Committee in making major decisions.

The Steering Committee is chaired by the Provost of the Professional Schools and Colleges.

It has l

as members two senior faculty, the Chair of the Decommissioning Program j

Committee, the Campus facility Planner, the Director of Design and Con-struction Services, the Campus Radiation Safety Officer and a student t

representative.

1.5.2 Decommissioning Program Committee 5

This committee consists of fourteen members, thirteen associated with l

the University and one representative of the City of Berkeley.

The University members include representatives from the Nuclear Engineering I

Department, Mechanical Engineering Department, Electrical Engineering j

I i

l 1-43 I

F:GLt E 1-8 UNIVERSITY ORGANIZATION FOR DECOMMISSIONING CHANCELLCR BERKELEY CRMPUS VICE CHRNCELLOR THE VICE CHANCELLOR BUSINESS RND ADMINISTRATIVE SERVICES I

DECOMMISSIONING CRMPUS ENVIRONMENTAL PLANNING ST " N

~~

{ POLICE HEALTH & SAFETY CC ITTEE FACILITIES UC OUALITY l

l l

MANAGEMENT PSSURANCE y

SUPERVISOR

...-__2.__________'.n._-__

?

7 l_,

PROGRAM COMMITTEE DECOMMISSIONING

_,l g

l l

RADIATION REACTOR

. PROGRAM REACTOR SAFETY HAZARDS l DIRECTION l------------

HEALTH COMMITTEE COMMITTEE y

PHYSICIST UC DECOMMISSIONING PROJECT ENGINEER DEPUTY UC I

---- CONTRACTUAL OIRECTION GECOMMISSIONING PROJECT ENGINEER EXECUTIVE ENGINEER CONTRACTORS 6-JPN-89 Z53: [ : 3 4.13 < E I C 754. E G'..

O O

O

and Computer Science Department, Reactor Hazards Committee, Radiation J

Safety Committee, Campus Planning Office, Department of Facilities Man-r agement, the Office of Environmental Health and Safety, and the student a

body. Their qualifications meet the standards of education, training, and experience as determined by University policy.

The committee's responsibilities include:

o Advise the Steering Committee and the Chancellor on all mat-ters pertaining to the decommissioning activities o

Provide direct oversight and planning o

Review and approve all decommissioning planning and operation documentation o

implement all actions regarding licensing, permitting, and document submittals o

Maintain and oversee interfaces with federal, state, and 1ccal agencies o

Provide technical and programmatic direction to the decommis-sioning contractors The Chairman of the Decommissioning Program 'ummittee is also the Reat-O tor Administrator, who is responsible for administering reactor opera-tions under the present NRC license.

He is the UC interface with the Nuclear Regulatory Commission and other federal agencies concerning reactor operations and decommissioning assisted by the UC Decommission-ing Project Engineer. He directs the work of the UC Decommissioning Project Engineer and the Deputy UC Decommissioning Project Engineer.

1.5.3 Reactor Hazards Committee The Reacter Hazards Committee reviews, approves and audits matters of radiation health and safety pertaining to reactor operations and to each stage of decommissioning.

The Committee is directly responsible to The

.i Vice Chancellor of the Berkeley Campus and advises the Decommissioning Steering Committee, the Decommissioning Program Committee, and the UC Decommissioning Project Engineer.

1.5.4 Office of Environmental Health and Safety The Office of Environmental Health and Safety (EH&S) provides the fol-lowing support to the reactor decommissioning program:

o The Reactor Health Physicist and a back-up Health Physicist in his absence o

Review of contractor's industrial safety, fire, and industrial hygiene programs 1

1-45 l

o Interface with the State of California concerning State licensing requirements for decommissioning operations, by the Campus Radiation Safety Officer 1.5.5 Radiation Safety Committeo The Radiation Safety Committee is responsible for the following activi-ties during the reactor decommissioning operations:

o Review of decommie.sioning operations with respect to their impact on the env'.ronment and on Campus areas under the juris-diction of the State of California Radioactive Materials License (State License) o Review of radiation monitoring results to assure that decom-missioning activities meet the requirements of the State License 1.5.6 Physical Security The Chief of UC Berkeley Campus Police is responsible for general physi-cal security in and around Etcheverry Hall during reactor operations and during dismantlement.

The UC Decommissioning Project Engineer is responsible for oversight of security measures for the reactor facility and for security measures employed by the decommissioning contractors.

1.5.7 Planning and Facilities Management, Design and Construction Services; Executivo Engineer The representative of this office will provide overall contractual direction to the Decommissioning Contractor.

The representative will be assisted by an Executive Engineer from a nuclear engineering firm who will prepare construction bidding documents, review bids and schedule the work in accordance with on-going needs of the College of Engineering and of the campus.

The Executive Engineer will supervise the worker safety training and will provide the following services:

o Review of contractors' planning and operations to assure adherence to procedures and quality assurance provisions o

Oversee monitoring equipment calibration by contractors o

Review to assure adherence to terms and conditions of the contract o

Inspection to insure satisfactory execution of the work o

Administer changes and additions to the contract o

Fiscal management O

l-46

l The Executive Engineer will also provide the following, under the pro-

\\

grammatic supervision of the Decommissioning Program Comittee.

o Radiation health and safety training to contractor operating personnel o

Initial radiation survey o

Final radiation survey 1.5.8 Decommissioning Contractors Contractors will be selected by the University through their normal procurement practices.

The selected contractors will be knowledgeable of the proposed work and experienced in the execution of each task and will provide all labor, equipment, and materials directly or through appropriate subcontracts.

The contractors will provide:

o A complete set of procedures o

Medical examination of operating personnel o

Execution of work in accordance with the contractual terms and conditions 1.5.9 UC Decommissioning Project Engineer The UC Decommissioning Project Engineer will be the Reactor Supervisor, who is the person responsible for present reactor operations, or a per-son with equivalent qualifications.

He is responsible for coordinating all decommissioning activities with cognizant Campus organizations and individuals and with outside contractors directly or indirectly involved with the decomissioning.

He ensures that all federal and state rules and requirements on decommissioning are followed.

He is responsible for supervising access control to the reactor facility, including interfac-ing with the Campus Police. He provides technical and programmatic direction to the decommissioning contractors through the Executive Engineer.

He reports to the Chairman of the Decommissioning Program Committee / Reactor Administrator and is assisted by the Deputy Decommis-sioning Project Engineer.

1.5.10 Deputy UC Decommissioning Project Engineer The Deputy UC Decommissioning Project Engineer is the Chief Reactor Operator, or a person with equivalent qualifications. He assists the UC Decommissioning Project Engineer and serves as his deputy. He coordi-notes the daily activities of the decommissioning program to assure compliance with all rules, regulations, and safety procedures.

He coor-dinates the maintenance of UC-owned and UC-operated equipment. He pro-O vides the operational interface with other University departments for maintenance, installation, and other services.

1 47

1.5.11 Reactor Health Physicist O

The Reactor Health Physicist is responsible for performing the following activities in the decommissioning program:

o Review of the contractor's industrial safety, fire and indus-trial hygiene programs by EH&S o

Performance of UC radiation safety surveys of the decommis-sioning operations as independent : hecks of the Executive Engineer surveys, to ensure compliance with appropriate Nuclear Regulatory Commission (NRC)

State, and Campus rules and regulations o

Certification that all radioactive material sent off Campus meets NRC, Department of Transportation State, and Campus rules and regulations 1.5.12 UC Ouality Assurance Supervisor The UC Quality Assurance Supervisor is responsible for establishing and implementing the quality assurance plan for decommissioning of the BRR, and working with the affected branches of the University organization, in order to assure his independence of action and the independence of the UC QA program from costs and schedule constraints, the UC QA Super-visor reports directly to the Chairman of the Decommissioning Steering Committee.

The UC QA Supervisor has sufficient authority and organiza-tional freedom to identify quality problems; initiate, recommend, or provide solutions; and verify implementation of solutions.

The UC QA Supervisor will prepare and carry out a ccmprehensive system of planned and periodic audits /surveillances using documented check lists to verify compliance with all aspects of the QA program and to determine its effectiveness.

1.6 REGULATIONS, REGULATORY GUIDES, AND STANDARDS While the terms regulation, guideline, standard, and criteria are often used interchangeably, it is necessary to make distinctions. Regulations are rules having the force of law issued by an executive authority or a government. A guideline is a recommended practice or guiding informa-tion supplied by an agent with implied intimate technical knowledge, A

standard is established by "authority" as a rule to follow.

In general, standards set forth limits or definitive ways of accomplishing an objec-tive, whereas criteria provide a yardstick for comparison as a basis for judging the acceptability of a practice.

This section identifies and discusses the regulations, guides, and stan-dards applicable to decommissioning of the BRR, O

1-48

1.6.1 Applicable Regulations Federal regulations that are applicable to decommissioning of research reactors appear in the Code of Federal Regulations (CFR) and in the legal codes of the State of California. While all the federal govern-ment regulations are contained in the CFR, different titles are associ-ated with various government agencies, commissions, and administrations.

For example Title 10 -- Energy, pertains to the NRC; Title 40 --

Protection of Environment, includes regulations of the Environmental Protection Agency (EPA); and Title 49 -- Transportation, deals with transportation of hazardous materials.

Some of the regulations under these Titles have immediate applications in the decommissioning, and I

some have application by implication of related subject matters.

1.6.1.1 State of California Title 8 Safety and Health Pequirements for Construction Title 14 Chapter 3 Guidelines for Implementation of the California Environmental Quality Act 1.6.1.2 Code of Federal Regulations 10 CFR Part 19 Notices, Instructions, and Reports to Workers; Inspections 10 CFR Part 20 Standards for Protection Against Radiation 10 CFR Part 30 Rules of General Applicability to Domestic Licensing of Byproduct Material 10 CFR Part 50 Domestic Licensing of Production and Utilization Facilities i

i 10 CFR Part 51 Licensing and Regulatory Policy and Procedures for Environmental Protection 10 CFR Part 61 Licensing Requirements for Land Disposal of Radioactive Waste 10 CFR Part 71 Packaging of Radioactive Material for Transport and Transportation of Radioactive Material under Certain Conditions 10 CFR Part 140 Financial Protection Requirements and Indemnity Agreements 49 CFR Parts Department of Transportatinn Hazardous Material 170-199 Regulations 40 CFR Part 260 Hazardous Waste Management System General 40 CFR Part 261 Identification and Listing of Hazardous Wastes 1-49

40 CFR Part 262 Standards Applicable to Transporters of Hazardous Waste 40 CFR Part 61 National Emission Standards for Hazardous Air Pollutants 1.6.2 Rogulatory Guidos In addition to regulations that carry the force of law, regulatory bodies such at the NRC and EPA prepare regulatory guides that, among other things, suggest agency-approved methodology and solutions to proolems. While compliance with them is not a legal requirement, they generally provide the most effective method of obtaining approval for a partir.ular courr9 of action.

1.6.2.1 NRC Regulatory Guides 1.8 Personnel Qualification and Training 1.16 Reporting of Operating Information 1.86 Termination of Operating Licenses for Nuclear Reactors 1.143 Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants e

3.X Draft Standard Format and Content of Decommissioning Plans for 10 CFR 30, 40 and 70 Licenses.

8.2 Guide for Administrative Practices in Radiation Monitoring 8.3 Film Badge Performance Criteria 8.4 Direct-Reading and Indirect-Reading Pocket Oosimeters 8.6 Standard Test Procedures for Geiger-Muller Counters 8.7 Occupational Radiation Exposure Records Systems 8.8 Information Relevant tc Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be As low As Reason-ably Achievable 8.9 Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program 8.10 Operating Philosophy for Maintaining Occupational Rcdiation Exposure As Low As Reasonably Achievable 8.15 Acceptable Programs for Respiratory Protection O

l-50

7 4

{v 1.6.3 Standards A number of institutions or technical societies, such as the Interna-tional Commission on Radiological Protection (ICRP), the National Com-mittee on Radiation Protection and Measurement (NCRP), and the American Ns.ional Standards Institute (ANSI), publish standards. While not car-ryiig the force of law, they do re9 resent the formal statement of tech-ni;al opinion of the bodies issuing them.

1.6.3.1 ANSI Standards AiC: li13.13 Control of Rhdioactive Surface Contamination of Material, Equipment, and Facilities to be Released for Uncontrolled Use (Oraft)

ANSI Z88.2-1980 Practices for Respiratory Protection ANSI N13.1 Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities ANSI N323-1977 Radiation Protection Instrumentation Test and Calibration ANS!/ANS-15.10-1981 Decommissioning of Research Reactors 1.6.4 Informal Guidance and Te( hnical Reports Informal guidelines published by the NRC can be found in NUREG docu-ments, Branch Technical Position papers, Inspection and Enforcement Branch rotices, and other external or internal docum2nts.

There are nuc.erous technicat reports published by the NRC and the 00E that support the subject of the decommissioning of research reactors.

The following documents are directly applicable:

o NUREG/CR 1756 "Technology, Safety, and Costs of Decommission-ing Reference Nuclear Research and Test Reactors" and addenda, o

NUREG/Cr. 2082 "Monitoring for Compliance with Decommissioning Terd nation Survey triteria."

o NUREG-0586 "Draft Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities."

o NRC's "Guidance and Discussion of Requirements For an Applica-tion to Terminate a Non-power Reactor Facility Operating License."

o NUREG/CR-2241 "Technology and Cost of TerF nation Surveys

(

Associated with Decommissioning of Nuclear Facilities."

i 1-51

1.6.5 Permits / Licenses Covering 1140 Etcheverry Hall Activities Agency LlCense/ Permit Number NRC Facility Operating License R-101 as amended State CA Radioactive Material License 1333-62 as amended State WA Radioactive Waste Disposal Permit 7221 1.7 TRAINING AND OUALIFICATIONS The following is a summary of the training program.

Details are dis-cussed in Chapter 2.

The training and qualifications of personnel will depend on the individ-ual task assignments and the experience of the University personnel and contractort assigned to the decontamination and dismantlement of the BRR.

1.7.1 Training Program Descriptions Training topics will depend upon:

o The health and environmental impacts of planned operations; o

Applicable regulations, standards, and guidelines pertinent to operations involving radiologically or chemically hazardous materials / waste; o

The purpose of the training; and o

The personnel to be trained (e.g. their education, training, and experiance).

Documentation of training shall be by appropriate Environmental Health and Safety (EH&S) form, currently ' Training Record Sheet."

Personnel having received substantial radiaticn safety t. raining within the past two years may, upon demonstration of their kncwledge to the satisfaction of the Reactor Health Physicist be cxempt frcm general em-ployee training.

The anticipated training programs follow:

o General Employee Training: General employee training in com-pliance with Title 10 CFR Part 19.12 will be required for all personnel involved with radioactive materials or working in the vicinity of radioactive materials.

o Respiratory Protection:

Respiratory protection training will be implemented to meet project requirements in compliance with ANSI Z-88.2, NRC Reg Guide 8.15, NRC NUREG-0041 "Manual of Respiratory Protection" and 29 CFR 1910.134, 1-52

o Hearing Conservation: A hearing conservation training program

)

will-be conducted to implement 29 CFR 1910.95.

o Hazard Communications: Hazard communications training in com-pliance with 29 CFR 1910.1200 will be conducted as applicable, o

Technical Training: Job activity simulations or briefings may be conducted to ensure proper handling and use of equipment, health and safety issues and ALARA considerations. These will depend u;on the task and personnel and will be documented in the trainir.g record 1.7.2 Administration and Recordkeeping The Reactor Health Physicist shall be responsible for the training pro-gram and maintenance of personnel training, qualification, and exposure records..The Executive Engineer / Contractor's responsibilities for training are specified in Section 2 1.2 and 2.2.2.

Complete up-to-date training, qualification and exposure records will be maintained on all personnel. The records will include the following:

o Bioassay analysis o

Personnel exposure records

[]/

o Individuti dosimeter readings as related to daily tasks and

(

work procedures o

Respiratory protection qualifications (medical clearance and fit test) o Audiogram esselts o

Training records 3

Visitor logs and exposure inforrratior, f

O 1-53

CHAPTER 2 p

OCCUPATIONAL AND RADIATION PROTECTION PROGRAMS Q

l

2.0 INTRODUCTION

The Occupational and Radiation Protection Programs (0RPPs) for the Berkeley Research Reactor (BRR) decommissioning project consist of a set of policies, procedures, and instructions enacted to protect workers, the general nublic, and the environment. Objectives of the ORPPs include:

I:nsuring the health and safety of personnel by providing pro-o tection programs which include a commitment to the principles of maintaining exposures as low as reasonably achievable (ALARA) o Minimizing the exposure of the general public and the environ-ment to the radioactive and/or ha ardous chemical effluents that may be released during the decommissioning activities c

Identifying and separating contaminated structures, surfaces, systems, and components from those which are not contaminated o

Disposing of contaminated and non-contaminated components and materials properly and safely V

o Ensuring that the facility meets all radiological decommis-sioning requirements and is ready to be released to unre-stricted use The ORPPs provide integrated ccctJ>ational health, health physics, indus-trial hygiene, and ufety elemer,ts.

To meet NRC reporting guidance, these clements are discussed in Sections E.1, Radiation Protection Prcgram, and P.2, Industrial Safety and 9ygiene Program. Inis format creates discussion repetition in the text because normally a respiratory protecticn program is provided for all airborne hazards for radiologi-cally and chemically hazardous substances, not separate programs.

Addi-tionally, personnel training is an integral facet of both portions of the ORPPs; however, discussion of certain training items has been pre-sented in Section 1.7, Training and Qualifications.

2.1 RADIATION PROTECTION PROGRAM The Radiation Protection Program (RPP) for the decommissioning project includes requirements to monitor radiation and radioactive materials, to control distribution and releases of radioactive materials, and to keep radiation exposure for individuals and the collective radiation exposure within the limits of 10 CFR 20 and at levels ALARA.

2.1.1 Personnel The Executive Engineer's health physics personnel will become familiar with the location and magnitude of sources of radiation to which per-sonnel may be exposed during the course of work, in addition, b'alth 2-1 l

physics personnel will become familiar with the use of approved Radiation /

Pazardous Work Permits (RHWPs), Standard Work Permits, and detailed Work Procedures.

The Executive Engineer's health physics personnel will include:

o Radiological Control and Safety Officer (RC&S0) o Radiological engineer / health physics supervisor o

Health physics technicians ORPPs procedures will be prepared, and guidance provided by the RC&SO and a Certified Health Physicist.

2.1.2 Training All persons working or frequenting the DECON area of Room 1140 of Etcheverry Hall during DECON activities will be presented with instruc-tions in radiation safety by the Executive Engineer prior to any such work activities. Although all persons will receive training, not all persons should receive the same tyne of training.

To prevent duplica-tion and to make efficient use of time, project personnel will be grouped in three categories and will be given training commensurate with poten-tial radiological problems to be encountered in their scope of work.

The three groups formed are:

o Non-radiation workers o

Radiation workers directly involved in handling radioactivc/

contaminated materials and entering radiation areas o

Persons directing the activities of radiation workers Personnel having received instruction in radiation and reactor labora-tory safety at the campus within the past two years may be exempt from training at the discretion of the project Reactor Health Physicist (RHP).

The OPPPs includes a training program for all pcrsons who will be in-volved in the decommissioning project. All decommissioning personnel will receive instruction concerning radiation protection through orientation / training.

Each worker will attend one of these orienta-tions and will be evaluated by an examination upon the conclusion of the training. A passing score is required before these personnel are occupt.tionally exposed to radiation.

The training program will adhere to the guidelines of the Institute for Nuclear Power Operations (INP0)

General Employee Training Program.

The RC&S0 or a designee will administer comprehensive radiation worker training, Cal OSHA hazardous communication training, and an indoctrina-tion session for all workers and to all on-site management personnel.

Additional ariefings and practical factors training will be performed on a routine basis to familiarize personnel with work procedures, equipment, radiation control requirements, and hazards associated with the various 2-2

work elements. Docuinentation of individual training and qualifications pgj will be filed and maintained in the health physics office at the work site. Records of training will be maintained which will include trainee's name, date, subjects covered during training, equipment in which training was received, results of written tests, and the instruc-tor's name.

Specialized training will be provided to e.nployees before they are allowed to undertake jobs with high exposure potential. Employees and supervisors will receive this training as part of the crew. Mockups and other training aids may be used to train the.,crkers so that time spent in high exposure rate areas is minimized.

Objectives of the training program are to accomplish the following:

o Provide involved personnel with information about radiologi-cally and chemically hazardous substances, sources and types, exposure routes, and effects o

Provide information on the ORPPs for the decommissioning proj-ect in order to enable each person to comply with health and safety rules and to respond properly to all conditions o

Provide instruction in the fundamentals of radiation and chem-ical protection to enable individuals to maintain their own exposure and collective exposure ALARA

\\ -

o Provide information on personal protection equipment, monitor-ing instruments, and equipment available, and how to use them o

Infonn eacn person about NRC, EPA, Cal OSHA, state and local license regulations and requirements, and Campus rules and regulations concerning health and safety 2.1.3 Administrative and Radio!ogical Co7trols 1

Ad:ninistrative and radiological controls comprise the measures taken to limit re.diation exposure, spread of contamination, and the numerical limits for exposure.

2.1.3.1 Exposure Limits Limits on the radiation exposure of individual workers involved in radiation-related work have been set for the nuclear industry by the NRC and are applicable here. These limits are stated in 10 CFR 20, "Stan-dards for Protection Against Radiation." However, in order to ensure that individual and collective doses are kept ALARA, the contractor will j

establish work procedures and a radiation / hazardous work permit system to ensure that all work performed is evaluated with respect to the ALARA l

philosophy during the decommissioning of the BRR.

P Personnel classified as radiation workers will have their whole body doses administrative 1y controlled according to the guidelines listed in j

Table 2-1.

These guidelines were developed specifically for this pro-gram. Occupation whole body dose limits may be permitted to exceed 2-3 l

Table 2-1 administrative guidelines provided that an approval has been signed by the RC&SO and the Reactor Health Physicist. Under no circum-stances will the limits In 10 CFR 20, "Standards for Protection Aqainst Radiation," listed in Table 2-2, be exceeded.

Visitors and non-radiation workers will be limited to occupation whole body doses not to exceed 100 millirem per week, 150 millirem per calendar quarter, and 300 millirem per calendar year (whichever is more restric-tive in the individual case).

Visitors, non-radiation workers, and radiation workers who have failed to meet or maintain the requirements as a radiation worker will be escorted by a radiation worker, according to requirements specified in standard procedures, whenever they enter the DECON area.

To assure compliance with ALARA principles, the Executive Engineer's Radiological Engineer / Health Physics Supervisor will be available to review the work permits and assist in preparation of work permits.

An ALARA checklist will be used to assure that all work is preplanned to minimize radiation exposure. The checklist includes the physical and udministrative implementation of the radiation exposure controls.

The implementation of the ALARA philosophy is directed by the RC&SO.

The RC&SO and the RHP review and approve procedures concerning work that has the potential for occupational exposure. Appropriate health physics procedures will be referenced in the Work Procedures to ensure that any occupational exposure is maintained ALARA.

Entrance to the restricted areas of the facility will be controlled by the BRR RHP and requires the issuance and a, proval of a Work Permit and Radiation / Hazardous Work o

F'ermits (RHWPs).

The description of the RHWP system is discussed below.

When the Permit is initiated, the work assignment and applicable pro-cedures are developed and listed., As determined on a case by-case basis, additional health physics procedures can be impicmented at that time, if needed.

Public radiation exposure resuiting from dea.ormissioning the BRR must conply with 10 CFR 20. The maximam public exposure limits for external exposure are specified in 10 CFP 20.105, "Permissible Levels of Radia-tion in Unrestricted Areas."

Limits for internal exposure pathways are given in 10 CFR 20.106, "Radioactivity in Effluents to Unrestricted Areas." As in the case of occupational exposure, 10 CFR 20.1(c) requires application of the ALARA principle to the control of public radiation exposures and releases of radioactive materials to the envi-rons. Appendix I of 10 CFR 50 provides numerical guidelines for estab-lishing design objectives and limiting conditions of operation to meet the ALARA criterion for radioactive materials in effluents from operat-ing light-water reactors.

The EPA public radiation exposure limits, defined in 40 CFR 190, "Envi-ronmental Radiation Protection Standards for Nuclear Power Operations,"

are now in effect. As currently written, the EPA limits apply to uranium fuel-cycle operations that directly support the production of electric-ity, but not to test reactors nor for waste management.

Limits for waste management are being developed and may apply to decommissioning.

2-4

Table 21 Administrative Guidelines for Radiation Whole Body Doses During Decommissioning Administrative Guidelines (mrem)

Non Radiation Workers Radiation Workers and Visitors Daily 20 100 Weekly 100 300 Calendar Quarter 150 500 Calendar Year 300 1,000 Table 2 2 Regulatory Limits for Radiation Doses During Decommissioning for a Calendar Quarter (mrem)

Fladiatica tNorkers Whole Body, Gonads, 1,250 Blood Forning Organs, Lens of Eye Hands and Forearms, 18,750 Feet and Ankles Skin of the Whole 7,500 Body O

2-5

In effect, the EPA limits, which are more restrictive for direct external exposure than those in 10 CFR 20.105, will govern all aspects of public radiation exposure.

(The appropriate sections of 10 CFR 20 are being revised to reflect this.)

However, since Appendix I of 10 CFR 50 is more restrictive than 40 CFR 190 for internal exposure from light-water reactor effluents, Appendix I should guide this aspect for the BRR decommissioning.

2.1.3.2 Radiation Hazardous Work Permits RHWPs will be established to assure that hazardous conditions and pro-tection measures are identified and communicated to those who sill per-form work in potentially hazardous areas or with material which may be radioactive er radioactively contaminated.

The procedures will also provide the mechanism for exposure accountability and ALARA assessment.

A RHWP will be issued for a specific task, job, or series of tasks to be performed within restricted areas; with material which is radioactive, radioactively contaminated, or chemically hazardous; or with high-hazard operation (confined space, platforms, crane operation, etc.).

RHWPs will be used for areas where hazards are significant and may change.

RHWPs will establish:

Radiation and contamination mapping based upon radiological o

surveys, analytical results, and calculations Segregation of the available work area into sections (e.g.,

o contaminated, "clean," working area, examination area) o Access limitation to cortrol the sp"ecd of contamination from conterrinated to "cleaning, areas and limit access to all per-sannel who are not directly involved in the specific task o

Description of the methods to identify and mark all removed items, and note their place of origin and any other pertinent radiological informatior, Par.kaging of contaminated wastes in appropriate containers (as o

prescribed by NRC and 00T regulations and radwaste disposal site criteria)

Maintenance of accurate shipping records throughout the o

operation Listing of work area monitoring requirements which warn of any o

unexpected changes in the radiological conditions o

Listing of personnel monitoring and protective devices o

Requirocents for maintaining accurate and updated records of personnel exposure, surveys, and lessons learned in order to improve and revise procedures as necessary 9

2-6

h The supervisor, of workers who perform work under RHWPs, will be respon-V sible for assuring that the workers have been properly prepared prior to entry into a restricted area.

Proper preparations include:

o The workers have successfully completed Radiation Worker Training and OSHA Hazards Communication Training and Respira-tory Protection Training (if applicable) o The workers' dose records will be checked to ensure entry and/or work without exceeding established limits (administra-tive and regulatory) o Pre-job ALARA briefings, training, or instruction have been provided if recommended or required o

Approved detailed procedures covering the total radioactive work aspects have been prepared prior to the start of work o

Appropriate procedures, tools, and equipment are available to perform the job o

All workers and supervisors of such workers have read and understand the RHWP and its requirements, as well as the work conditions and special controls necessary The RC&S0 will be responsible for assuring that personnel have been O

properly prepared for entry before approving their entry to restricted areas.

The contractor, health physics supervisct, the work supervisor, and the workers responsible for performing the work will assure that all radiation / hazard controls are properly impler.ented thioggbout the job cycle.

RHWPs will be valid orly for the period of the task (s) to be performed and only for the specific task (s) indicated oa the RHWPs.

RHWPs are provided for entry and work in areas where radiological conditions are subject to significant or unexpected change; therefore, additional ir,-

structions or requirements will be incorporated in the RHWPs, as changes warrant, through the fo!1owing nrocedures:

o Any supervisor responsible for completion of a task to be per-formed within the Restricted Area requiring a RHWP may request one by completing the applicable portions of Radiation Hazard-ous Work Permit Request Form and forwarding it to the RC&SO and BRR RHP.

Requests for RHWPs should be submitted a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to scheduled job initiation, o

The RC&SO and the RHP will review the request and prepare the RHWP, after determining the following:

The radiological status of the work area, through the appropriate contractor monitoring and surveys.

O V

The necessary precautions, based on the radiological i

l status, including protective equipment, special control l

measures, and actions to reduce exposure to ALARA.

2-7 L

Each RHWP anall be 6 signed a number indicating the year and month of issuance, respectively, and the total number of RHWPs issued that month.

Example:

for the first RHWP issued in November 1988, the RHWP number would be 88-11-1.

o All RHWPs will be reviewed and approved by the RC&SO and RHP.

The work supervisor should sign all RHWPs, indicating cogni-zance of the work to be performed, the work location and conditions / restrictions, and approval to enter and perform the

work, o

Copy 1 (original) of approved RHWPs should be posted at the entrance to the job location area to allow review immediately prior to entry to the worksite.

Copy 2 should be posted near the entrance to the Restricted Area and copy 3 should be re-tained by the RC&SO.

Copy 4 of the RHWP should be retained by the Work Supervisor.

o Exposure time sheets should be provided for each RHWP and all persons should provide the appropriate information requested on the time sheet upon entry and exit from the work location.

o All information entered on exposure time sheets should be printed legibly in ink.

Information entered on RHWP time sheets indicates that the individual for whom information is entered has read, understands, and will comply with the requirements of the RHWP, and that entry and work will be in accordance with the established radiation protection rules and

policy, o

Upon job completior, the individual (s) responsible for per-forming the werk should inspect the work area to assare thht it is clear of materials, tools, equipment, or other ite:ns used or produced by performance of the job.

The individual (s) resporsible for the work should assure that the work area is in a condition equal to or better than that at the commence-ment of the job, and the responsible supervisor should sign the back of copy 1 of the RHWP posted at the jobsite, request-ing termination of the permit.

o The RHWP termination request should be forwarded to the RC&SO and RHP who will verify the condition of the jobsite and sign approval to terminate the RHWP.

o All copies of the RHWP should then be collected.

The original (Copy 1) and Copy 2 should be retained as file copies by the RHP and RC&SO, respectively.

The copies should have the ter-mination date recorded on them.

All other copies should be discarded.

O 2-8

2.1.3.3 ControIIed Surface Contamination Area

.,i r Contaminated or potentially contaminated items, materials, and surfaces will be handled, dismantled, and decontaminated within a Controlled Surface Contamination Area.(CSCA). Radicactive waste material will be placed in designated containers and stored in radioactive material stor-age zones. To minimize areas designated as CSCAs and the potential that contamination will be spread throughout these areas, small CSCAs will be established to promote work efficiency.

These CSCAs may correspond to locations where cutting, dismantling, and decontamination operations are performed.

When materials with loose surface contamination are properly wrapped and carefully handled to pre-vent breaking the wrapping, they may be carried through or handled in areas which are not controlled for surface contamination.

All work involving contaminated material will be performed inside the boundaries of a CSCA.

A Contamination Control Point (CCP), through which all entries and exits will be made, will be located on the perimeter of each CSCA.

The floor of the CCP will be covered with paper, plastic sheet, or other material.

This is to provide an easily removable surface within the CCP to prevent the spread of contamination from the area. A "step-off" pad will be placed at the exit of the CCP.

This will be used when removing clothing during exit from the area.

Receptacles for waste and contaminated clothing will be maintained at the CCP.

Instrumer.ts for monitoring personnel and equipment will be cr, hand. All equipment, parts, materials, surfaces, and wastes which have been ex-posed to radioactive contamination or to neutrons from the reactor or sources will be handled as radioactive and will not be released for unrestricted handling until they are surveyed and show results in com-pliance with NRC Regulatory Guide 1.86.

If loose contamination is sus-pected to be in excest of the limits prescribed above on surfaces not accessible for neasurement, the traterial will be handled as radioactive.

The actual frisking will be performed wherever possible in low radiation background areas where audible response of the frisker can be distin-guished more easily. Adequately trained personnel will be permitted to frisk themselves.

l Radiation tags and labels will be available at the CCP to identify the l

contaminated items being removed from the area.

The entrance to the contaminated area will be posted to provide:

l o

Approved RHWP i

o Information concerning radiation and contamination levels o

Precautions for entry O

2-9

o Precautions for exit o

Step-cff points o

Frisking instructions The RHP or a designee will be responsible for the CCP and will ensure that personnel and equipment are surveyed and logged as required. Work involving contaminated material inside the CSCA may require the use of glove boxes and enclosures.

The installation, use, and dismantlement of these items will be supervised by the Executive Engineers' health physics personnel.

2.1.4 Radiation Protection Facilities, instrumentation, and Personal Protective Equipment Radiation protection facilities, instrumentation and personal protective equipment used during DECON are discussed below.

2.1.4.1 Facilities Facilities may be provided to enhance the effectiveness of the ORPPs to include the following:

o Facilities and equipment to clean, repair, and decontaminate personal protective equipment, monitoring instruments, tools, and other materiel o

Change areas which allow for the segregation of contaminated frcm non-contaminated clothing o

Control stations fur entrance or exit af personnel into radia-tion or contaminated areas, fct rrovetent of radiocctive waste material, and for movement of potentially contaminated equip-ment and instruments o

Eoaip:nt nt to facilitate ccmmunication between workers and supersiso y pereennel between radiation and non-radiation areas o

Calibration facilities for the instruments which will be used during decommissioning It is intended to make maximum use of the facilities and personnel already existing at the University of California to minimize the cost impact and to take advantage of the University staff experience.

Coordination with city services, such as those provided by the Fire Department and Emergency Ambulance, will be requested through the UC police.

Procedures will then be developed should events requiring these services arise during the decommissioning activities.

The University will arrange to provide care of injured workers, if needed, with Cowell Hospital, or Herrick Hospital.

2-10

2.1.4.2 Instrumentation (n1 A wide range of portable and non-portable instruments and lab-counting equipment will be used during the decommissioning for radiation surveys, radioactive contamination surveys, personnel monitoring, area monitor-ing, air monitoring, and sample analysis.

Table 2-3 lists types of instruments / equipment in the Nuclear Engineering Department at BRR or part of the University Office of Environmental Health and Safety (EH&S) that :an be made available for decommissioning activities described in this plan. The Executive Engineer may supplement the instruments and equipment with their ow' Portable and non-portable radiation protection instruments supplied by the Executive Engineer will be calibrated in accordance with established guides and practices.

Detailed calibration records (including date, method, source description, results, and person) will be kept as quality assurance records and will be auditable under a quality assurance program.

On a daily basis, or as frequently as required, each type of instrumen-tation will be checked to verify that it is properly functioning.

Table 2 3 Radiation Survey and Monitoring instrumentation and Equipment Available for Decommissioning Activities o

Portable Ion Chamber Rate Meters o

Portable GM Survey Meters o

Alpha Survey Meters o

Neutron Surycy Meters o

Pocket Ion Chamber 00simeters o

Single Channel Analyzers with Scintillation and Pancake GM Probes o

Area Monitors o

Windowless Gas Flow GM Counting Systems o

Liquid Scintillation Counter System o

Hand and Foot Monitor o

Pressurized Ion Chamber (to be obtained)

O 2-11

2.1.4.3 Personal Protective Equipment Other personal protective equipment will be provided by the Executive Engineer / contractor for use as needed. Typically, such equipment includes:

o Anti-contamination clothing o

Contamination control equipment, such as hoods, plastic con-tainers, bags, filters o

Signs, labels, tags o

Special tools o

Decontamination equipment o

'tobile or temporary shields o

Respiratory protection devices Respiratory Protection Program: A respiratory protection program in compliance with ANSI Z-88.2, NRC Regulatory Guide 8.15, and Cal OSHA will be developed by the Executive Engineer / contractor to provide pro-tection against airborne radioactive and/or chemically hazardous sub-stances. The following elements are included in the program:

o Written standard operating procedures governing selection and use of respirators o

Assigrment of responsibilities o

Types or records o

Training of employees and supervisors o

Qualitative testing o

Work area surveillance o

Medical surveillance o

Special respirator use problems and limitations o

Maintenance end repair of respirators 1

2 The Executive Engineer / contractor will supply only MSHA /NIOSH approved respiratcry protection equipment.

Potential exposure hazards, radio-active particulates and vapors, and airborne chemical waste will be monitored and evaluated.

Workers will be instructed and supervised to O

1 Mine Safety and Health Administration 2 National Institute for Occupational Safety and Health 2-12

. (~s ensure that all respiratory protection equipment provided is used in Q

accordance with the training and instructions received. The contractor will routinely inspect respiratory equipment, and protect it from out-side contamination and damage.

The RC&SO and Certified Industrial Hygienist (CIH) will direct, eval-uate, and provide guidance on all aspects of the facility respiratory protection program and ensure that employees required to wear respira-tory protection are physically and medically able to do so.

Immediate facility program administration will be accomplished under the guidance and direction of the RC&S0.

In addition, the RC&S0 will be responsible for maintaining an adequate supply of respirators and car-tridges on-site for personnel use. All purchasing of respiratory equip-ment will be under the guidance of the RC&SO.

Respiratory selection will be based on the following criteria:

o Nature of the hazard o

Physical properties of the contaminants involved o

Location of the hazard o

Time frame for which respiratory protection will be required o

Operational activities of personnel required to wear respira-tory equipment o

Functional capabilities and limitations of respiratory equipment o

Potential for the presence of conditions immediately dangerous to life and health o

Potential for the presence of oxygen deficient atmosphe es All facility personnel required to use respiratory protection will re-ceive periodic training portaining to all respiratory protection. This training will be given under the guidance of the RC&S0.

Training will include but not be limited to the following:

Proper use of all available respiratory protection including o

hands-on training o

Reasons for the selection of a particular type of respiratory equipment based on potential hazards o

Functional capabilities and limitations of all available respiratory equipment A

o Proper methods of donning respiratory protection equipment o

Reasons for determining respirator fit, methods to be used, and factors affecting respiratory fit 2-13 c

o Proper care and maintenance of respiratory protection o

Training in the use of respiratory protection as it relates to the recognition and handling of emergency situations o

Discussions of potential contaminants against which the wearer is to be protected, Including physical properties, physiolog-ical action, toxicity, and means of detection o

Discussion of the application of various cart-idges and canis-ters available for air respiration o

Instruction in emergency action to be taken in the event of malfunction of the respiratory protection devices All personnel required to use respiratory protection will be advised that they may leave the work area at any time for relief from physical or physiological distress, procedural or communication failure, signifi-cant deterioration of operational conditions, or any other condition that micht require such relief.

The RC&S0 will designate which personnel are qualified to give instruc-tions on respiratory protection.

Qualifications will be based on the designee's knowledge of the application and use of respiratory protec-tion and the hazards associated with potential chemical and radioactive contaminants.

All BRR decommissioning personnel required to wear respiratory protec-tion and who have demonstrated that they are physically and medically able to do so will receive a qualitative respirator fit test to be administered under the guidance and direction of the RC&S0 or CIH.

The progran will include enp!oyce orientation, employee seas;'tisity tests, performance of the fit test in a test enclosure, and respirator as sigr;nent.

lhe qualitative fit test protocol will use Iscamyl Acetate or Irritant Smoke tubes as apprcpriate.

In additior., all p3rsonnel will be instructed in the proper procedure for the performance of the positive and negative pressure tests.

These quick respirator fit checks will be performed by all personnel immedi-ately after donning approved respirators and prior to entering an area designated for respirator use.

Personnel will only wear air purifying respiratory equipment for which they have successfully passed the qualitative fit test protocol.

No respirators will be worn by personnel who have facial hair such as beards or long sideburns which interfere with the sealing periphery of tle respirator face piece or with respirator valve function.

Contact lenses will not be worn by any employee while in an area desig-nated for respirator use.

Prescription glasses may be worn as long as the seal of the respiratory face piece to face is not directly affected.

O 2-14

Only the RC&S0 or a properly trained designee will be permitted to issue

(_)

respiratory protection to facility personnel or outside contractors and V

visitors.

All outside contractors and visitors will be required to adhere to the same respiratory protection procedures as regular decommissioning personnel.

For unusual operational instance or special projects, respirator issuance will be made under the guidance and direction of the RC&S0.

All workt i for whom the potential of contact with hazardous materials exists shall participate in a medical surveillance program.

As a mini-mum this program must provide baseline health assessments to investigate existing conditions that may predispose a worker to illness following exposure to hazardous substances or to the physical demands of using protective equipment.

In addition, periodic health assessments must be provided to screen workers for signs of occupational exposure to toxic agents and to determine their subsequent assignments.

2.2 INDUSTRIAL SAFETY AND HYGIENE PROGRAM The Industrial Safety and Hygiene Program (ISHP) for the decommissioning project is concerned with the protection of all personnel occupying Etcheverry Hall from detrimental non-radioactive exposures and hazards.

The ISHP will be administered in accordance with Cal OSHA regulations.

('

In the absence of a particular regulation, guidance will be obtained from t: ' NIOSH or the American Conference of Governmental Industrial

'(~

Hygienists (ACGIH).

Chapter 7 of the DP discusses the detailed specifications of the health and safety limits and their implementation for workers and the members of the public.

2.2.1 Personno!

The ISHP will be admint.tered by the Executive Engineer's RC&SO.

In addition to responsibilities previously discussed as aspects of the RPP, responsibilities will include:

o Inspections and audits o

Occupational health Medical surveillance Hearing conservation First aid o

Emergency services i

o Operational activities (v3 1

2-15

2.2.2 Training O

To supplement the comprehensive training program described in Section 2.1.2, the Executive Engineer / contractor will supply all workers with instruction concerning the project safety program through orientation /

training. Each new-hire or transferee will attend one of these orienta-tions, which consists of instruction in job safety action plans, hazard recognition and correction, fire extinguisher training, and safety awareness films. Specialized training applicable to specific conditions will be given as the progress of decommissioning activities mandates.

Supervisory safety training is an integral part of the safety training program. Supervisors will receive a safety orientation detaiiing the safety responsibilities of their positions.

Training courses and a qualified staff roster will be documented and updated, with followup training conducted as needed. Topics for presen-tation at these ISHP training sessions include:

o Specific project safety procedures o

Fire protection and prevention o

Work practice procedures and tool-box safety o

Special housekeeping requirements o

Material handling techniques o

Safety and warning devices o

Hazard identification and reduction o

Enforcement policy o

First aid anc emergency procedures and equipment Records will be kept of all persennel attending, level of accomplish-ment, follow up sessions, etc., as necessary, to ensure that the appro-priate awareness and competency have been demonstrated.

2.2.3 Administrative and Work Practice Controls These controls comprise the measures taken to limit chemical exposure and safety hazards and to reduce the risks associated with the decommis-sioning activities.

Essential to the ISHP are RHWPs discussed in Sec-tion 2.1.3.2 and access controls discussed in Section 2.1.3.3.

2.2.3.1 Exposure Limits Personnel exposures to toxic / hazardous materials will not exceed limits established by Cal OSHA or those recommended by ACG1H in "Threshold Limit Values and Biological Exposure Indices for 1987-1988."

2-16

2.2.3.2 Inspection and Audit Programs Inspections will be conducted routinely by the Executive Engineer /

contractor during active work phases and on at least a weekly basis.

Safety. violations will be recorded, identified, and corrective actions then taken immediately.

Copies of any infraction notices will be main-tained by the RC&SO and documented on a Weekly Safety Report to the UC Decommissioning Project Engineer. The following items will be specifi-cally addressed during these routine inspections:

o Barricades i

o Safety signs o

Scaffolds o

Confined-space entry o

Floor and roof entry o

Hearing protection o

Radiation / Hazardous Work Permits Supervisors will be required to participate actively in the investiga-tion of any accident occurring in their area which results in any per-sonal injury to employees under their direction, equipment or property damage, and near misses with the potential for serious injury or loss.

The investigation should be aimed at determining facts, not fault, so that recurrences can be prevented.

In order to provide verification of the program, an audit procedure of the ISHP will be developed, incorporating approved evaluation criteria.

Audits are conducted for:

o Compliance with all saff.ty requirements o

Implementation of health and safety procedures i

o Health and safety organization o

Job descriptions and tasks i

o Review of records and documenta; ion relating to health and safety o

Site layout and inventory o

Training materials f

O 2-17

. ~,

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

The audit criteria includes, at a minimum, the evaluation of:

e Written procedures o

Qualifications, education, and training of management and staff o

Communications and coordination of the various health, safety, and medical elements of the program o

Environmental surveillance o

Facilities, apparatus, and monitoring equipment o

Medical surveillance o

Emergency planriing Using this procedure, an assessment is completed and the results prop-erly communicated to the UC Decommissioning Project Engineer, the RC&SO, and the contractor.

Written recommendations are, prepared to improve deficient areas.

It is the responsibility of the UC Decommissioning Project Engineer, in con-junction with the RC&SO, to ensure correction of deficiencies and docu-mentation of actions taken.

2.2.3.3 Medical Surveillance Program A medical surveillance program will be established by the Executive Engineer / contractor for all workers who may be occupationally exposed to radiological or hazardous chemical agents.

Th2 program may include the following items as appropriate in baseline health assessments:

o Occupational history o

Medical history o

Family history o

Physical examination o

Pulmonary function testing o

Audiometric testing o

Baseline bioassay Criteria for this assessment will be developed to identify the need for pre-employment and periodic health assessments, termination examina-tions, and return-to-work and other special examinations.

h Consideration will be given to on-site facilities and equipment and to access to local clinics and hospitals for non-routine and emergency treatment.

2-18

Medical records will be maintained with attention given to federal

,s f

T O

requirements, inclusion of exposure data, appropriate update frequency, and access privileges.

The worksite will be monitored for health hazards associatec with the work environment, including chemicals that may be present in liquid, dust, fume, mist, vapor, or gaseous forms.

Physical hazards such as noise, pressure, vibration, and illumination will also ba monitored and controlled.

2.2.3.4 Hearing Conservation Program A hearing conservation program will be established by the Executive Engineer / contractor for all workers who are exposed to noise levels of 85 dBA or greater (as an 8-hour, time-weighted average exposure).

This program will include:

o Noise monitoring in areas where the levels exceed 80 dBA Audiometric testing for all workers (as part of the medical o

surveillance program) to determine baseline hearing perfor-mance before exposure and test results after exposure o

Personnel training and education o

Recordkeeping o

Hearing protection devices Personnel who are assigned tasks in known noise-hazardous areas

(> 90 dBA) will be enrolled in the hearing conservattor, program prior to beginning their work. Other personnel may be required to participate after monitorirg reveals thtt their 8-hour, time weighted average expc-sure exceeds 85 dBA.

i Noise control measures, Including the requirement to wear hearing pro-tectinn equipment, will be determined by the P,C&S0 after appropriate noise monitoring is completed throughout Etcheverry Hall and the sur-rounding area.

Records will be maintained that document all noise monitoring conducted, employee training done, control measures implemented, and protective I

equipment issued.

2.2.4 Operational Activities The requirements for fire protection and other equipment / tools are dis-cussed below.

2.2.4.1 Fire Protection and Prevention Fire protection devices will be made available by the contractor in the O

Reactor Room during decommissioning tasks.

Portable Type A and B/C fire extinguishers will be strategically located in Room 1140 to serve areas partitioned for the various decommiss bning activities.

2-19

.,.. _ - - - _ _ - - - _ ~ -..__

Fire prevention measures will be implemented to avoid ignition hazards from electrical wiring and equipment and from conbustible materials.

Smoking will not be permitted in areas where a potential fire hazard is present.

No..u ing will be permitted in the Reactor Room.

2.2.4.2 Hand and Power Tools and Cutting Equipment The condition of the hand and power tools used during decommissioning activities will be routinely checked for proper operation and for com-pliance with the applicable provisions of the Cal OSHA regulations cited in Section 1.6.1.1.

2.2.4.3 Lifting Equipment Lifting equipment (fuel handling monorail hoist, fork lift, and work bridge crane) used in the decommissioning activities will comply with the applicable provisions of Cal OSHA.

These provisions include:

o Compliance with the manufacturer's specifications and limita-tions applicable to equipment operation o

Posting of rated load capacities, operating speeds, and special hazard warnings or instructions o

Inspection of equipment by competent oersonnel prior to each use and during use to make sure it is in safe operating coridition o

Limiting the travel of rail-mounted equipment with limit stops o

Removing from service any equipment which has damaged wire ropes 2.2.5 Personal Protective Measures To minimize the effects of the hazards associated wit'n decommissicning, specific health and safety measures will be impiemented.

These are listed by hazard in Table 2-4.

O 2-20

Table 2 4 O

Mitigation ;nd Monitoring of Hazards During Decommissioning r

Hazard Mitigation Monitoring airborne dust /

radionuclides o water fogger o whole-body counts o respirators o continuous air sampling o HEPA filtration units o grab air sampling falling and flying debris o safety glasses o incident reports o limited access, safety shoes o hard hats high sound levels o ear protectors o physical exams o dB measurements beta-gamma exposure rates o limited access o personnel dosimetry O

o daily surveys high heat and humidity o ventilation o temperature measurements o work breaks o stay time limits loose surface contamination o anti-contamination o frisking clothing o daily surveys o cleanup / decontamination airborne hazardous i

chemicals / vapors o respirators o air sampling o cleanup / decontamination t

l O

2-21 l

2.2.6 Excavations Excavations required during decommissioning activities will comply with applicable provisions of Cal OSHA.

These provisions include:

o Protection of workers with personal protective devices as discussed in Section 2.2 above o

Provisions to prevent workers from standing under loads handled by lifting equipment o

Daily inspection of excavations by contractor industrial safety personnel for evidence of cave-ins or slides o

Supporting systems (e.g., underpinning, etc.) designed by qualified contractor personnel and inspected daily o

Excavated materials and other material stored at least 2 ft from the edge of the excavation o

When using heavy equipment in the vicinity of excavation, the sides of the excavation braced to resist extra pressure by superimposed loads o

Adequate barrier physical protection provided around the excavated area o

Ladders and scaffolding used in excavrtion provided around the excavated area o

Ladders and scaffolding used in excavation complying with the applicable provisions of Cal OSHA 2.3 EXECUTIVE ENGINEER: CONTRACTOR ASS; STANCE Outside contractors will be used to supplement the University's own staff for the following activities:

o Supervision of day-to-day decoimissioning activities including direction of craft supervisors and crew leaders o

Health physics assistance, including equipment installation, calibration, and testing; and conducting of radiological surveys o

Quality assurance assistance, including procedures preparation and quality compliance during decommissioning o

Crafts and labor to provide temporary construction work, per-form decontamination and demolition tasks, and to process, package, and ship radioactive materials O

2-22

These will be ongoing activities during the entire decommissioning period, O

with the University's personnel overseeing and reviewing the work as it takes place. The University will retain overall responsibility for health and safety during all aspects of decommissioning.

2.4 COST ESTIMATE 2.4.1 Cost Estimate Elements The estimated cost for performing the decommissioning of the BRR is approximately $1,350,000 as shown in Table 2-5.

In general, the work consists of removing the reactor core components and all appurtenant equipment, removing all activated and contaminated materials, and ship-ping the debris to a nuclear waste disposal facility.

The noncontami-nated balance of the reactor pool structure will be demolished together with other non-contaminated items and shipped to a municipal dump or scrap yard. A new finished floor will be installed in the demolished area. The estimate is broken down into the specific tasks, described in detail in Chapter 3.

The basis of the estimate follows.

o Labor and Burden - Union labor rates (1987) for Alameda County were used.

o Equipment Usage Equipment usage includes rental and operat-j ing cost of cons:ruction equipment.

It also includes power tools, scaffolding, and consumables such as drill bits, etc.

Special construction equipment purchased for this project will become the property of the University; hence, no salvage value is included in this estimate.

3 o

Material - Consumab'e materials prices are based on available data and vendor quotes, o

Sales Tax - A 7% state and Alameda County tax is included with material and equipment purchases.

I l

o Subcontract - Subcontract costs were estimated for:

Task 5, removal of material with radioactive concrete Radiation survey and monitoring Medical examinations.

I o

Indirect Costs-Contracter's Overhead and Profit included the following markups:

20% On Labor and Burden 10% On Equipment Usage 10% On Material Purchases 5% On Subcontract and Fees.

Construction Management included a 15% contractor's cost.

2-23 4

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Table 2-5 Estimated Cost for Decommissioning the Berkeley Research Reactor QUANTITY FIELD LABOR LABOR & EQUIP'T CONSUMABLE SUBCONTRACT

& UNIT MANHR RATES BURDEN USAGE MATERIAt

& FEES TOTAL TASK 1-CONTRACTOR MOVE-!N 936 36720 21900 43700 25000 127,320 TASK 2-INITIAL RADIATION SURVEY 9700 9.700 l

IASK 3-INSTAttAT10N Of CONF!NEMtNT BARR!tR$

400 13850 11400 6200 31,450 i

TASK 4-RIMOVAL OF REACTOR COMPONENIS AND POOL LINER 840 29400 4900 2500 21300 58.100 TASK 5-REMOVAL OF MATERIAL WITH POTENTIAL SURFACE CONTAMINATION ANO 2080 154000 154,000 OTHER ACTIVAllD MATERIALS 716 25060 2400 27,460 TASK 6-CLEANUP AND REMOVAL Of TOOLS AND EQUIPMENT 312 10920 1000 269850 281,770 TASK 7-PACKAGING AND SHIPMENT Of RAOlUACTIVE WASTES 30000 30,000 TA5" 8-TERMINAI!ON RADIATION SURVEY 3260 97800 35500 16200 2400 151,900 T ASK 9-Of MOLIIION OF NON-RADIDACIIVE PORTION OF REACIOR INSTAL L All0N 7 S.UBT0TAL (TASKS 1-9) 8544 213750 17100 68600

$12250 871,700 SALES TAX - 7% OF Material Cost 4802 4,802 SUBIOTAL CONTRACTOR'S DIRECT COST 213750 77100 73402 512250 876,502 CONTRACTOR'S OH&P

$343E 7710 7340 25613 94.100 SLBTOTAL DECONMISSIONING COST 267188 84810 80742 537863 970.602 CUNSTRUCTION M/.NAGEMENI

$145,590 SUB10TAL 1.116,193 ESCALATION - 4% of the Total Project Cost less Nuclear Disposal Fee J4,808 TOTAL PROJECT COST (INCLUDING ESCALATION) 1.151,000 C6NTINGENCY - 17% of Total Project Cost (Includir.g Escalation) 195.670 101At l.346,670 1,350,000

('N o

Contingency - A 17% contingency Is recommended and is included

(,)

in this estimate.

o Escalation - Included 4% of total cost except nuclear disposal fee.

2.4.2 Assumptions Assumptions used in this estimate are listed below.

o The University will assign personnel to function as the UC Decommissioning Project Engineer, the UC Deputy Decommission-ing Project Engineer, and the UC Reactor Health Physicist, and the UC Q/A Supervisor, to perform the functions described

herein, o

The contractor will have access to the use of the bridge crane and fuel-handling hoist.

o All spent fuel is already disposed of by others.

o Free power and utilities will be made available to the contractor, o

The toilet adjacent to Room 1140 will be made available to the cont-actor.

The shower facility will be available for emer-p) gency situations only, o

The nuclear waste disposal fee used in this estimate is the 8

anticipated 1988 rate of $36.60/ft.

o Sufficient trailer and parking spacc will ba provided by the Ur.iversity near by.

o The decommissioning will be carried out continuously from start to completion, o

unrestricted access to jobsite is already available.

o Double shift is estimated for Task 5.

\\

/NJ 2-25

CHAPTER 3 O

DISMANTLING AND DECONTAMINATION TASKS AND SCHEDULES

3.0 INTRODUCTION

This chapter discusses the following items in regards to dismantling and decontamination:

the tasks and activities involved, the schecule relat-ing to tasks and activities, and the procedures involved in accomplish-ing these tasks.

3.1 TASKS AND ACTIVITIES Tnis section is divided into two parts:

tasks and activities to be per-formed prior to the actual decontamination and dismantling operations, and tasks and activities to be performed as part of decontamination and dismantling.

Figure 3-1 shows the general plan of the first floor of Etcheverry Hall giving the location of Room 1140 and its relationship to the..'ther por-tions of the buildiiig including DEC0n aperations access prior to the start of DECON.

3.1.1 Prior to D!smantling There are three tasks to be performea prior to the actual dismantling O.

activities:

o Removal a.1d relocation of certain classroom laboratory equip-p ment from Room 1140 o

Removal and shipment of fresh reactor fuel o

Removal and shipment of irradiated reactor fuel 3.1.1.1 Removal and Relocation of Classroom Equipment The University has commissioned a study to evaluate the best means to reuse the laboratory and its research and classroom equipmer,t iccated in Room 1140, for future activities in the post-reactor era.

The plan is to continue using Rnom 1140 to support axisting laboratory instruction and research and new laboratory research.

After the reactor is dis-mantled, the operations with radioactive material in the refurbished laboratory aould be operated under a State of California Radioactive Material License.

The University has elected to tempor:rily clear portions of Room 1140 to provide working space for the dismantlement acti"ities.

Temporary con.

finement barric.s will be erected to protect re, tining equipmcnt and University personnel from potential cortamination.

3.1.1.2 Removal and Ship nent cf Ficsh Rcactor Fuct The nine fresh fuel assemblies will be shipped off-site under the University's current operating license.

3-1

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3-2

3.1.1.3 Removal and Shipment of Irradiated Reactor Fuel The Am-Be neutron source will be removed by University personnel and stored in a shielded storage area for other use prior to the start of fuel removal.

The 111 irradiated reactor fuel assemblies will be removed and shipped from the site. This activity will be carried out under the current oper-ating license. The removal, packaging, and shipment of the irradiated fuel (including disassembly of instrumented fuel) will be accomplished by an outside contractor under the supervision of the University staff.

Current pians are to ship the irradiated fuel to a designated U.S. 00E facility.

The consignee will be the U.S. 00E.

3.1.2 During Dismantling The dismantling activities will be conducted in two phases: demolition and removal of activated or contaminated material; and final demolition and removal of non-activated / contaminated materials.

To minimize disruption to the teaching activities, the first phase of the dismantling activities is planned for the summer recess, May 15 tc September 1, 1988. At the conclusion of the second phase the site will be ready for restoration and new construction as required by the University.

O Eight tasks are planned for the 'trst phase period based on a one shift, eight-hour five-day per week basis.

If necessary to meet the schedule, some activities may be conducted on a two shift per day basis.

Schedule compression may also be achieved by performing some tasks concurrently, These are discussed in Section 3.1.3.

After fuel elements have been removed, the r6maining items of equipment and structure which are either contaminated or activated, and are to be removed under this plan, are grouped into four categories:

Group 1 - Reactor components that have induced activity Group 2 - Reactor equipment or structural components with potential sur-face contamination Group 3 - Reactor tank structure and components that have become acti-vated due to proximity to the core Group 4 - Equipment, systems, or tools which may have become contami-nated during the DECON operations These contaminated or activated components are summarized in Table 3-1, Table 3-2, Table 3-3, and Table 3-4.

O 3-3 I

7 Table 3-1 Reactor Components with Induced Activity Group 1 Items Quantity Rotary Specimen Rack 1

Control Rod Guide Tubes 1

Ion Chamber Assembly 1

Top Grid Plate Assembly 1

Bottom Grid Plate Assembly 1

Core Support Structure 1

Central Thimble 1

Pneumatic Tube (rabbit) 1 assembly TRIGA graphite dummy elements 31 (15 in the core)

Transient Rod 1

Am-Be Neutron Source l 1

Po-Be Neutron Source:

1 i

i l

l l

I I This source will be retained by the University.

2 This source was previously in the reactor core, 3-4

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Table 12 Miscellaneous Components witn Potential Surface Contamination Group 2 3

)

Items Quantity j

Pool water purification system consisting of:

Heat exchanger!

1 Demineralizer 1

l Filter 1

1 3

Circulat:ng pump 1

Conductivity cells 2

i Flow meters 7

t Cooling coils (embedded in concrete shield) 10 l

assemblies Valves!

19 r

Connecting piping 1 lot l

h 4

Core bridge assembly 1

j Rolling shield assembly 2

f i

l Water level indicator 1

i Misc. conduits and cables 1 lot j

Fuel storage rack 1

Fume hood 1

Storage well 1

j Fuel Handling Monorail Hoist 4.5-tonne (5-ton) capacity 1

t 8 May possibly be donated to UC Irvine Campus.

O i

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Table 3 3 Reactor Tank Structure with Induced Activity Group 3 Items Quantity Reactor tank aluminum liner 1 assembly Inner concrete surface near horizontal thermal column (Hohlraum)

-5m (176 cu ft) 8 Beam ports 4

Through ports 2

Horizontal access ports 2

Reactor tank concrete floor at three reactor 3

positions

-3.9m (136 cu ft)

Concrete embedments 1 lot Steel shielding slabs over thermal column 3

Inner concrete surface of exposure room

-73.6m (2600 cu ft) 3 Thermal column graphite and aluminum lining 1 assembly Table 3 4 Contaminated Tools and Equipment Group 4 Items Quantity Tools used in DECON operations 1 lot Solid and liquid radwaste handling and treatment equipment I lot Confinement barriers 1 lot Temporary ventilation system 1 lot Facility ventilation system 1 lot Work clothing 1 lot 3-6

p A ninth task has been planned for the second phase of the work.

This V

task covers the final demolition and removal of the non-activated / con-taminated portions of the structure, and is planned to occur after the NRC has confirmed the final site survey and granted approval for license termination and release for unrestricted use. This task may be carried out directly after approval of license termination, or delayed until the following summer recess in 1989.

Task 9 is included in this OP to demonstrate the complete project approach.

3.1.3 Task Descriptions These task descriptions cover the planned demolition and removal of con-taminated or activated material followed by the demolition and removal of the remaining uncontaminated material.

They are planned to be per-formed in a sequence which will minimize the spread of contamination and provide as low as reasonably achievable (Al. ARA) exposure to operating personnel while ma::imizing public safety.

In addition, the task sequence takes into account the need to minimize the impact on the University's academic activities.

3.1.3.1 Task 1: Contractor Move In p

The first task will consist of the activities by the selected decommis-Q sioning contractor in setting up the office facility and organizing to start work.

The activities include:

o Setting up temporary office facility o

Installing / connecting temporary utilities o

Clearing and securing working areas, installing temporary traffic and pedestrian control barriers outdoors and within Etcheverry Hall o

Completing radiation health and safety training of all employees 3.1.3.2 Task 2: Initlal Radiation Survey Prior to commencement of DECON activities, evaluation of existing radio-logical and hazardous conditions will be conducted by the Executive Engineer. This evaluation will be based on current survey data obtained under the University's survey program and supplemented by additional surveys conducted to meet the following objectives, o

To establish radiation levels throughout Room 1140 and all spaces that are specified to be within the boundary of the NRC licensed facility.

This will permit development of detailed and comprehensive radiation protection procedures.

l 3-7 l

o To confirm the results of analytical computations so that final decisions can be made as to how much activated struc-tural material (concrete, aluminum liner) will be removed, o

To establish "background" radiation levels in other areas of Etcheverry Hall which may be occupied during the decommissioning.

o To establish "background" radiation levels in the immediate outdoor environment so the radiological impact (e.g., contami-nation) of decommissioning activities on the surroundings can be assessed.

3.1.3.3 Task 3: Installation of Confinement Barriers Confinement barriers are to be installed prior to starting decommission-ing activities. They will assist in confining dust and potential air-borne radioactive particulate within the working area, controlling move-ment of personnel and materials, and provide some noise attenuation.

Three temporary barriers will be erected:

two to divide Room 1140 between the DECON area and the occupied laboratory areas, and one to be erected over the reactor pit.

These are shown on Figure 3-1 and Figure 3-2 respectively.

The barrier shown on Figure 3-1, which isolates the southern side of Room 1140, will consist of a solid wall approximately 3 m (10 ft) high extending in an east-west direction. Sloping supports connecting the top of the temporary barrier to the south wall will encase the existing room ventilation air inlets within the enclosure formed.

The roof structure will be covered with plastic film to exclude dust and cause a slight positive air pressure within the laboratory space relative to the DECON area. The solid part of the barrier will be provided with a personnel door to permit construction personnel entry to the DECON area via the change room trailer.

A barrier of similar construction isolates the laboratory equipment in the northwestern corner of Room 1140 from the DECON area.

The third barrier, as shown on Figure 3-2, consists of a structure span-ning the reactor tank area and covered with plastic film to control con-l tamination spread.

This will allow a slight negative pressure created l

by the exhaust fan equipped with a high efficiency particulate (HEPA) l filter to exist within the reactor tank.

The planned activities include:

o Erecting temporary confinement barriers in Roou 1140 and over the top of the reactor tank (see Figure 3-1 and 3-2) o Connecting temporary portable exhaust fan-HEPA filter unit to the reactor tank and test (Figure 3-2) e Inspecting and testing existing HEPA room air ventilation exhaust system in Room 1140 (see Figure 3-3) o Installing and testing temporary room air radiation monitoring and alarm system 3-8

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o Inspecting and testing existing 4.5-tonne (5-ton) bridge crane O

spanning Room 1140 and the 4.5-tonne (5-ton) monorail fuel handling hoist over the teactor.

3.1.3.4 Task 4: Removal of Reactor Components and Pool Liner (Groups 1 and part of 3)

After all preliminary work has been completed (Tasks 1 through 3), the next task will be the. removal of all activated reactor components includ-ing top and bottom grid plates, core support, rotary specimen rack and drive, ccatrol rod guide tubes, graphite dummy elements, fission cham-bers, and miscellaneous fastener hardware.

The pool aluminum liner is also removed in this tank.

3.1.3.S Task 5: Removal of Material with Potential Surface Contamination and Other Activated Materials (Group 2 and part of Group 3)

Schedule constraints may dictate that some of the activities listed below be performed in parallel with those of Task 4.

This may lead to potential cross-contamination between such activities. To minimize this potential, work sequences will be planned in detail and protective measures will be taken.

Major activities associated with this task are:

o Removal of activated concrete and embedments (Group 3 less pocl liner) o Component disassembly o

Surface decontamination o

Packaging for disposal or reuse 3.1.3.6 Task 6: Cleanup and Removal of Tools and Equipment After completion of the decontamination operations described in Tasks 3 1

through 5, Room 1140 and any existing laboratory equipment that is con-i taminated will be cleaned up in preparation for the shipment of all low-level radioactive wastes. The cleanup will be to levels low enough that state regulatory control is not required. The room ventilation air exhaust HEPA filters will be removed and new filters installed. Used filters will be shipped as low level waste, if contaminated.

The fuel storage well and the pool water demineralizer system will remain operable to assist in this task. The fuel storage well will be used as a wash down water holding tank. Prior to discharge to the sewer system, wash down water will be demineralized to remove any residual contamination to a level sufficient to meet state and local discharge standards.

O i

3-11 4

~~-

3.1.3.7 Task 7: Packaging and Shipment of Radioactive Waste Ouring the decontamination procedures outlined in Tasks 3 through 6, low level wastes will be packaged as they accumulate and stored within the storage area of Room 1140. A detailed discussion of the packaging and shipping of radioactive wastes is given in Chapter 6.

3.1.3.8 Task 8: Pcrform Termination Radiation Survey Upon completion of Tasks 2 through 7 a termination radiation survey will be conducted of all spaces that are specified to be within the boundary of the NRC licensed facility to demonstrate that surface contamination levels and ambient radiation levt:ls are not in excess of applicable limits.

Details of the praposed termination radiation survey plan are given in Chapter 8 of this document.

3.1.3.9 Task 9:

Demolition of Non Activated Contaminated Portion of Reactor Installation Upon satisfactory completion and confirmation of the radiation survey, and receipt of approval for termination of license and unrestricted use, the remaining non-radioactive portions of the reactor installation, which includes concrete, steel walkways and stairs, miscellaneous wiring, and conduit, will be demolished by conventional means.

This task also includes obtaining the necessary dumping permits and identification of the disposal area.

The major activities will include:

o Removal of reactor control console, conduit, and wiring.

o Removal of structural steel walkways and platforms and handrails, o

Removal of 4.5-tonne (5-ton) fuel handling monorail hoist, if found not contaminated, o

Dismantlement of exposure room plug door and thermal column plug door and demolition of concrete portions, o

Demolition of concrete shield walls.

o Shipment of all removed materials off-site to a local dump and/or scrap yard, o

filling excavated reactor floor area and water pit with con-crete and finishing level with existing finished floor.

The 4.5-tonne (5-ton) bridge crane spanning Room 1140 will remain in place for future use.

3.2 SCHEf;ULE The master schedule for implementing the decommissioning plan for the BRR is shown on Figure 3-4.

3-12

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RECElvE NRC COMMENTS. PUBLISH FINAL ER EEE; (9Ja FAL C@L D1FiRIDO r;o S. DEFUELING AND SH]PMENT ll llll lllll ll ll lllllllllll'lllll NEMI A

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DCE/INEL SHIPPING ALCjEST REv!EW D/

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CONTRACTOR MOVE-IN fTASK 1) l 8.

INITIAt, RA0]ATION SUh(EY (TA$K 2) 9.

INST ALLATION OF cot #1NEMENT BARRIERS (TASK 3)

10. REMOVAL OF REACTOR COMPONENTS AND POOL LINER tTASK 4)

11. AEM0 VAL OF MATERIAL WITH POTENTIAL SURFACE CONTAMINATION AND OTHER ACTUATE 0 MATEAtAt tTASK S)

12. CLEANUP AND REMOVAL OF TOOLS AND EQUIPMENT (TASK 66

13. PACKAGING AND SHIPMENT OF PR0!OACT!vE WASTE tTASK 7) 14 PERFORM TERMINATION FA0!ATION SURVEY tTASr 8) 15 NRC FACILITY INSPECTION

16. OEMOL1 T !CN OF NON-PADIOAC 7 !VE PORT!CN OF REACTOR INSTALLATION tTASK 96 NOTE: FIGURtS IN t ) ARE DURAT!ONS OF ACTIVITY IN WCRKING DAYS i

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.4-hEvlEW CYCLE f

- TASa DURAtlON 1

Figure 3 4 Berkeley Research Reactor Decommissioning Schedule 1988 1989 JON JUL AUG SEP OCT Nov DEC JAN FEB MAR

27 3

to I? 24 8

8 15 22 29 5

12 89 26 2

9 16 23 30 7

14 El 29 4

Il it 25 2

9 16 23 30 6

13 20 27 3

14 17 24 3

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The schedule is based on the Task Descriptions provided in section 3.1.3 and the following assumptions:

v o

NRC review cycle for the DP and ER will not exceed 17 weeks.

i o

Fuel elements are shipped to 00E within approximately seven months of notification of intent to ship (Kaiser Engineers letter dated September 15,1987).

3.2.1 Milestones Referring to Figure 3-4, the milestones shown at the bottom of the schedule are defined in Table 3-5.

It is planned to complete the OECON Tasks 1 through 7 during the summer recess, May 15 to September 1, 1988, as these tasks are the most disrup-tive to academic activities. The final radiation survey and the NRC follow-up inspection are planned to occur during the academic year start-ing in September with planned license termination scheduled for November 18, 1988. After termination and release for unrestricted use, the remainder of the non-activated / contaminated portions of the reactor installation may be removed as part of a separate demolition contract.

The summer recess duration is equivalent to 75 working days while Tasks 1 through 7, if done sequentially, require 92 working days.

To meet the 75-day limitation scoe tasks will be performed concurrently and other durations will be reduced by an extended work week or multiple shift operations.

g The time period when final demolition will occur (Task 9) is dependent on the University's schedule for reuse of the site.

It is shown as a continuing follow-on activity but it may be postponed until the summer recess of 1989. This has no impact on the DECON tasks.

3.3 TASKS ANALYSES The dismantling and decontamination tasks are planned in a sequence oriented towards limiting the spread of radioactivity, ensuring person-nel safety and ALARA doses, and providing schedule and cost efficiency.

These tasks are described and analyzed below. Task 1, "Contractor Move-in," does not involve radiation exposure and consequently is not covered in this section.

3.3.1 Task 2: Initial Radiation Survey The initial decommissioning activity 111 be a comprehensive radiation stfrvey of the reactor room, other areas of the Etcheverry Hall building, and the immediate outdoor environment of the building as described in Section 3.1.3.2.

In order to achieve the objectives listed in Section 3.1.3.2, the foi-lowing tasks will be performed by the Executive Engineer / contractor.

Included in the description of each task is a brief discussion of the C

radiological health problems anticipated during the performance of the task.

3-15

Table 3 5 Doce,mmissioning Milestonos Scheduled Completion Milestone Description Date 1.

Submit the DP and ER to NRC for review / approval January 8, 1988 Shut down reactor, begin fuel cooldown period 2.

NRC reviews of OP and ER complete, receive April 22, 1988 coments from NRC 3.

Final DP & ER published by UC May 13, 1988 Summer recess period begins Decommissioning contractor mobili:ation starts 4.

Reactor defueling complete May 20, 1988 OP approved by NRC, dismantling order issued 5.

DECON complete (Tasks 1 through 7)

August 26, 1988 Summer recess ends 6.

Termination Radiation Survey complete, September 30, 1988 report submitted Decommissioning Contractor demobilizes 7.

NRC license terminated November 18, 1988 Demolition of non-activated / contaminated components begins 8.

Demolition complete, site available for reuse December 30, 1988 9

3-16

(

3.3.1.1 Reactor Room Initial Conditions v

o A gross-gamma and surface radiation contamination survey of facilities under NRC licensing jurisdiction listed in Section 1.1.1 shall be performed.

Particular attention will be given to those portions which, based on the computational results, are expected to be highly radioactive.

o Subsequent detailed beta-gamma and contamination surveys of the facilities will be planned and conducted to prevent undue exposure to personnel and to identify and assess "hot spots."

o Removable surface contamination levels will be characterized by wipe surveys.

o Measurements of airborne radioactivity will be made to estab-lish "reference" levels before any decommissioning activity begins.

The direct result of this task will be the generation of representative radiation zone maps which will reflect the radiological status of the licensed facilitier.

The primary risk o' exposure during performance of this task is exposure of personnel conducting the survey to underestimated radiation exposure n

rates.

It is expected that respiratory protection devices will not be V

required during this phase unless measurements of airborne concentrations dictate so.

3.3.1.2 Reactor and Biological Shield Assembly Determination of the extent of neutron-activation of concrete will con-firm thickness of activated concrete to be removed and disposed of as radioactive waste.

The extent of neutron-activated concrete surrounding the reactor has been approximated by calculation and is given in Sec-tion 1.3 of this report.

Activation levels will be further determined by the contractor taking samples of the concrete at various depths by core drilling.

Laboratory analysis and measurement of the samples will allow determination of the activity concentration of gamma and beta emit-ters in the concrete.

During boring of the concrete to extract the samples, dust will be generated which will contain radioactive nuclides.

This necessitates monitoring of airborne concentrations and the use of protective clothing (e.g., coveralls, gloves, boots).

It is expected that respirators will be used to prevent inhalation of airborne particulates.

In addition, the level of induced radioactivity of the remaining reactor structure, beam ports, thermal columns, neutron source stordge tubes in the east wall and the floor of Room 1140, and miscellaneous components will be determined by measurement and labcratory analysis in conjunction with calculati.as provided in Section 1.3 g

U 3-17

3.3.1.3 Etcheverry HallInterior Conditions It is necessary to establish "background" radiat'en levels in the other areas of Etcheverry Hall in order to be able to assess the radiological impact of the decommissioning activities on these areas.

This task of the radiation survey will address two main concerns:

o Radiation shine in areas adjacent to all licensed facilities listed in Section 1.1.1 o

Airborne radioactivity concentration in areas (laboratories, classrooms, and offices) where occupancy may possibly continue during decommissioning operations The radiation survey will be conducted as follows:

o A gross-gamma survey of areas adjacent to Room 1140 will be conducted o

Air sampling of the rooms adjacent to Room 1140 will be conducted These radiation measurements will be performed periodically during the decommissioning operation.

These measurements will demonstrate that radiation exposure and contamination are not presenting an unacceptable hazard to the public and workers.

Since the radiation fields expected to be encountered during this phase of the survey will not be significant, the use of protective clothing and respiratory devices is not foreseen.

3.3.1.4 Etcheverry HaII Exterior Conditions The immediate external areas around Etcheverry Hall will have been moni-tored during the initial radiation survey to establish background con-ditions to permit an assessment of the impact of decommissioning on the surroundings.

o A gross-gamma survey wil) be conducted on the truck access driveway, sidewalk, and loading platform located on the west side of the building, and connecting to the bordering city streets.

This driveway is planned to be used for transporting radioactive waste from the demolition activity.

The Univer-sity's Environmental Surveillance Program will continue during and after DECON activities.

o The survey will also include the east side of the building which consists of a parking lot and patio that are located on the roof of Room 1140.

(Room 1140 is below street grade level.)

o Air sampling of the exhaust duct from Room 1140 will be con-ducted along with a gross gamma survey of the roof area sur-rouncing the exhaust duct.

3-18

l

/'

o Samples taken from the upper 5 to 15 cm (2 to 6 in.) of soil l

will be counted with appropriate instruments to determine the background contamination level in the soil.

7 It is not expected that special safety and/or radiation protection mea-t sures will be necessary during this phase of the survey.

3.3.1.5 Evaluation of Initial Survey Results A comprehensive evaluation will be performed of the results of the ini-l tial survey. This evaluation will assess the impact of radiation and contamination levels on the planning and conduct of DECON activities.

In addition, baseline conditions in adjacent areas will be determined.

3.3.2 Task 3:

Installation of Confinement Barriers

[

A barrier will be installed to contain airborne contaminants generated during demolition, and to prevent their spread in Room 1140 and possibly in the surrounding areas. The confinement barrier will consist of a plastic enclosure on a rigid wooden frame over the top of the reactor pit and around the entrance to the exposure room. Associated with this enclosure will be an independent localized ventilation system, which will ensure a negative pressure with respect to Room 1140 while provid-ing high efficiency filtration of the exhausted air, and a source of clean air supply within the enclosure (see Figure 3-2).

Additional confinement barriers (see Figure 3-1), approximately j

3 m (10 ft) high, will be constructed as described in Section 3.1.3.3.

The free standing walls of the confinement barriers will be of plywood i

on a wooden frame covered with heavy plastic sheets. Wood studs will be installed along the existing vertical wall and a heavy gauge plastic sheet attached on the face. The ceiling of the enclosure will be a wooden frame supported from the floor and covered with heavy gauge plastic sheets. All seams and penetrations will be sealed.

A portable cold / hot change-room trailer will be moved into the eastern-l most corner of Room 1140, formed by the temporary wooden barrier and the truck entryway. The trailer will be arranged to permit entrance on the J

"cold" side through a door in the partition connecting to the corridor leading into the personnel air lock. A door on the opposite side of the trailer will open to the DECON area as shown on figure 3-1.

The change-room trailer will provide a cold locker room, a hot change room, a decontamination shower with holding tank, and a means for receiving contaminated work clothing.

The water from the decontamina-i tion shower will be temporarily stored in a holding tank.

After mont-toring for radiological contamination it will be released into the sani-l tary sewage system if the conditions of 10 CFR 20.303 can be met.

If not the water will be treated to remove the radioactive contamination or l

processed for disposal as low level radioactive waste.

i For entering the OECON area, shown as the cross-hatched area on figure l

3-1, personnel will dress in appropriate work clothes and leave their l

street clothes in the cold change area of the trailer.

Upon exiting the 1

3-19

DECON area, personnel will leave their work clothes, wash and shower in the hot change area of the trailer, and then enter the cold change area to dress in their street clothes.

Because of the short duration, non-permanent, nature of this work site, only minimum toilet facilities will be provided within the change area; toilet facilities will be made available to be used by personnel after they change into street clothes.

There is a permanent decontamination shower located adjacent to Room 1140; however, this will only be used in case of an emergency.

Ventilation within the reactor pit / exposure room confinement barrier will be provided by a mobile air cleaning / exhaust unit which will ex-haust the air from the enclosure.

The unit will be equipped with a roughing filter and a HEPA filter.

The reactor pit / exposure room confinement barrier will be maintained at a negative pressure with respect to the other areas of Room 1140.

Ventiiation for the other areas of Room 1140 will be provided by the existing supply roof exhaust system and fan which is equipped with rough-ing and HEPA filters (see Figure 3-3).

The DECON area will be maintained at a negative pressure (with respect to the surrounding areas of Etcheverry Hall) greater than the pressure maintained in the reactor pit / exposure room.

This will ensure that the air will travel from the non-contaminated area to the increasingly con-taminated areas.

During this task, modification of the lighting system to serve the new room areas may be made, utilities will be connected, and the health physics and radiation monitoring equipment will be installed, calibrated, and tested. Air locks / check point areas will be provided at entries to Room 1140.

The structural loading capability of the DECON area will be investigated and reinforcing provided as necessary to permit the tem-porary storage of radwaste packages.

An estimated laydown area measur-ing 3.7 m (12 ft) x 19.5 m (64 ft) is required for storage.

Storage will be against the north wall.

3.3.3 Task 4: Removal of Reactor Components and Pool Liner Reactor components, as listed in Table 3-1. will be stored underaater in the reactor tank until they can be removed, air dried, and packaged for disposal.

Major activities associated with this task are:

o Taking dose-rate measurements o

Disassembling components (as needed) o Packaging for disposal o

Removing, testing, and cleaning (if necessary) water from reactor tank 3-20

[

1 t

O Contaminated waste will be packaged in suitable containers for shippi~g, in accordance with the NRC and 00T regulations for transportation of l

radioactive materials.

Packaging may be performed underwater if indi-t cated by measured activity levels.

1 The induced activity of the rotary speciaen-rack is estimated to be on i

the order of 100 R/hr at contact after that three month decay. Actual activity levels will be determined from measurements taken as the first r

step of this task.

Based on present estimates it appears that the rotary specimen rack can be packaged in one piece in a shielded container of 122 cm x 122 cm x i

244 cm (48 in. x 48 in. x 96 in.) high, enclosing 3.6 m (128 f t').

l 3

Avoiding the cutting of the rack will significantly reduce worker doses l

and will prevent generation of contaminated metal dust.

l j

The water from the reactor tank will be processed (using the existing purification system), checked for radioactivity, and then discharged by

)

app'opriate means.

If the conditions of 10 CFR 20.303 can be met the j

water will be released into the sanitary sewage system.

If not the water will be treated to remove the radioactive contamination or pro-cessed for disposal as low level waste.

i Ventilation during performance of this task will be the same as that t

i described in Task 3 above.

Removal of the aluminum pool liner will be l

accomplished as part of this task.

As the demolition of activated material proceeds, the radioactive mate-rial will be packaged for shipment and disposal as discussed in a

j Chapter 6.

}

The principal radiological hazards to personnel performing these tasks l

j is exposure to direct radiation and possible inhalation of airborne radionuclides.

The most significant hazards will be non-radiological i

and include falling debris, and loud sound levels.

[

To minimize the potential effects of these hazards, specific require-f ments will be implemented. These are shown for each hazard in Table 2-4 i

(Chapter 2). Monitoring of safety program efficiency will be performed l

I by the health physics technicians and supervisory personnel.

I 3.3.4 Task 5:

Removal of Components with Potential Surface Contam.

l ination and Other Activated Materials j

The level of fixed and removaole surface contamination of components will be determined in Task 2.

Based on these results the surface con-tamination of non-usable components will be removed until the levels l

prescribed in Regulatory Guide 1.86 are met to permit disposal as non-radioactive materials. A few components may be appropriately packaged l

j and transported for reuse and these may not require complete decontami-1 nation.

In some instances the level of effort required to decontaminate may be unacceptable for economic reasons and the material / equipment will j

be disposed of as low-level radioactive waste.

1

)

3-21

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

Examples of equipment and material that may require remwal are:

o Portions of the cor. rete biological shield o

Horizontal source storage conduits at the northeast corner o

Vertical source storage conduits at the north side o

"Rabbit" conduit to fume hood in northwest correr o

Fuel storage well at the southeast corner o

Activated concrete b ' logical shield Major activities associated with this task are listed in Section 3.1.3.5.

Neutron activation of a portion of the concrete in the exposure room and reactor pool area will require removal, packaging, and disposal as low level waste.

This will require demolition techniques and equipment.

The concrete will be removed a section at a time and supports w:11 be placed in the cavity formed as needed, before proceeding with the next section.

At the completion of the activated concrete removal, dose rate measure-ments will be made to ensure that all necessary portions have been removed.

The limiting requirement is 5 pR/h above background at 1 meter.

As the demolition of activated material proceeds, the radioactive mate-rial will be packaged for shipment and disposal as discussed in Chapter 6.

The principal radiological hazards to personnel performing concrete removal is inhalation of airborne radionuclides and accessible gamma exposure rates of as high as 5 to 10 mR/h.

The most significant hazards will be non-radiological and include falling debris, loud sound levels during concrete breaking, flames from torches possibly used to cut rebar, and high humidity and heat levels during concrete breaking operations.

To minimize the potential effects of these hazards, specific require-ments will be implemented as described in Section 3.3.3 above.

Monitoring for airborne radioactivity will be done using both continuous and grab samples of exposure room air.

A continJous beta-gamma dir mon-itor will be stationed in Room 1140, with a sampling collection tube extended into the working area.

Such systems have operational sensitiv-ities of about 25% of the occupationa! maximum permissible concentra-tions (MPC) for the nuclides of interest.

A water spray hose, out-fitted with an atomizer tip to valve at the nozzle will be provided to the concrete removal crew to use during concrete breaking for dust con-trol.

The valve allows the operator to turn off the water during periods when dust is not being created, thus minimizing water on the work area floor.

3-22

Those components connected by pipes and tubes will be disconnected and (Q/

swipes taken at connections to determine if the internals have been con-taminated.

If the contamination levels at these points fall within Regulatory Guide 1.86 limits, the components will be discarded as scrap metal.

If the contamination levels are above Regulatory Guide 1.86 limits, the components will be dismantled and further cleaned; valves, fittings and irregular pieces will be appropriately disposed of as low level radwaste as discussed in Chapter 6.

The pool water purification system will be left intact for subsequent cleanup of the pool water and later removal.

3.3.5 Task 6: Cleanup and Removal of Tools and Equipment Upon completion of all dismantling operations the Reactor Room will be cleaned up.

Group 4 material that may require cleaning include such things as:

o General and localized ventilation system o

Confinement barrier o

Waste treatment equipment o

Contaminated tools o

Contaminated clothing The major activities performed during cleanup are:

o Removing all loose contamination from DECON zone including confined enclosure area o

Decontaminating waste packages, making dose rate check, and storing in lay-ous area for shipping o

Cleaning and reush',9 or disposing of contaminated tools and equipment; packaging for disposal o

Removing surface contamination in the reactor pit o

Dismantling confinement barrier and packaging for disposal o

Removing surface contamination from the entire DECON zone Equipment and supplies used during decommissioning will be checked for potential contamination and either properly dccontaminated for reuse or disposed of as solid radwaste.

The confinement barrier will be dismantled and the plastic sheets pro-perly packaged for disposal as los level waste. Contaminated clothing

(,)

will be appropriately packaged and either sent to a laundry licensed for V

cleaning contaminated clothing or disposed of as radioactive waste.

3-23

3.3.6 Task 7: Packaging and Shipment of Radioactivo Waste The iellowing major activities will take place during this task:

o Lowering the empty container into the reactor pit o

filling the container with activated metal components or acti-vated concrete rubble to meet packaging requirements o

Lifting of loaded containers from the pit and removal of the external disposal Jacket, and closing and sealing container o

Performing radiation survey of package exterior for possible contamination o

Cleaning package exterior as necessary and marking and labeling according to 49 CFR Part 173, 10 CFR Part 20 and waste disposal site criteria o

Loading and securing packages on over-the-road transport trailers as well as inspecting shipment, making final radia-tion survey, preparing all necessary documents, and notifying the appropriate local authorities of shipment time o

Making final dose rate check of packages and of loaded trailers o

Shipping packages to disposal site or to new place of use During the decontamination procedures outlined in Task 3 through 6, low level wastes will be packaged as they accumulate and stored within Room 1140 in the storage area.

A detailed discussion of the packaging and shipping of radioactive wastes is given in Chapter 6.

It is expected that the number of packages generated will not require more space than that available in the reactor room.

Therefore the pack-ages will be collected for shipment to the disposal site at the end of the decommissioning activities.

3.3.7 Task 8: Perform Termination Radiation Survey Upon completion of Tasks 2 through 7 a termination radiation survey will be conducted to demonstrate that surface contamination levels and ambient radiation levels are not in excess of applicable limits.

The following major activities v111 be performed in this task:

o Radiation survey of the Reactor Room o

Radiation survey of Etcheverry Hall o

Radiation survey of potentially contaminated outside areas 3-24

o Sampling of soil and water O

o Preparation of termination radiation survey report Further description of the termination radiation survey is provid d in Chapter 8.

3.3.8 Task 9:

Demolition of Non Activated /Contaminatod Portion of Reactor installation When Task 8 is completed, and prior to the start of Task 9, it is re-quired that the NRC conduct an on-site survey to verify that the activity and contamination levels are satisfied. When the requirements are satis-fied NRC will issue an order that terminates the license and any further NRC jurisdiction over the facility releasing it for unrestricted use.

The schedule SF'wn on Figure 3-4 allows three weeks for the NRC inspec-tion plus four additional weeks for the receipt of the termination order.

Task 9 consists of the removal of the remaining non-activated /contami-nated portions of the reactor installation which includes concrete, steel walkways and stairs, miscellaneous wiring, and conduit to be demolished by conventional means.

This task is based on the assumption that the radiation level of the remaining concrete shield is at the low value previously established.

Samples of the remaining concrete will be taken to determine whether any residual hot spots exist.

If none are found, it is assumed that the concrete rubble may be disposed of at an off-site disposal location without restriction. This task also includes v

obtaining the necessary dumping permits and identification of the dis-posal area. The major activities include those listed in Section 3.1.3.9.

3-25

I CHAPTER 4 SAFEGUARDS AND PHYSICAL SECURITY r

4.0 INTRODUCTION

All BRR fuei, which accounts for the vast majority of special nuct:ar material, will be removed from the rcactor and shipped off-site under the operating license before physical decommissioning activities are initiated. Thus while the existing ucurity access controls and alarm systems may continue to be used they are not required by the decommis-sioning activities, 4

?

4.1 DECON AREA During decontamination activities Room 1140 will be divided into two regions by temporary wooden and plastic partitions for contamination d

control purposes as described in Section 3.1.

Access to the region con-taining the BRR, which is to be decommissioned, will be via the person-nel air lock and by the vehicle door through a te.mporary trailer which serves as a change room. Access to the uncontaminated laboratory and experimental equipment region will normally be via the personnel air-i lock.

Barriers will be installed at these two entrances to control per-i sonnel access and spread of contamination.

During times when Room 1140 is unoccupied, security such as locked doors O

will be provided to prevent casual, inadvertant, and unauthorized entrance.

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CHAPTER 5 RADIOLOGICAL ACCIDENT ANALYSIS Chapter 5 is not required since nuclear fuel will not be on-site during decommissioning.

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5-1

G CHAPTER 6

/b RADIOACTIVE MATERIALS AND WASTE MANAGEMENT i

6.0 INTRODUCTION

i I

During the decomissioning activities, radioactive materials in liquid, solid, and particulate forms are expected to be genereted. Management of these wastes is an integral part nf the DP.

Provisions for minimiz-ing the amount of waste generated, and waste collection, treatment, pack-aging, and shipment off-site for disposal are discussed in the following sections.

6.1 FUEL DISPOSAL The University intends to remove the fuel from the BRR under the current operating license. Therefore, at the beginning of decommissioning oper-ations, there will not be fuel on-site.

6.2 LIQUID RADWASTE Liquid radwastes generated during decomissioning activities will be collected, monitored, and released to the sanitary sewage system if the conditions of 10 CFR 20.303 can be met. Otherwise the water will be

[

treated to remove the radioactive contamination or processed for disposal as low level radioactive waste.

Expected sources of liquid radwaste are:

f o

Decontamination of components and parts o

Personnel decontamination liquids f

4 o

Oecontamination of structures and floors Efforts will be made throughout the decomissioning activities to mini-mize the generation of liquid waste.

Whenever possible, scrubbing with swabs will be used instead of spraying.

The water mist used during the i

I demolition of activated concrete will be closely controlled, and approved liquid absorbers will be used around the floor to absorb any runoffs.

Liquid waste will be absorbed or solidified using absorbent materials or solidification agents specified in the license requirements of the rad-waste disposal site. Absorbent material will be provided to absorb at least twice the volume of radioactive liquid contents in all radwaste packages.

6.3 SOLID RADWASTE A

The solid radwastes generated during decomissioning activities will be U

packaged on-site in containers suitable for shipping and disposal to one of the three licensed disposal sites in the United States.

6-1

6.3.1 Packaging The types of solid radwaste to be packaged include the following:

o Demolition Materials - These include the aluminum liner, con-crete rubble, reinforcing steel, and steel anchors.

The total area of activated aluminum to be removed from the reactor pit 2

floor and wall is expected to be approximately 42 m2 (450 ft ),

The total volume of activated concrete to be removed from the exposure room and the reactor pit floor and wall is expected to be 82 m2 (2,900 ft ).

Assuming!a packing density of 50%

8 and 3.6 m2 (128 f t ) metal or wood containers, approximately 3

52 containers will be necessary. Other potentially activated demolition materials are some of the reactor tank anchors, beam ports, and the reinforcing steel from the activated por-tion of the reactor pit floor and exposure room.

The small volume of these materials will permit their packaging in a fraction of a container, most probably mixed with other solid radwastes. This volume of waste will require 5 to 12 truck shipments to the disposal site. The loading for each shipment is dependent on the individual truck's weight limits. A flat-3 3

bed truck may carry up to six 3.6 m (128 ft ) metal boxes.

Equipment and Tools - This includes such items as saws, Jack o

hamm; s, forklift, shovels, pumps, tanks, ventilation system components, filters, and piping.

Not all of this equipment is expected to be discarded as radwaste. A determination of volumes of solid radwaste generated from this category will be possible only during cleanup (fask 6) when measurement of contamination level and evaluation for decontamination will be made.

o Auxiliary Materials and Clothing - This includes the con-finement barrier plastic sheets, protective mats, rags, work platforms, and protective clothing.

It is assumed that these materials will be compacted and packaged in six 208 1 (55-gal) drums.

o Activated Equipment - This includes primarily the rotary spec-imen rack which is planned to be packaged in one piece in a 3.6 m (128 ft ) shielded container. Other small comp" ents 3

3 would probably be packaged in the same container since

.e rotary specimen rack will occupy only part of the volume.

Other activr.ed equipment includes the core support assembly, top and bottom grid plates, and associated equipment, Wooden boxes will require a variance at the Richland, Washington, site l

but are accepted at Beatty, Nevada.

6-2

9 6.3.2 Container Handling (V

Because of the small work areas ~available, the demolition materials resting at the reactor pit floor and exposere room floor will require removal as the demolition proceeds. An area of the floor will be kept clear during demolition to allow laydown of an empty container. After the equivalent of a container content is demolished, a container will be lowered with the aid of the bridge crane.

It is anticipated that a pro-tective plastic jacket will be placed on the outside of the container to minimize deposition of contaminated dust; the igper edges will be folded inside the container.

After the container is laid down, the loading of concrete rubble will proceed. The filled container will then be lifted from the pit with the aid of the bridge crane and moved inside the lay-down area within the confined enclosure of the Reactor Room floor. At this point the container cover will be placed on the container and the plastic jacket cut off around the cover (fold-over plastic left in the container). The containers will then be moved with a fork 14ft to the storage area within Room 1140 and the cover sealed.

6.4 VENTILATION SYSTEM There are two separate ventilation systems operating in the Reactce Room:

a general ventilation filtered exhaust provided by the existing roof fan and a new localized ventilation filtered exhaust serving the confined enclosure around the reactor pit.

V The two ventilation systems will operate to ensure a negative pressure in the Reactor Room with respect to surrounding areas of Etcheverry Hall and a (greater) negative pressure within the confinement barrier with respect to the Reactor Room.

The purpose of this arrangement is to assure that the air flows from the non-contaminated area towards in-creasingly contaminated areas.

Both ventilation exhaust systems will be equipped with roughing filters to capture large particles, and with high-efficiency particulate absorp-tion filters (HEPA) to provide up to 99.99% particulate retention. HEPA filters will be changed in the event of high radiation level readings /

alarm in the exhaust duct or based on maximum pressure differential readings indicating that the filters are filled with dust.

Radiation monitoring will be provided for the exhaust air to atmosphere; readings above prescribed limits will shut off the exhaust fans. The ventilation unit serving the confined enclosure will be a moLile type and connected with flexible ducts to the enclosure and to the exhaust duct. The exhaust duct will discharge the air to the general ventilation e:.haust near the ceiling on the west wall.

The exhaust duct will be provided with gravity louvers which will automatically close in the event of failure of the ventilation exhaust unit.

Fresh air intake to Room 1140 will be provided by the existing systems.

Air supply to the confinement barrier will be through intake grills from Room 1140.

OU 6.5 WASTE CLASSIFICATION The criteria for waste classification for low-level waste disposal is contained in 10 CFR 61.

6-3

The significant radionuclides generated as a result of neutron activa-tion are shown in Table 1-6.

In addition the BRR contains a very small quantity of stainless steel, present in components such as dowel pins, captive screens, and spacers.

It is concluded that the radioactive wastes from the BRR can be classified as Class A.

Class A wastes do not need to be segregated for disposal, providing they meet the stability criteria described ir, paragraph 61.56 of the regulation (i.e., wastes do not structurally degrade and affect the overall stability of the dis-posal site through lurrping, collapse, etc).

The radioactive waste from the BRR decommissioning will meet these cri-teria because they will be in solid form (aluminum liner, and concrete and other metal components). To further comply with the regulations, the containers will be filled so that voids will be kept to a minimum for ensuring structur al stability when overburdened or when other packages are placed over them.

6.6 SHIPPING OF RADIOACTIVE WASTES Department of Transportation Regulation 49 CFR 173 provides radiation level limitations for transportation of packages of radioactive mate-rials in closed, exclusive-use transport vehicles as follows:

o External radiation levels must not exceed 1000 mR/h on the accessible surface of the package if the shipment is made in a closed transport vehicle, the package is secured, and there is no loading or unloading operations during transit (49 CFR g

173.441)

W o

The radiation level must not exceed 200 mR/h at any point on the outer surface of the transport vehicle, 10 mR/h 2 meters from the vehicle sides and 2 mR/h in the tractor cab (49 CFR 173.441)

A quality control program will be established to ensure that radioactive waste shipping regulations are enforced.

Such a quality control program will include the following requirements:

o Waste Containers All containers are 00T specification 7A, Type A packages or Type B packages if the A quantity is exceeded (49 2

CFR 173.425, 178.350, and 173.24).

Exemptions to this must be specified.

All containers are in good physical condition with no evidence of damage, corrosion, or leakage (49 CFR 173.475 and 173.24).

All tretal drums with a capacity of 208 1 (55 gal) or greater will have 1.6 cm (5/8 in.) or larger bolt for securing the closure device (ring assembly). All metal containers will have an intr.ct heavy-duty closure device when presented for disposal.

6-4

Ring bolt is torqued to approximately 45-ft pounds 7~. -

(recommended).

Drum lids are a proper fit, and bungs (if any).are tight (49 CFR 173.475).

Radiation levels at the container surface do not exceed 200 mR/h unless excepted by 49 CFR 173.441.

Surface contamination levels are below 00T limits (49 CFR 2

2 173.443) 220 alpha dpm/100 cm and 2200 beta dpm/100 cm.

o labels and Markings Each container has two radioactive labels affixed to it (49 CFR 172.403 and 49 CFR 172.406).

Each container has been marked "Radioactive Material, LSA, n.o.s. UN 2912" in greater than 1/2-in.-high letters (49 CFR 172.101, 172.301).

Each container has been marked "USA 00T 7A, Type A" or "00T Type B" in 1/2-in.-high letters (49 CFR 178.350 and 172.310).

The waste class has been marked on each container using h,

greater than 1/2-in.-high letters (10 CFR 61.57 and WA V

License).

Waste Class Example Class A Unstable Class A Stable Class B Class C For each container in excess of 110 lbs., the weight and unit of measurement have been marked on the container (49 CFR 172.310).

Markings must be durable and legible, and displayed on a background of sharply contrasting color, unobscured, and located away from any other marking, such as advertising, (49 CFR 172.304).

The name and address of the shipper has been attached to each container if vehicular transfers are infolved (49 CFR 172.306).

O v

6-5

o Transport Vehicle The total transport index number does not exceed 50 (49 CFR 177.842).

This is not applicable to exclusive-use vehicle shipments.

The containers have been loaded, blocked, and braced so that they cannot change position during conditions nor-mally incident to transportation (49 CFR 177.425 and 177.842d).

o Exclusive-use vtihicle Exceptions Specific instructions for maintenance of exclusive-use shipment controls must be provided by the shipper to the carrier and be included with the shipping paper informa-tion (49 CFR 173.425 b9).

An exclusive-use vehicle shipment of Type A Low Specific Activity (LSA) radioactive material (RAM) is exempt from Type A packaging specifications if it meets a strong, tight packaging criterion (49 CFR 173.24).

Shipments must be loaded by consignor and unloaded by consignee.

There must be no loose radioactive material in the conveyance.

Shipment must be braced so as to prevent shifting of lad-ing under conditions normally incident to transportation (173.425).

The transport vehicle must be placarded, and the placards must be located away from any markings (such as advertis-ing) by at least 3 in. (49 CFR 172.500 and 172.516 c4).

Exclusive-use vehicle shipments of LSA RAM are exempt from specified markings and labeling if the exterior of each package is stenciled or otherwise marked "Radioactive -

LSA."

o Documentation The generator has a valid disposti site and (where required) state User Permit.

All required documents are fully completed and are legible.

The Radioactive Shipment Manifest (RSM), formerly RSR/

Manifest, is complete in all details.

(49 CFR 172.200 Subpart C, 10 CFR 20.311, and radwaste disposal site requirements.)

6-6

The number of containers listed on the RSM agrees with O

i the physical count of containers loaded.

All required certifications, RSMs, and other documents as appropriate are signed.

The use of abbreviations must conform to 00T and NRC specifications, o

Shipping Routes The route out of Berkeley to the disposal site is chosen using the PC Miler Program which calculates the shortest highway routes.

Local, state, and federal ordinances may impact the routes selected to one of the three licensed disposal sites in the United States.

I

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t 6-7

CHAPTER 7 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS xy

7.0 INTRODUCTION

The technical and environmental specifications will be implemented to control conditions, parameters, and variables so that:

During DECON activities the radiation exposure to workers, all o

personnel occupying Etcheverry Hall, and the public shall be maintained "ac low as reasonably achievable" (ALARA).

o After DECON, portions of the facility can be released to unre-stricted use. Other areas within Room 1140 of Etcheverry Hall will remain under the State of California Radioactive Materials License.

The technical and environmental specifications will include items in the following categories:

o Health and safety limits o

Surveillance requirements (S

o Administrative controls o

Design features j

7.1 HEALTH AND SAFETY LIMITS Health and safety limits for DECON must be maintained for adequate con-trol of the DECON activities and for subsequent unrestricted use.

The limits cited in this section will be enforced by adhering to the appro-priate procedures.

During DECON activities the following limits shall be enforced.

7.1.1 External Exposure External exposure for individuals in restricted areas during decommis-sioning shall not exceed the limits specified in 10 CFR 20.101.

7.1.2 Internal Exposure Internal exposure from inhalation of radioactive material in air in restricted areas shall not exceed that which would result from the inhalation of the limiting quantities specified in 10 CFR 20.103.

7.1.3 Concentration of Airborne Radioactive Materialin Restricted Areas Concentration of airborne radioactive material shall not exceed the limits set in 10 CFR 20.103.

7-1

7.1.4 Concentration of Airborne Radioactive Materialin Unrestricted Areas Concentration of airborne radioactive material shall not exceed the limits specified in 10 CFR 20.106.

7.1.5 Concentration of Non Radioactive Substances in Restricted Areas Concentrations of non-radioactive substances shall not exceed limits established by the American Conference of Governmental Industrial Hygienists (ACGIH) as listed in "Threshold Limit Values" (TLV) and Bio-logical Exposure Indices for 1987-1988."

7.1.6 Concentration of Non. Radioactive Substances in Unrestricted Areas Concentration of non-radioactive substances shall not exceed 1/30 the ACGIH TLV limits.

7.1.7 Noise Levels Noise levels shall not exceed the permissible noise exposures of Cal OSHA.

7.1.8 ALARA The University is committed to the concept of ALARA in terms of both individual and collective doses. Administrative controls; training, protective devices, and measures will be used to achieve this commitment to ALARA and minimize the hazards to health and safety.

7.1.9 Health and Safety Limits for Unrestricted i'se The material directly related to the BRR operat hns, remaining in those areas / structures which will be verified as meeting the requirements for release to unrestricted use or materials released off-site for unres-tricted use, shall meet the following criteria:

Totalmaximumannualdosetomanfromallexposureythways c

shall not exceed 10 mrem per year above background The dose rate at 1 meter from an exposed surfqc shall not o

exceed 5 p-Roentgen per hour above background \\

1 Surfacecontamination1qvelsshallnotexceedthoselistedin o

Regulatory Guide 1.86 (1) Guidance and Discussion of Requirements for an Application to Ter-minate a Non-Power Reactor Facility Operating License.

Rev. 1, September 15, 1984, USNRC.

(2) Source:

Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors, June 1974.

7-2

7.2 SURVEILLANCE REQUIREMENTS bV The surveillance activities during and after DECON are described below.

7.2.1 The Environmental Dosimeter Program This program was carried on during reactor operations and will be con-tinued during DECON activities.

Locations, type, and dosimeter change schedule will be adjusted to accommodate DECON activities.

An abbrevi-ated environmental surveillance program will be conducted for one year following DECON.

7.2.2 The Routine Swipe Program This program will continue through DECON.

Frequency and locations will be adjusted to accommodate DECON activities.

Following release, sur-veillance of the facility will be conducted under the State of California Radioactive Material License.

7.2.3 The Routine Instrument Survey Program This program will be adapted to reflect the change in status.

Neutron monitoring will be discontinued when fuel is removed from the site.

Locations, frequency, and type of instruments will be adjusted to accom-modate DECON activities and radiation levels.

Routine instrument sur-O veys of Room 1140 of Etcheverry Hall will continue post DECON under the State of California Radioactive Materials License.

7.2.4 The Air Sampling Monitoring Program This program will continue through OECON.

The continuous air monitor will be relocated outside of the reactor enclosure. Monitoring inside the enclosure will be conducted for personnel exposure under the Health and Safety Program.

All instruments / systems used for monitoring purposes shall be properly calibrated prior to use.

7.3 ADMINISTRATIVE CONTROLS The administrative controls which will be used during DECON are dis-cussed below together with the responsible organization and documenta-tion requirements.

7.3.1 Administrative Controls During DECON Administrative controls during DECON are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure completion of the decommissioning of the BRR in a safe manner.

OG 7-3

7.3.2 Responsibility The UC Decommissioning Project Engineer shall be responsible for the completion of DECON activities at the University.

7.3.3 Organization The organizational structure for management and performance of the decom-missioning activities is shown in Figure 1-8.

The functions and respon-sibilities, and minimum required qualifications and experience of each position are detailed in Section 1.5.

7.3.4 Records and Reports Accurate and complete records and reports shall be maintained by the University of the performance and completion of all activities which may result in exposure of workers or the public to radiation or other hazardous / toxic materials, o

Records Records which will be maintained during DECON activities are listed below.

Health and Safety Related Activities:

Work permits Work procedures Radiation survey reports Contamination survey reports Airborne survey reports Environmental survey reports Counting data on air samples, smears, and gamma spectrum analysis Instrument calibrations Source inventory and storage Radioactive material inventory and storage Shipment records Waste disposal - surveys and records Package certifications / records Incidents and accidents Personnel Records Bioassay analysis Personnel exposure records Individual dosimeter readings as related to daily tasks and work procedures Respiratory protection qualifications (medical clearance and fit test)

Audiogram results Training records Visitor logs and expcsure information 7-4

-s o-Reports Reports pertaining to DECON activities shall be written and

[-

submitted to the proper authorities pursuant to Regulatory Guide 1.86 i

o Review Responsibility for review of procedures, practices, and per-

.formance shall rest with the appropriate individuals and/or committees detailed in Section 1.5 7.4 ENGINEERING CONTROLS Confinement barriers and HEPA filtered ventilation systems will be employed during DECON activities. Task 3, as outlined in Chapter 3, details installation of confinement barriers.

These systems will provide:

o Maintenance of a negative pressure in the confined area of Room 1140 in respect to the rest of Etcheverry Hall.

This will be accomplished by a HEPA filter system in conjunction with the installed confinement barrier.

o Maintenance of a negative pressure in Room 1140 with respect O'

to the surrounding environment.

This is accomplished by the existing HEPA filter system.

o Provisions for dust control during DECON activities which have the potential to generate airborne particulates.

O 7-5 t _.

/~T CHAPTER 8 V

PROPOSED TERMINATION RADIATION SURVEY PLAN

8.0 INTRODUCTION

A termination radiation survey of the facilities under NRC jurisdiction will be conducted in order to ensure that the DECON area satisfies the radiation requirements contained in Table 1 of NRC Regulatory Guide 1.86 and also satisfies the NRC acceptance criteria for unrestricted use of 5 pR/h above background at 1 m from a surface which has become radio-active as a result of reactor operations or decommissioning activities.

The termination radiation survey will be conducted after all other decommissioning activities have been completed.

The detailed plan for the final radiation survey will depend on:

The details cf the dismantling / decontamination process o

o The results of radiation surveys during that process The radiological history and other specific characteristics of o

the BRR facility o

The results of the preliminary survey 8.1 PRELIMINARY SURVEY A brief preliminary radiation survey will be made in order to formulate plans for an efficient, comprehensive survey. Before designing the pre-liminary survey, decisions will be made concerning logical divisions of the BRR and the neutronics laboratory into separate survey units or strata. A minimum number of 30 measurements will be made in each survey 2

unit, and each survey unit should cover an area of at most 30 m.

Dimensions of the survey units will be obtained so that a scaled drawing of each unit can be prepared prior to the termination radiation survey.

The BRR site has been continuously under a surveillance program includ-ing environmental monitoring and indoor radiation and contamination sur-veys during operation. Because of this data base, current survey data may be incorporated into the preliminary survey data if it meets the l

criteria of NUREG/CR-2082. The preliminary survey ensures the presence of grid markers for defining survey unit numbers and locations, and a grid map for locating and recording preliminary measurements.

The pre-liminary survey will aid in deciding how to sample the site; that is, whether by random sampling, stratified random sampling, systematic sam-pling, or other methods.

8.2 TERMINATION SURVEY PROCEDURES 8.2.1 Indoor Survey for the termination radiation survey, each indoor survey unit is divided into two subunits:

(1) lower surfaces comprised of floor surfaces, wall surfaces up to a height of 2 m, and any other surface easily accessible 8-1

to a surveyor standing on the floor; and (2) overhead surfaces comprised of ceiling surfaces, wall surfaces more than 2 m above the floor, and all other surfaces not described in (1).

The floors and lower walls will be divided by a rectangular grid system such as that shown in Figure 8-1.

The smaller blocks formed in this manner are referred to as "survey blocks," and the corners of the survey blocks are called "grid points." The choice of the particular gri<l system is guided by the following rules:

o No survey block should measure less than 1 m on a side.

Sur-vey blocks of less than 1 m on a side would require an imprac-tically large number of measurements in the buildings.

o No survey block should measure more than 3 m on a side.

Sur-vey blocks larger than 3 m on a side could lead to large uncertainties as to the precise location of the contamination.

o There should be at least 30 measurements per stratum.

o The maximum number of survey blocks needed in the survey should be calculated according to NUREG/CR-2082.

The radiological conditions to be characterized on the lower surfaces include alpha contamination levels (by direct reading), beta-gamma dose rates at 1 cm above the surface, external gamma radiation levels at 1 m above the floor, and removable alpha and beta contamination levels.

At 1 m above the center of each survey block, the external gamma radia-tion level is measured. At the surface in each survey block, five direct measurements each of alpha contamination levels, beta-gamma dose rates, and gamma radiation levels will be made at uniformly spaced points in a 1 m area in the center of the survey blocks as shown in 2

Figure 8-2.

(If the entire survey block has an area of approximately 2

2 1 m, then the "corner" measurements shown in the 1 m area in Figure 8-2 are moved 30 cm toward the center of the block.)

For each type of 2

measurement, the average value and the local variability in this 1 m 2

area can be estimated.

For an area of only 1 m, five alpha or beta-gamma measurements will be used to yield an estimate of the average in that area. The survey block is next scanned with a G-M m (open window),

the point showing where the maximum reading (if any) is located; and each type of measurement (including smear samples of measurements of transferable alpha and beta contamination levels) is made at this "beta-gamma maximum point."

The manner of data recording for an indoor survey is illustrated by the headings in Table 8-1. The following minimum data are needed:

o Survey unit numbers, identifiable on a scale drawing, and the building name or number; the building floor number; the sur-faces surveyed; and types of measurements and the units (dpm/100 cm, millirad /h, and/or R/h; mGy/h and microGy/h) 2 8-2 I

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e AND y LEVELS AT SURFACE Figure 8 2 Example of Maximum Observed Octa Gamma Dose Rate and Direct Alpha Beta Gamma Points In a Survey Unit.

8-4

O Table 81 bl Alpha, Beta Gamma, and External Gamma Radiation Levels in Room 1140, Etcheverry Hall, including Floor and Lower Wall Surf aces Directly measured Directly measured contamination at contamination surface at External gamma Survey center of block maximum beta-gamma point radiation level block Beta-gamma dose Beta-gamma dose 1 m above floor Alpha rate at 1 cm Alpha rate at 1 cm (R/hr) 2 2

(dpm/100 cm )

(millirad /hr)

(dpm/100 cm )

(millirad /hr)

A2 100 0.05 100 0.45 NAa b

3 100 0.08 NR 0.08 NA 4

200 0.13 NR 0.13 NA 5

210 0.13 NR 0.13

~~NA 6

50 0.13 NR 0.13 NA 7

160 0.05 NR 0.10 NA NA 8

220 0.15 270 0.23 NA 9

100 0.15 210

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10 210 0.15 NR 0.15 NA 11 100 0.15 NR 0.15 NA 12 120 0.15 NR 0.15 NA 13 120 0.28 120 0.28 NA B1 210 0.05 NR Q.'05[

HA O2 150 0.10 270 OMO 40

'd 3 160 0.10 1'

8.5 35 4

220 0.18 26a 6.0 45 5

260 1.8 160 2.0 40 6

210 0.10 280 3.5 50 7

150 0.35

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90 0.50 75 8

280 0.25 210 1.5 70 9

160 0.30 160 2.8 100 10 290 0.25 240 3.0 130 11 260 0.25 210 0.35 120 12 150 0.20 110 0.70 130 13 100 60 150 0.75 220 14 160 NR 0.15 NA C1 0.05 NR 0.05 NA 2

0.09 NR 0.09 35 3

0) 0.08 NR 0.15 40 4

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0.13 140 0.35 60 5

0.15 130 1.3 80 6

120 0.13 290 1.3 75 7

140 0.13 80 0.30 70 8

220 0.30 280 3.5 50 9

260 0.08 200 0.65 55 10 110 0.11 270 0.20 55 11 210 0.15 260 0.33 90 f

A 12 170 0.15 100 0.20 85 V 13 170 0.40 210 1.0 170 l

b

NA = not applicable HR = no reading taken 8-5

o Name of surveyor taking measurements, date of survey, location data relative to the grid coordinates, and log book pages for original data o

Surface smears, plaster chips, etc., taken, and the indoor block number from where they were taken, the container number, and whether matched to any air readings Type, model number, calibration data, sensitivity limit, and o

any other information needed about the portable survey instru-ments to interpret the data obtained with these instruments and to ensure quality control on the data so obtained o

When the unit surveyed is below the sensitivity of the instru-ment, the fact that such a measurement was made will be included as a significant datum 8.2.2 Outdoor Survey Procedures The land will be divided by a rectangular grid system.

The choice of the particular grid system is guided by the following set of rules:

No survey block should measure less than 5 m on a side o

o No survey block should measure mord than 15 m on a side o

There should be a minimum of 30 measurements The number of grid points should be calculated according to o

NUREG/CR-2082 At each grid point, beta-gamma measurements will be made within 1 cm of the surface and a second gamma measurement made at 1 m above the sur-face.

These grid-point measurements are considered "unbiased" and are used to estimate average gross gamma and beta-gamma radiation levels.

Each outdoor survey block may be quickly scanned with a gamma scintilla-tion survey meter.

If soil samples collected independently of the gamma and beta-gamma readings during the preliminary survey indicate that the contamination consists largely of beta-emitting nuclides such as U-238 with its short-lived daughters, the survey block should be scanned with a G-M meter, with the cpen-window probe held no more than a few centime-ters from the surface.

If a maximum gamma (or beta-gamma) point in the survey block is found during the scan, gamma measurements at the surface and at 1 m and a beta-gamma measurement at the surface will be recorded for this point.

The outdoor survey will include collection of surface soil samples for determination of radionuclide concentrations.

These samples will be taken in the upper 5 to 15 cm of soil.

Maximum concentrations of radio-nuclides in surface soil will be estimated from samples collected at points showing the highest gamma or beta-gamma radiation levels.

Aver-age radionuclides concentrations are estimated from "unbiased" samples taken at randomly selected points within each stratum.

8-6

O The outdoor area to be surveyed.will be divided into survey units 10 m x C/

10 m, which in turn can be subdivided into smaller blocks if the pre-liminary survey indicates the need. By transit survey, the entire site will be staked out with markers (grid points) and a scale drawing made.

Next, if stratification seems warranted in terms of local site condi-tions, subdivision into three or more large survey units or strata on the scale drawing will be made. Stratum 1 represents the highest poten-tial hazard area from the preliminary survey.

This scale drawing can then be used for various recording and collating purposes.

8.2.3 Instrumentation and Methods for Contaminated Surface Surveys In determining instrumentation to be used in the decommissioning activi-ties, isotopes of concern must be identified.

Those isotopes that may be present as contamination are listed in Table 8-2.

The instruments to be used in the radiation surveys will be selected on the basis of the type and level of radiation anticipated as a result of reviewing BRR historical data, the decontamination process, and the results of the preliminary survey.

Both the direct and indirect (wipe) monitoring methods will be used to obtain a complete assessment of the surfaces being examined during the termination survey.

For direct methods of surface monitoring, the scan-ning speed will be slow enough to ensure detection of at least 50% of all the small discrete sources, the emission rates of which, if averaged 3

over 100 cm, would be &t the guide levels specified in Table 1 of NRC 2

Q Regulatory Guide 1.86.

When surface activity is detected, the probe will be held stationary for a quantitative measurement.

The amount of removable radioactive material will be determined by wiping representative portions of the surface being surveyed with soft absorbent material and applying uniform moderate pressure.

Efforts should be made to standardize, as much as practical, the procedure for taking the wipe.

8.2.3.1 Instrument Selection The instruments used for direct and indirect monitoring of surface contamination will be capable of measuring surface activity at the guide levels specified in Table 1 of NRC Regulatory Guide 1.86.

Instruments will be tested and calibrated in accordance with the specifications contained in the American National Standard, "Radiation Protection Instrumentation Test and Calibration," ANSI N323-1977, or the most recent revision.

Table 8-3 lists the detection capabilities of the various types of sur-vey instruments that could be used to support decommissioning activities.

OL.)

8-7 t

Table 8-2 Isotopes Potentially Present at the BRR and Their Principal Decay Characteristics Major X-Ray or Gamma Ray Energies Isotope Half-Life Particles (MeV)

C-14 5730 y B-None Ca-45 163 d 0.077(99.98%)

Hn-54 303 d EC(a) 0.835(100%)

Fe-55 2.7 y EC 0.005(60.7%)

Co-57 270 d EC 0.122(87%);0.136(11%)

Co-58 71 d EC, +

0.511(30%);0.810(99%)

Ni-59 7.5x104 y EC 6.00075(134%)

0.006(54.9%)

Co-60 5.26 y B-1.173(100%);1.332(100%)

Ni-63 100.1 y B-0.017(0.0365%)

Sr-90, Y-90 27.7y/64 h B-None Ru-106/Rh-106 368d/30 s B-0.512(21%);0.622(11%)

Sb-129 2.71 y B-0.427(31%);0.463(10%);0.599(24%);

0.634(11%)

7 I-129 1.7 x 10 y B-Xe x-rays, 0.040(9%)

Cs-134 2.05 y B-0.57(23%);0.605(98%);0.796(99%)

Cs-137/Ba-137m 30.0y/2.6 m B-0.662(85%)

Ce-144 284 d B-0.080(2%);0.134(11%)

Eu-152 12.7 y B,EC 0.122(37%);0.344(27%);0.799(14%);

0.965(15%);1.087(12%;1.113(14%);

1.408(22%)

Eu-154 16 y B-0.123(38%);0.724(21%);0.876(12%);

1.00(31%);1.278(37%)

Eu-155 1.81 y B-0.087(32%);0.105(20%)

Ac-227 21.6 y B-(99%),a 0.0704(1.2%);0.084(2.7%);0.0995(2 0.160(1.4%)

U-232 72 y o

0.058(21%);0.129(.082%)

8-8

=

Table 8 2 (Cont.)

Major X Ray or Gamma Ray Energics Isotope Half Life Particles (MeV)

U-233

.1.62x10 y a

many low-yield 0.37;(lines 0.029 5

to 0.366)

U-234 2.47x10 y a

0.0533(0.068%).0.1029(0.023%)

5 U-235 7.1x10.8 y a

0.143(11%),0.185(54%)0.204(5%)

0.068(0.6%),0.142(0.07%)

Th-230 8.0x10' y a

U-238 4.51x10' y a

.050(0.324%)

Pu-238 86.4 y a

Uranium L-shell x-rays Pu-239 2.4x10' y a

0.039(0.006%),0.052(0.021%),

0.129(0.006%),0.375(0.0016%),

O.414(0.0015%)

Pu-240 6600 y a

0.45(0.045%),0.104(0.007%)

Pu-241 13.2 y 8-Uranium x-rays Am-241 458 y a

0.060(36%)

(a)ElectronCapture.

a O

4 8-9

Table 8 3 Typical Minimum Detection Capabilities for Various Survey Instruments Surface Types Nuclide Instrument or Method Detection Levels 8

2 Soil Gross Thin-walled, shielded

> 10 pCi/cm, very slow scan Alpha Pancake G-M 2

2 ZnS scintillator 1 10 pCi/cm, source dependent 2

2 Thin NaI or CaF

> 100pCi/cm, source dependent 2

Gross Thin-walled, shielded 500-5,000pCi/cm,

Beta-gamma Pancake G-M Emax > 0.15 MeV 2

ZnS scintillator 300 pCi/m, source dependent 5

2 NaI spectrometer 500-5 x 10 pCi/cm 2

2 Ion chamber

> 10 pCi/cm 2

Phoswich 2 200pCi/cm 2

Intrinsic Germanium

> 200pCi/cm 2

Portable scintillator 5,000-50,000pC1/cm 2

Facility walls, Gross Thin-walled, shielded

-> 100pci/cm

floors, Alpha Pancake G-M equipment 2

ZnS scintillator

> 100pCi/cm 2

2 Thin flat or CaF

> 30pCi/cm,

for alpha delay x-rays 2

Proportional chamber

> 50pCi/cm very slow scan on surface 2

Gross Thin-walled, shielded

-> 30pCi/cm,

2 Beta-gamma G-M

> img/cm window 2

Zns scintillator 30-300pCi/cm 2

tial

> 500pCi/cm 2

Pressurized ion 50-50,000pCi/cm chamber 2

Air ionization 500-5,000pCi/cm chamber 8-10

Table 8-3 (Cont.)

Surf ace Types Nuclide Instrument or Method Detection Levels 2

Phoswich 10-100pCi/cm 2

Intrinsic germanium 30-3,000pCi/cm 2

Proportional chamber 10-100pCi/cm e

2 Portable scintillator 50-500pC1/cm O

l l

O 8-11 l

8.2.4 Documentation Radiation measurements and analytical results will include the following data:

o Location of the measurement or sample o

Date or dates of measurements or sample collection 3

o Measured concentration of the specific nuclides in pCi/m or 3

mBq/m for air samples; pCi/g or mBq/g for soil samples o

Measurements of radiation sources should be reported as 2

follows:

alpha in dpm/100 cm, beta-gamma dose rate at 1 cm in pR/h, and gamma at 1 meter above surface.in pR/h o

Analytical error at 95% confidence level should be reported for all analyses o

Name of surveyor, sampler, or analyst o

Analysis date o

Instrument specifications and calibration data o

ConfidenTe level, standard error, etc. attached to analytical results o

Name of person verifying results The actual net measured values (including negative values) and their associated errors will be reported.

For values lower than the lower limit of detection (LLD), the LLD will be provided.

Whenever possible, values lower than the LLO will be reported in the following manner:

11.1 1 18.5 pCi/L or mBq/L or 7.4 i 18.5 pCi/g or mBq/g.

The following supplemental information will be included:

o Description of survey and sapling equipment o

Survey and sampling procedures, including sampling times, rates, and volumes o

Analytical procedures o

Calculation methods o

Calculation of the lower limit of detection o

Calibration procedures o

Discussion of the program for ensuring the quality of results 8-12

The data will be presented so that the radiological condition of the f-g site is completely and accurately depicted and the radiological condi-tion of the site can be ascertained without further analysis and manipu-lation of the data.

A report will be written and submitted to the NRC on the Termination Radiation Survey as required by NRC Regulatory Guide 1.86.

The report will include a description of the survey methods, instruments, analyses, and an evaluation of the results.

The report is expected to conclude that the site is suitable for release to unrestricted use.

b 4

I I

t I

I 8-13 i

I I

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

.

ML20147D769

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6.0 INTRODUCTION

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ML20147D769

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3.0 INTRODUCTION

4.0 INTRODUCTION

6.0 INTRODUCTION

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