2016 Site Description: Santa Susana Field Laboratory (SSFL) 

Proposed Corrections to Technical Basis Documents 1 & 2
 

ORAUT-TKBS-0038-1 / ORAUT-TKBS-0038-2
 

CORE Advocacy for Nuclear & Aerospace Workers 

Prepared by:

CORE Advocacy for Nuclear & Aerospace Workers • COREAdvocacy.org
 

For Review by:
 

National Institute for Occupational Safety and Health (NIOSH)
 

Presidential Advisory Board on Radiation and Worker Health (ABRWH)
 

Presented to NIOSH / ABRWH at the Idaho Falls, ID ABRWH Meeting, August 9, 2016
 
 


 


 


 


 


 


 

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Note: Building numbers and operations are non-sequential at SSFL.

Table 3 (p. 28) contains a numerical list of buildings and their page numbers.

The table also provides radionuclides of concern at each location, operations, number of

incidents, whether operations may have involved DOE-NASA or Area I, II & III workers, and 

whether the location was included/excluded from the 2006 Site Profile.

This document pertains to DOE - LMEC - ETEC Operations at SSFL. 

CORE Advocacy assumes no claim to authorship of data contained herein, nor liability
 

for its misinterpretation or misuse. Every effort has been made to credit appropriate sources. This document is 

intended for review by NIOSH, ABRWH, and federal agencies tasked with the administration of EEOICPA.
 

© CORE Advocacy 2016
 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 

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TABLE OF CONTENTS
 

Section Page 

1.0 INTRODUCTION: CORE Advocacy for Nuclear & Aerospace Workers 11
 

1.1 Purpose of the 2016 SSFL Site Description 12

1.2 Information Provided in the 2016 SSFL Site Description 13

1.3 Source Documents Used in the 2016 SSFL Site Description 14
 

1.4 Acceptance of Source Documents by Federal Agencies 15
 

2.0 OVERVIEW: 2006 SITE PROFILE 16
 

2.1 Site Nomenclature: 2006 Site Introduction / Site Description 16

- Misrepresentation & Exclusion of Buildings, Facilities, Operations

- “Non-Radiological” Facilities Involved in Radiological Processes 

2.2 ETEC Site Description - ORAUT-TKBS-0038-2 17
 

- Questionable Source Data 

3.0 2016 SSFL SITE DESCRIPTION & GENERAL INFORMATION 18
 

3.1 Brief Description & Site History 18
 

3.2 AEC’s Support of Space-Nuclear Research 19
 

3.3 Corporate Mergers & Site Nomenclature 20
 

3.4 Environmental Monitoring & Sampling / Questionable Environmental Data 20

- Questionable Sampling Practices Observed by EPA:

 [Soil - Water - Vegetation - Environmental Dosimetry] 

- Observations on Environmental Data

 [DOE Office of Environmental Audit & Tiger Team] 

4.0 NUCLEAR REACTOR DEVELOPMENT 23

4.8.1 Isotopes of Concern 24

4.8.2 Magnitude of Site Activity 25

Table 1: Research Reactors at SSFL / Radioactivity at the End of Operations 26
 

Table 2: Criticality Test Facilities at SSFL 27
 

TABLE 3: SSFL FACILITIES: RADIONUCLIDES & BUILDING USE* 28
 

* Table 3 provides the following information: SSFL buildings/functions in numerical order; changes to 

building numbering schemes; Visitor Log Entry Codes (partial list); radionuclides used at each location; 

existence of Tritium / Neptunium; specifies DOE-NASA joint projects; number of incidents associated with 

each location; and shows whether the location was included/omitted from the 2006 Site Profile. 

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5.0 DEVELOPMENT & TESTING OF NUCLEAR REACTORS 44
 

5.1 Homogenous Water Boiler Reactors (HWBR) 45
 

5.1.1 Building 4073 - Kinetics Experiment Water Boiler (KEWB) Reactor 46

5.1.2 Building 4093 - Water Boiler Neutron Source (WBNS) AE6/L-85 Reactor 52
 

5.2 HWBR: Additional Locations / Operations to Include in Site Profile 56
 

5.2.1 Building 4074 - KEWB Support 57
 

5.2.2 Building 4083 - KEWB Support 58
 

5.2.3 Building 4123 - KEWB Support 59
 

5.2.4 Building 4643 - KEWB Exhaust Stack 61
 

5.2.5 Building 4453 - AE6/L-85 Fuel Handling Facility 62
 

5.3 Sodium Cooled Graphite Reactors 63
 

5.3.1 Building 4143 - Sodium Reactor Experiment (SRE) Reactor 64

5.3.2 Building 4041 - Component Storage 95

5.3.3 Building 4686 - Temporary Hot Waste Storage 96 

5.3.4 Building 4163 - Box Shop / CERF 98

5.3.5 Building 4695 - Cold Trap Vault	 103

5.3.6 Building 4653 - Liquid Radioactive Waste Vault	 104

5.3.7 Building 4654 - Interim Waste Storage Area	 106

5.3.8 Building 4689 - Intermediate Contaminated Waste Storage 109
 

5.4 SRE: Additional Locations / Operations to Include in Site Profile 110
 

5.4.1 Building 4153 - Sodium Service Vault	 111
 

5.4.2 Building 4723 - Steam Cleaning Pad	 115
 

5.4.3 Building 4733 - Sodium Cleaning Pad	 117
 

5.4.4 Building 4773 - Sodium Pond (SRE Waste Disposal)	 119
 

5.5 Systems Nuclear Auxiliary Power (SNAP) Reactors	 122
 

5.5.1 Rockwell International Adopts Its Own “Acceptable Limits” for SNAP 122

5.5.2 Perspective: AEC-NASA Announce Joint Project - SNAP	 123
 

5.5.3	 Building 4010 - SNAP Experimental Reactor (SER) 124

 SNAP 8 Experimental Reactor (S8ER) 

5.5.4	 Building 4024 - SNAP 2 Developmental Reactor (S2DR) 128

 SNAP 10 Flight Systems Reactors 1,3,5 (S10FS1-3-5)

5.5.5	 Building 4028 - Shield Test Reactor (STR) 143

 Shield Test & Irradiation Reactor (STIR)

5.5.6 Building 4059 - SNAP 8 Developmental Reactor (S8DR)	 151
 

6.0 CRITICAL TEST FACILITIES	 161
 

6.1 SNAP Development Test Facilities	 161
 

6.1.1 Building 4373 - First SNAP Critical Facility	 162

6.1.2 Building 4012 - Second SNAP Critical Facility / NASA’s HMRFSR 173

6.1.3 Building 4019 - SNAP Flight System Critical Facility	 179

6.1.4 Building 4024 - SNAP Transient Test Facility	 182
 

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6.2	 SNAP: Additional Locations / Operations to Include in Site Profile 197
 

6.2.1	 Building 4013 - SNAP Assembly & X-Ray 198
 

6.2.2	 Building 4025 - Remote Handling / Mock-up 200
 

6.2.3	 Building 4027 - Shock / Vibration / Acceleration / Hazardous Storage 202
 

6.2.4	 Building 4032 - Thermal Vacuum Environmental Tests 206
 

6.2.5	 Building 4036 - SNAP Offices 210
 

6.2.6	 Building 4042 - Thermal Test Facility / LMFBR 212
 

6.2.7	 Building 4057 - SNAP / LMEC / ETEC / NASA 214

6.2.8	 Building 4228 - SNAP Critical (Formerly Building 4012) 216
 

7.0 CIVILIAN NUCLEAR POWER & OTHER CRITICAL TEST FACILITIES 217
 

7.1	 Building 4009 - Organic Moderated Reactor (OMR) 218

 Sodium Graphite Moderated Reactor (SGR)

 SGR Hot Lab*

 Van de Graaf Particle Accelerator* 

7.2	 Building 4100 - Advanced Epithermal Thorium Reactor (AETR) 223
 

8.0 NUCLEAR SUPPORT OPERATIONS 225
 

8.1 REACTOR FUEL MANUFACTURING 225
 

8.2 Building 4003 - Engineering Test Building (ETB)	 226
 

8.3 Building 4055 - Nuclear Materials Development Facility (NMDF)	 232
 

8.4 Building 4005 - OMRE / UCPFF / Molten Salt Tests* / NOx-Sox	 239
 

8.5 Building 4064 - Fuel Storage Facility	 245
 

9.0 DISASSEMBLY & EXAMINATION OF REACTORS
 

USED REACTOR-FUEL ASSEMBLY 251
 

9.1	 Building 4020 - Hot Laboratory / Component Development Hot Cell 252

Also see Section 7.1, Building 4009 SGR-Hot Laboratory 

10.0 FABRICATION, USE & STORAGE OF RADIOACTIVE SOURCES 271
 

10.1 Building 4029 - RMF	 272
 

11.0 PREPARATION OF RADIOACTIVE MATERIALS FOR DISPOSAL 277
 

11.1 Overview: Radioactive Materials Handling Facility (RMHF)	 278
 

11.1.1	 Buildings & Facilities Associated with the RMHF 280
 

11.1.2 Perspective: “Plan of Action for the RMHF” c. 1975 280
 

11.2 Building 4021 - Decontamination / Processing / Packaging	 286
 

11.3 Building 4022 - Radioactive Material Storage Vault	 298
 

11.4 Building 4034 - Shipping & Receiving	 305
 

11.5 Building 4044 - Offices	 307
 

11.6 Building 4075 - Low Specific Activity Waste Storage	 310
 

11.7 Building 4654 - Interim Radioactive Waste Storage Area 313
 

11.8 Building 4665 - Storage 317

11.9 RMHF Leach Field 318
 

(* Add t ona Hot Laboratory and Van de Graaf Part c e Acce erator, exc uded from 2006 S te Descr pt on.) 

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12.0 RMHF: Locations / Operations to Include in Site Profile 321
 

12.1 Building 4622 - Entry Point / Guard Shack	 322
 

12.2 Building 4563 - Storage Yard 324
 

12.3 Building 4621 - Radioactive Waste Storage	 325
 

12.4 Building 4658 - Entrance / Exit	 329
 

12.5 Building 4663 - Storage	 330
 

12.6 Building 4664 - Low-Level Radioactive Waste Incinerator	 333
 

12.7 Building 4668 - Storage & Sodium Cleaning	 338
 

13.0 RESEARCH ON REPROCESSING USED REACTOR FUEL 340
 

13.1 Building 4003 - Hot Cave 341
 

14.0 PARTICLE ACCELERATORS 347
 

14.1	 Building 4030 - Van de Graaf Particle Accelerator* 348

*Also see Section 7.1, Building 4009, Particle Accelerator #2 

15.0 RESEARCH USING RADIOISOTOPES 351
 

15.1 Building 4064 - Fuel Storage Facility 352
 

15.2 Building 4023 - Corrosion Testing Laboratory (SNAP) 359
 

16.0 MISCELLANEOUS OPERATIONS 363
 

16.1 Building 4583 - Conservation Yard 363

16.2 Building 4884 - Sodium Disposal Facility (Burn Pit) 365
 

17.0 ADDITIONAL FACILITIES & OPERATIONS TO INCLUDE IN SITE PROFILE 370
 

17.1 COAL GASIFICATION & LIQUEFACTION 371
 

17.1.1	 Building 4048 - Plant Development Unit (PDU) 372
 

17.1.2	 Building 4049 - Hydraulic Test Fac. / Piqua Test Loop / PDU 373

(Coal Gasification)

17.1.3 SSFL Energy Bowl Test Facility 375
 

17.2 SODIUM COMPONENTS TEST INSTALLATION (SCTI) 377
 

17.2.1	 Building 4356 - SCTI X-Ray 378
 

17.2.2	 Building 4006 - Sodium Laboratory 381

17.2.3	 Building 4355 - Control Center 383

17.2.4 Building 4357 - Heat Transfer Loop 385
 

17.2.4	 SSFL Energy Bowl - VTS III SABER 386

(Steam Accumulation Blowdown Evaluation Rig) 

17.3 SMALL COMPONENTS TEST INSTALLATION 388
 

17.3.1	 Building 4026 - SCTI Control Building 388
 

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17.4 LASER TESTING FACILITIES 389
 

17.4.2 Building LETF - The “RACHL” 390
 

17.5 MISCELLANEOUS OPERATIONS & FACILITIES 391
 

17.5.1 Building 4273/4283 - Radioactive Laundry 392
 

17.5.2 Building 4133 - Hazardous Waste Materials Facility (HWMF) 393

17.5.3 Building 4011 - Radiation Instrument Calibration Laboratory 396
 

17.5.4 Building 4641 - Shipping & Receiving 399

17.5.5 Building 4066 - Instrumentation Calibration / Repair 400
 

18.0 ASSOCIATED FACILITIES 402
 

18.1 Downey Facility 402
 

18.2 Canoga / Van Owen Facility 403
 

18.3 DeSoto Facility 406
 

19.0 CONCLUSION 417
 

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LIST OF ACRONYMS & ABBREVIATIONS
 

ABRWH Advisory Board on Radiation and Worker Health 

ABTSWH Advisory Board on Toxic Substances and Worker Health 

AEC U.S. Atomic Energy Commission 

ARRA American Recovery and Reinvestment Act 

AETR Advanced Epithermal Thorium Reactor 

AI Atomics International Division, North American Aviation 

Atomics International Atomics International Division, North American Aviation 

CDPHE California Department of Public Health and Environment

CERCLA Comprehensive Environmental Response, Compensation, and

Liability Act 

Ci Curie 

COC Contaminants of Concern 

CRIEPI Central Research Institute of the Electrical Power Industry 

DEEOIC Division of Energy Employee Occupational Illness Compensation

D&D decontamination and decommissioning

DD&D decontamination, decommissioning and demolition

DHS Department of Health Services

DOE U.S. Department of Energy

DOD Department of Defense

DOL U.S. Department of Labor

dpm/100 cm2 disintegrations per minute per 100 square centimeters

DTSC Department of Toxic Substances Control 

EEOICPA Energy Employee Occupational Illness Compensation Program Act 

EPA U.S. Environmental Protection Agency 

ESSAP Environmental Survey and Site Assessment Program 

ETEC Energy Technology Engineering Center 

FAB Final Adjudication Branch 

FCEL Fast Critical Experiment Laboratory

FOIA Freedom of Information Act 

HEPA high-efficiency particulate air

HGL HydroGeoLogic, Inc.

HSA Historical Site Assessment

HWMF Hazardous Waste Materials Facility

HQ headquarters 

KEWB Kinetic Experiment Water Boiler 

KHI Kawasaki Heavy Industries

kW kilowatt 

kWt thermal kilowatt 

LLTR Large Leak Test Rig 

LMEC Liquid Metal Engineering Center

LMFBR Liquid Metal Fast Breeder Reactor 

μR/hr micro roentgen per hour

mR/hr milli roentgen per hour

MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual 

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MCP 

MDA 

MFP 

MOU 

MWd 

maximum permissible concentration

minimum detectable activity

“mixed fission products”

memorandum of understanding

megawatts per day

NASA 

NBZ 

National Aeronautics and Space Administration

Northern Buffer Zone

NIOSH 

NMDF 

NRC 

NSA 

National Institute for Occupational Safety and Health

Nuclear Materials Disposal Facility

Nuclear Regulatory Commission

nuclear safety analysis 

ORISE Oak Ridge Institute for Science and Education

PDU 

pCi/g

POC 

Plant Development Unit

picocuries per gram

Probability of Causation

R/A

RCRA 

RFI 

RHB 

RMHF 

RMF 

radioactive

Resource Conservation and Recovery Act

RCRA Facility Investigation

Radiological Health Branch

Radioactive Material Handling Facility

Radioactive Measurements Facility 

S2DR SNAP 2 Demonstration Reactor

S8DR 

SB 

SNAP 2 Development Reactor

Senate Bill

SBZ Southern Buffer Zone

SC&A Sanford Cohen & Associates

SCTI 

SCTL 

SETF 

SGR 

SHEA 

SNAP 

SNAPFS 

SNAPTRANS 

Sodium Components Test Installation

Small Components Test Laboratory

SNAP Environmental Test Facility

sodium graphite reactor

Safety Health and Environmental Affairs

Systems for Nuclear Auxiliary Power

SNAP Flight Systems

SNAP Transient Reactor

SPTF 

SRE 

SSFL 

SEM 

Sodium Pump Test Facility

Sodium Reactor Experiment

Santa Susana Field Laboratory

Site Exposure Matrix 

TM Technical Memorandum 

TO Task Order

TRU Transuranic 

TRUMP-S TRU Management by Pyropartitioning - Separation

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1.0 Introduction 

CORE ADVOCACY FOR NUCLEAR & AEROSPACE WORKERS 

CORE Advocacy for Nuclear & Aerospace Workers provides advocacy and Authorized 

Representation for Santa Susana Field Laboratory (SSFL), Canoga, Van Owen, DeSoto, and 

Downey claimants of the Energy Employee Occupational Illness Compensation Program 

(EEOICPA or, “the Act”). We are members of the Alliance of Nuclear Worker Advocacy Groups 

(ANWAG), the Energy Employee Claimant Assistance Project (EECAP), and the former Division 

of Energy Employee Occupational Illness Compensation Interim Advisory Board (DIAB). CORE 

Advocacy participates in annual meetings held in the spirit of improving EEOICPA, which are 

attended by advocates and members of federal agencies tasked with EEOICPA’s administration. 

CORE Advocacy supports full disclosure of SSFL site operations and history to better 

understand work environment and practices, and to effectively implement EEOICPA legislation. 

Our extensive research into SSFL and contractor history is based on official documentation, 

original facility records, and technical reports authored by the Atomic Energy Commission 

(AEC), Department of Energy (DOE), the National Aeronautics and Space Administration 

(NASA), Department of Defense (DOD), state and federal regulatory agencies, North American 

Aviation (NAA) Atomics International, Rocketdyne, Rockwell International, Energy Systems 

Group, Boeing, and actual worker records generated during employment. 

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1.1 The Purpose of the 2016 SSFL Site Description 

CORE Advocacy for Nuclear and Aerospace Workers submits the following 2016 Santa Susana 

Field Laboratory (SSFL) Site Description (“2016 SSFL Site Description”) for review by the 

National Institute for Occupational Safety and Health (NIOSH) and the Presidential Advisory 

Board on Radiation and Worker Health (ABRWH). This document is intended to revise and 

replace the 2006 Site Introduction / Site Description, which are currently considered part of the 

SSFL Site Profile. 

As indicated by NIOSH, the purpose of a Site Profile is to provide an accurate physical 

description of the work site’s appearance and layout, work processes, types of materials used at 

the site, potential sources of radiation, exposure monitoring practices employed over time, and 

other details that were/are relevant to the worksite. In addition, a Site Profile is intended to 

assist NIOSH with completing dose reconstructions when there is a need to understand or add 

to an energy worker’s personal exposure information. A Site Profile should provide a robust, 

comprehensive and accurate depiction of site operations. It should be consistent with historical 

facility data, employment records, incident reports, and current environmental and scientific 

studies. 

The SSFL Site Profile was issued as six separate Technical Basis Documents (TBDs) numbered 

ORAUT-TKBS-0038-1 through ORAUT-TKBS-0038-6. It has not been updated or revised since 

2006. Several problems have been identified but remain unaddressed, including an insufficient 

description of operations, processes, incidents, and the omission of several important

radiological facilities. The 2006 Site Profile provides an inaccurate depiction of facility operations 

that calls into question current environmental data currently used in the dose reconstruction 

process. In addition, several of the issues identified have direct bearing on the Special Exposure 

Cohort (SEC). 

NIOSH indicates that that TBDs “are general working documents that will be revised in the 

event additional relevant information is obtained.” CORE Advocacy has obtained the following 

additional relevant information about SSFL for inclusion in the Site Profile, beginning with the 

Site Introduction (ORAUT-TKBS-0038-1) and Site Description (ORAUT-TKBS-0038-2).

Thank you for the opportunity to provide additional information on behalf of former personnel of 

the Santa Susana Field Laboratory (SSFL) and its associated facilities, and Energy Employee 

Occupational Illness Compensation Program Act (EEOICPA) claimants. 

D’Lanie Blaze

CORE Advocacy for Nuclear & Aerospace Workers

COREAdvocacy.org 

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1.2 Information Provided in the 2016 SSFL Site Description

For the purposes of this document, the word “facility” is used as a general term for an area, 

building, or group of buildings that served a specific purpose at SSFL or its associated sites. It 

does not necessarily connote an “atomic weapons employer facility” or a “Department of Energy 

(DOE) Facility” as defined in the Energy Employee Occupational Illness Compensation Program 

Act (EEOICPA) under 42 U.S.C. § 73841(5) and (12). 

This document follows the original format of the 2006 Site Profile; buildings/facilities are listed 

according to operations. In each category, additional relevant buildings and facilities that were 

omitted from the 2006 document are described. Section 17.0 provides information about entire 

site projects and operational programs that should be included to the Site Profile to ensure an 

accurate site characterization. 

The 2016 SSFL Site Description contains the following information about SSFL:

• Corrects and clarifies facility nomenclature and years of operation

• Describes site’s physical appearance, history, DOE operations & processes, environmental 

and worker monitoring practices

• Summarizes incident reports involving each location

• Provides a partial list of Visitor Log Entry Building Account Codes to locate monitored workers 

• Summarizes radionuclides of concern at each building/facilities, according to EPA 

• Provides EPA MARSSIM Classifications for each location 

“MARSSIM" refers to Multi-Agency Radiation Survey & Site Investigation Manual 

Classifications. The United States Environmental Protection Agency (EPA) assigned MARSSIM 

Classifications to each building/facility contained in the 2016 SSFL Site Description. The United 

States Environmental Protection Agency (EPA) based these classifications on historical data

documenting building operations, incidents, and other information that may aid NIOSH in 

assessing the potential radiological exposures among workers during past site operations. 

MARSSIM Classifications are assigned based on the following criteria: 

• Class One: Applies to areas with the highest potential for contamination that meet the 

following criteria: impacted; has had, or currently has the potential for delivering dose above 

the release criterion; has had, or currently has the potential for small areas of elevated activity; 

and where insufficient evidence exists to support a lower classification (Class 2 or Class 3) 

under MARSSIM.

• Class Two: Applies to areas that are impacted; has had, or currently has low potential for 

delivering a dose above the release criterion; has had or currently has little or no potential for 

small areas of elevated activity.

• Class Three: Applies to areas that are impacted; has had, or currently has little or no potential 

for delivering a dose above the release criterion; and has had or currently has little or no 

potential for small areas of elevated activity. 

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1.3 Source Documents Used in the 2016 SSFL Site Description 

In 2009, EPA conducted a radiological characterization study within the area of SSFL known as 

“Area IV” and the Northern Buffer Zone (NBZ). This combined study-area is hereafter called the 

“Area IV Study.” 1 The Area IV Study was conducted with the participation of DOE and its 

contractor, The Boeing Company (“Boeing”). It was prepared by HydroGeologic, Inc. (HGL). 

EPA’s Area IV Study consisted of a Historical Site Assessment (HSA), gamma scanning of 

accessible areas, geophysical surveys, and evaluation of past soil sampling results. The HSA

component of the study involved a comprehensive investigation that identified, collected, 

organized and evaluated historical information relevant to nuclear research operations that 

resulted in radiological contamination of the Area IV Study area. EPA published the HSA in 2012 

as an eight-volume set of DVDs. 

The EPA conducted a physical inspection of the site and a review of documents that were 

obtained through a series of formal information requests sent to Boeing, DOE, the Nuclear 

Regulatory Commission (NRC), the California Department of Public Health (CDPH) under § 

104(e) of the Comprehensive Environmental Response Compensation and Liability Act 

(CERCLA), and the National Aeronautics and Space Administration (NASA). 

In addition, EPA directed Boeing to identify and provide pertinent documents within a number of 

document databases comprising approximately 1.4 million documents relating to all areas of the 

SSFL including Area IV and some off-site facilities. In response to EPA’s request, Boeing 

provided an Incident Database containing original incident reports documenting a variety of 

incidents over the course of site history, many of which resulted in releases of radionuclides to 

the environment and/or worker exposure. 

EPA reviewed individual documents and photographs, first screening over 80,000 documents 

amassed for the project. In addition, the documents requested from the document databases 

comprising approximately 1.4 million documents were reviewed. EPA’s screening effort 

produced 720 documents relevant to past operations at SSFL, which warranted in-depth 

evaluation. In addition, EPA and DOE jointly conducted over 300 formal employee interviews as 

part of the HSA. The interviews (in their entirety) are provided on the disk that accompanies the 

2016 SSFL Site Description. 

The quality of data gathered by EPA exceeds NIOSH’s criteria for a robust, comprehensive Site 

Profile. CORE Advocacy has drawn largely from the EPA HSA and other official historical 

documents, including employment records. In the interest of worker confidentiality, cited worker 

records and/or Case File numbers will be provided to NIOSH and/or ABRWH upon request. 

1 Final Technical Memorandum, Historical Site Assessment, Santa Susana Field Laboratory Site, Area IV Radiological 

Study, 2012. Environmental Protection Agency (EPA) and HydroGeologic, Inc (HGL) 

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1.4 Acceptance of Source Documents by Federal Agencies 

As summarized in the preceding section, in response to EPA’s directive to provide information 

for use in the Area IV Radiological Study and HSA, DOE and Boeing provided historical 

documentation authored by the Atomic Energy Commission (AEC), DOE, facility contractors 

over the course of site operations, NASA, and state and federal regulatory agencies. In addition, 

Boeing provided EPA with an Incident Database that contained hundreds of incident reports 

pertaining to SSFL (all areas of operation), Canoga, VanOwen, DeSoto, and Downey Facilities. 

There are indications that Boeing may have provided an identical Incident Database to NIOSH. 

In addition, EPA and DOE conducted over 300 formal interviews with former personnel. EPA

cross-referenced the Incident Database, historical documents, and worker interviews during the 

site investigation. 

Based on DOE and Boeing’s participation in the Area IV Study and HSA, and what appears to 

be NIOSH’s possession of the Incident Database, it is reasonable to expect that DOE, NIOSH 

and Boeing will accept the same information, which is cited herein, for inclusion to the SSFL Site 

Profile.

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2.0 OVERVIEW: The 2006 Site Profile 

This section provides an overview of the 2006 Site Introduction (ORAU-TKBS-0038-1, “Atomics 

International - Introduction”) and the 2006 Site Description (ORAUT-TKBS-0038-2, “Energy 

Technology Engineering Center Site Description”). 

In 2008, ABRWH directed Sanford Cohen and Associates (SC&A) to evaluate all six Technical 

Basis Documents (TBDs) that comprise the 2006 Site Profile. SC&A identified numerous issues 

that were likely to interfere with dose reconstruction, many of which have direct bearing on the 

Special Exposure Cohort (SEC). SC&A provided recommendations based on its review. 

SC&A noted that the 2006 Site Profile contained an incomplete and insufficient characterization 

of facilities, operations, and processes associated with DOE activities at SSFL. In addition, 

SC&A noted that the presentation of incidents and environmental releases was incomplete and 

inaccurate. These inaccuracies call into question the validity of environmental data, and cast 

doubt upon the relevance of dose reconstruction outcomes for employees of SSFL and its 

associated facilities. In addition, SC&A brought attention to inaccurate site and contractor 

nomenclature, and incorrect years of facility operations.1 2 

To date, it does not appear that SC&A’s observations have been fully investigated; their 

recommendations have not been implemented. CORE Advocacy has identified numerous other 

issues that are in need of correction. 

2.1 Site Nomenclature: 2006 Site Introduction & Site Description 

SSFL began site operations in 1948. Currently, SSFL Area IV is considered the “covered area” 

under EEOICPA, beginning in 1955. Area IV is comprised of SSFL’s westernmost 290 acres. 3

In 1966, the AEC established the Liquid Metals Engineering Center (LMEC) on a 90-acre area 

located inside Area IV. This area became known as the Energy Technology Engineering Center 

(ETEC) in 1978. “ETEC” and “Area IV” are the most common terms used to identify the 

“covered area” under EEOICPA. 

The 2006 Site Introduction and Site Description state: “This document describes the features 

and history of the Energy Technology Engineering Center (ETEC) ... The AEC established 

ETEC in 1966.” The statement implies that ETEC is the extent of the covered area and that 

EEOICPA coverage does not begin until 1966, likely creating confusion about the site’s 

nomenclature, years of operations, and locations of some facilities. 

The SSFL Site Profile should indicate that it contains information about DOE operations 

involving Area IV, ETEC, and LMEC, without exception. Both should clearly indicate that 

EEOICPA coverage begins in 1955. 

1 SC&A, “Review of the NIOSH Site Profile for the Santa Susana Field Laboratory,” August, 2008. Task Order No. I 

SCA-TR-TASK1-0027 

2 SC&A, “Issues Resolution Matrix for the Santa Susana Site Profile Review,” November, 2008. 

3 http://webapps.dol.gov/FederalRegister/HtmlDisplay.aspx?DocId=26761&AgencyId=13 

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2.2 Overview: 2006 ETEC Site Description - ORAUT-TKBS-0038-2 

The 2006 Site Description is severely flawed. Below is an overview of the general problems, 

which are explained in detail in this section.

Questionable Source Data 

- Boeing assumed SSFL operations in 1996, nearly 48 years after the site’s inception (1948). 

However, the majority of source documents cited in the 2006 Site Profile consist of Boeing’s 

own summary data and reports authored by Boeing contractors, after 1996.

- Historical facility documents and technical reports authored by DOE and Boeing’s 

predecessors are abundant and widely available. Yet, citations to historically significant 

documents are severely lacking throughout the 2006 Site Profile.

- When the post-1996 summary data and Boeing-contracted reports are compared alongside 

original historical documents, persistent conflicts and irregularities are apparent. The 

post-1996 summary data and Boeing-contracted reports consistently conflict with original 

facility documents, employee records, and the Incident Report database.

- It is reasonable to expect that any summary data provided by Boeing would be accompanied 

by supportive, authentic historical documents from a wide variety of sources.

Misrepresentation of Buildings/Facilities, Operations & Processes 

- In many instances, the years of facility, reactor, building, and program operations contained in 

the 2006 Site Profile, and the description of operations, are incomplete or incorrect.

- Descriptions of processes, materials, incidents, and releases are incomplete or incorrect.

Buildings, Facilities, Operations Excluded from 2006 Site Profile 

- Numerous radiological facilities have been excluded in their entirety, along with all 

corresponding incident reports, descriptions of operations, documented releases, and 

environmental data.

“Non-Radiological” Facilities Involved in Radiological Processes 

- Several designated “non-radiological” facilities have documented involvement in radioactive 

processes. These facilities have been excluded from the 2006 Site Profile, along with all 

corresponding incident reports, descriptions of operations, documented releases, and 

environmental data. 

- These locations may have lacked proper licensing and radiological use authorizations. In 

addition, they may have failed to meet criteria for safe handling of radioactive materials.

- There are strong suggestions that worker monitoring practices were determined based on a 

worker’s designated work location. An employee assigned to a “non-radiological” facility may 

not have been adequately monitored for radiation exposure or considered to be a “radiation 

worker,” yet historical documentation shows the assigned location regularly participated in 

radiological processes. This problem, especially when combined with worker rotation, has 

resulted in an inability to reliably determine a worker’s likelihood of exposure based on work 

location designation, job title, assignment to a “non-radiological” location or duties, or a lack of 

dosimetry data in employment records. In addition, this problem likely creates difficulty in 

assigning appropriate co-worker data, when the need arises.

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3.0 SSFL: Site Description & General Information 

This document describes the features and history of Department of Energy (DOE) facilities at 

Santa Susana Field Laboratory (SSFL), including the site areas or buildings, site processes, 

periods of operation, radionuclides of concern, incidents resulting in releases to the environment 

or worker exposure, and other information that is relevant to estimating worker radiation dose. 

3.1 Brief Description and History 

SSFL occupies 2,850 acres of rocky terrain located in the Santa Susana mountains between 

Simi and San Fernando Valleys, approximately 30 miles northwest of Los Angeles, California 

(Ventura County). The site’s elevation is 1,880-2,150 feet feet above sea level. The terrain 

consists largely of weathered bedrock and alluvium that have been eroded primarily from the 

surrounding Chatsworth and Santa Susana formations. Several geological faults cross the area. 

The climate is classified as Mediterranean Subtropical, corresponding to an average 

temperature of 50 degrees Fahrenheit in the winter and 70 degrees Fahrenheit in the summer. 

Rainfall averages approximately 18 inches per year. 

SSFL was established by North American Aviation (NAA) in 1947 to meet requirements for a 

remote field test laboratory to test static-fire large rocket engines and to conduct nuclear 

research for the Atomic Energy Commission (AEC, DOE’s predecessor agency). On February 

28, 1948 the AEC entered into a contract with NAA to perform research and to develop nuclear 

reactors. NAA was permitted its discretionary use of its facilities and locations owned by the 

company or leased by AEC to perform functions affiliated with fulfilling its contractual 

obligations. AEC-sponsored nuclear research began under contract with NAA at SSFL in 1953. 1

SSFL is divided into four areas of operation (Areas I, II, III and IV). Currently, under EEOICPA, 

Area IV is the “covered area” from 1955-present. Area IV is comprised of the site’s westernmost 

290-acre area. A Special Exposure Cohort (SEC) is in effect from 1955-1958 and 1959-1964. 

Expansion of the SEC into subsequent years is being considered by NIOSH and ABRWH. 

The Liquid Metal Engineering Center (LMEC) was created in 1966 as a government-owned and 

contractor-operated organization to provide development of non-nuclear testing of Liquid Metal 

Reactor (LMR) components, and to establish the Liquid Metal Information Center (LMIC) for the 

AEC’s Liquid Metal Fast-Breeder Reactor (LMFBR) program. LMEC was renamed the Energy 

Technology Engineering Center (ETEC) in 1978 to reflect DOE’s desire to broaden its mission 

beyond the LMFBR program. 

LMEC-ETEC consisted of a 90-acre area located within the confines of Area IV. Use of the term, 

“ETEC” is understood to encompass all DOE operations associated with Area IV, ETEC and 

LMEC. 

ETEC conducted a broad range of energy-related research, testing and development projects 

that included component testing for the U.S. liquid metal reactor program. Liquid metal test 

1 Division of Energy Employees Occupational Illness Compensation (DEEOIC) Peter M. Turcic Memorandum to 

Christy Long, District Director of Seattle, Subject: Atomics International and Energy Technology Engineering Center, 

September 7, 2005 

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facilities were maintained and operated for evaluating components such as heat exchangers, 

steam generators, pumps, valves, piping, vessels and instrumentation. In addition to the reactor 

component testing program, ETEC provided engineering support, technical management and 

monitoring for a number of DOE solar, conservation, geothermal, coal liquefaction and 

gasification, laser, and fusion energy programs. 

CORE Advocacy recommends that all Technical Basis Documents (TBDs) related to SSFL’s Site 

Profile be updated to correct and clarify the use of site nomenclature, the entirety of important 

radiological facilities located throughout Area IV, the years of facility operations, and the years of 

EEOICPA coverage. 

3.2 AEC’s Support of Space-Nuclear Research 

During the 1950’s the Atoms for Peace Initiative resulted in AEC’s pursuit and sponsorship of 

civilian uses for atomic power. AEC Chairman, Glenn Seaborg referred to the “marriage 

between space and the atom” as “something that was bound to occur.” NASA and AEC formed 1

joint committees and enthusiastically pursued funding and projects to foster space-nuclear 

propulsion research; SSFL was well suited for this type of research and development (R&D). 

In 1955, coinciding with NAA’s creation of Rocketdyne as its own division, AEC invited industry 

to bid on the development of a space nuclear reactor system. A concept by NAA Atomics 

International was chosen in 1956. Shortly thereafter, a need for a small atomic battery to power 

space missions was recognized. Under AEC-NASA-NAA contract, the Systems for Nuclear 

Auxiliary Power (SNAP) program resulted in several zirconium hydride uranium reactors (SNAP

2, 6, 8 and 10) that were researched and tested at SSFL.2 In 1965, the SNAP 10-A Space 

Nuclear Reactor was successfully launched into orbit (where it remains). In its 1965 Annual 

Report, NAA heralded SNAP’s success as the result of the company’s philosophy of 

3 interdivisional collaboration. AEC celebrated NASA’s commitment to perfection, which inspired 

AEC’s approach to devising nuclear safety standards for use in the operation of terrestrial power 

plants here on earth. 

Throughout SSFL’s well documented history, NASA is shown to have operated in Area IV while 

DOE is documented as having operated in Areas I, II and III. In addition, DOE facilities, 

operations and proprietary interests in Areas I, II and III fulfill statutory criteria used to determine 

a DOE Facility under the Act, pursuant to U.S.C. § 73841. 

1 Roger D. Launius, Powering Space Exploration: U.S. Space Nuclear Power, Public Perceptions, and Outer 

Planetary Probes, Smithsonian Institution, 6th International Energy Conversion Engineering Conference (IECEC),

July 28-30, 2008, Cleveland, Ohio. 

2 George P. Dix and Susan S. Voss, The Pied Piper - A Historical Overview of the U.S. Space Power Reactor 

Program 

3 NAA, Annual Report, 1965. 

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3.3 Corporate Mergers & Nomenclature 

Several corporate mergers and organizational changes occurred over the years. In 1967 North 

American Aviation (NAA) merged with Rockwell Standard to become North American Rockwell. 

In 1973 the corporate name changed to Rockwell International (RI). RI with Atomics 

International and Rocketdyne continued to exist as independent divisions until 1984, although 

“Rocketdyne” or “Rockwell” were common terms used to identify the contractor during this 

period. In 1984, Atomics International was absorbed by the Rocketdyne division. The Boeing 

Company purchased RI in 1996, becoming Boeing North American (BNA), and Rocketdyne 

became a division of Boeing. During this period, “Rocketdyne” was typical terminology used to 

identify the contractor. Currently, any of these terms, including “ETEC” are used to identify both 

the contractor and covered area at SSFL. 

The review of historical documents authored over the course of several eras of facility 

operations can be confusing because the contractor is routinely identified by any combination of 

the above-referenced company names. For instance, an incident report authored in the 1960’s 

may identify “Atomics International,” but a radiological investigation of the site conducted in the 

1980’s may refer to “Rockwell” or “Rocketdyne” when referencing the same report, while a D&D 

contract authored in the 1990’s may refer to “Boeing.” 

For the purposes of this report, any references to North American Aviation, Atomics 

International, Rocketdyne, Rockwell, Rockwell International, Energy Systems Group (a division 

of Rockwell International), Boeing North American, or Boeing refers to the DOE Contractor. 

References to Area IV, ETEC and LMEC refer to the covered area, under EEOICPA. 

CORE Advocacy recommends that an explanation and clarification of corporate mergers and 

changes in DOE contractors be reflected in every Technical Basis Document (TBD) associated 

with SSFL throughout the SSFL Site Profile. 

3.4 Environmental Monitoring and Sampling Practices 

On July 12, 1989 the U.S. EPA assessed the relative magnitude of health hazards, health risks, 

past, present, and future environmental problems, and how those concerns may be addressed 

with the site contractor.1 EPA reviewed previous Rockwell-Rocketdyne environmental reports, 

contractor reports, DOE site reviews, and conducted interviews with laboratory personnel to 

review procedures for sampling. In addition, EPA visited specific locations at the site, and took 

samples and measurements. 

EPA questioned the validity of some, if not all, of the environmental data. According to laboratory 

personnel, the laboratory had never had a thorough review or audit by Rockwell-Rocketdyne or 

DOE (DOE’s environmental audit had yet to be finalized at the time of the EPA site visit). 

1 Dempsey, Gregg: “Site Visit to Santa Susana Field Laboratory,” EPA Memorandum to Daniel M. Shane, On-Scene 

Coordinator, Emergency Response Unit. July 28, 1989 

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3.4.1 Soil - Questionable Sampling Practices Observed by EPA 

EPA found that the SSFL laboratory personnel used poor methods for assessing environmental 

radioactivity, including procedures that resulted in temperatures sufficient to volatilize most manmade radionuclides of concern, including cesium-137 and strontium-90. In addition, the 

Rocketdyne procedure had highly variable results due to attempt to make corrections, which 

were inadequate. Further, the SSFL laboratory personnel were consistently applying the wrong 

procedure to measure alpha and beta activity in the soil (the procedure applied was applicable 

to the analysis of americium-241 by alpha spectroscopy, which is entirely different). Laboratory 

technicians could not provide EPA with any documentation or references on the procedure. 

Further, EPA discovered that spike samples had never been prepared, or run through a 

procedure to provide any internal quality control. EPA concluded that gross alpha and beta data 

on soil, even though it has indicated some radiation areas on the site, is not a true 

representation of conditions that were present in the environment. 

3.4.2 Water - Questionable Sampling Practices Observed by EPA 

EPA then inspected water monitoring practices, and again found that alpha and beta selfabsorption was likely to be a problem. EPA doubted the validity of the Rockwell-Rocketdyne 

water analyses, as well, and noted that Rocketdyne had never had a procedure for testing for 

tritium. 

3.4.3 Vegetation - Questionable Sampling Practices Observed by EPA 

Vegetation samples were collected by Rocketdyne until 1986, but was stopped two years after 

an internal SSFL review determined that problems existed with alpha and beta counting, and 

that changes should be made. EPA reviewed the procedure and found it to be similar to the soil 

counting procedure, in that vegetation was essentially “ashed” before counting, and only one 

gram of ash was analyzed. 

EPA noted that the procedure manual instructed laboratory technicians to “gently wash the 

vegetation with warm tap water to remove external foreign matter.” EPA points out that past 

operations that had produced airborne contamination that settled on the surface of the 

vegetation (instead of being absorbed through the roots) would be washed off before counting. 

Or, it would likely be volatilized off during the ashing process (500 ̊ C). EPA noted that animals 

at SSFL were not sampled. 

3.4.5 Environmental Dosimetry - Questionable Monitoring Observed by EPA 

The SSFL laboratory also provided environmental thermoluminescent dosimetry for the facility 

and offsite areas. EPA noted that certain questionable practices were alluded to, in the 

environmental report. The first was that data obtained by dosimeters was normalized to a 1,000-

foot altitude, by using an adjustment factor to 15 mR/1000 ft. elevation difference to obtain site 

averages. EPA spoke to two nationally recognized dosimeter experts, and neither had ever 

heard of such a practice. Both experts felt the normalization was meaningless. 

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EPA noted that in both the calendar year of 1987 and the unpublished calendar year of 1988 

SSFL environmental reports, comparisons for the dosimeters placed by the State of California 

and a DOE inter-comparison project were “not available” for inclusion at the time the report was 

published. Bill Watson of CDHS EMB assured EPA that the data was available, and provided to 

Rockwell-Rocketdyne. Even if the data was unavailable for inclusion in the previous year’s 

report, it should have been added as an addendum for the following year’s report. The 

unpublished 1988 report did not contain information about 1987 emissions, leading EPA to 

conclude that the SSFL dosimetry program might not compare favorably with the other groups. 

Systematic error that might be present in dosimetry analyses might make SSFL dosimetry data 

look comparable to itself but still may make these analyses invalid or suspect. EPA 

recommended more thorough review. 

EPA commented on the audit conducted by DOE in February, 1989. The preliminary copy was 

supplied to EPA to assist in the 1989 site visit. EPA notes that DOE attempted to review many 

aspects of the SSFL Environmental Program within the 1989 document, and EPA echoed their 

concerns about the well and air sampling at SSFL and offsite. EPA agreed with DOE’s 

conclusions that well and air sampling, in addition to environmental sampling in general, needed 

to be reviewed for adequacy. According to EPA, DOE also identified some problems in the 

Radiological Laboratory but did not do an extensive review. 

EPA echoes DOE’s concerns that the lack of a meteorological tower onsite presented problems. 

SSFL used the EPA code AIRDOS to define dose to affected offsite areas. However, tower 

information used was from the Burbank Airport, nearly 25-miles away from the site. 

EPA’s conclusions at the end of its 1989 site visit were that the SSFL sampling, placement of 

sample locations, and analyses cannot guarantee that past actions have not caused offsite 

impacts. EPA advised that, if the environmental program were to stay uncorrected, SSFL could 

not guarantee that unforeseen or undetected problems onsite would not impact the offsite 

environment in the future. 

EPA made it clear: “Rocketdyne does not have a good “handle” on where radiation has been 

inadvertently or intentionally dumped onsite. Most of the evidence of on-site spills is 

incompletely documented or anecdotal.” 

3.4.6 DOE Environmental Audit & Tiger Team Observations on Environmental Data 

1 In 1989 and 1991, DOE’s Office of Environmental Audit and Tiger Team 2 concluded that worker 

and environmental monitoring practices at SSFL were inadequate and provided unreliable dose 

assessments among the worker population throughout the site’s operational history. 

Both the Office of Environmental Audit and the Tiger Team concluded that SSFL’s lack of a 

meteorological tower presented inaccurate dose assessments following radiation releases. Both 

noted that on-site sampling locations were inappropriate because they had been chosen based 

1 U.S. DOE Office of Environmental Audit, “Environmental Survey Preliminary Report, U.S. DOE Activities at SSFL,” 

1989. 

2 DOE Office of Environment, Safety and Health (ES&H), Tiger Team Assessment - ETEC, April 1991. 

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on the intended locations of reactor operations. However, the selected locations for reactor 

operations were finalized, the sampling locations did not change accordingly. 

Both research teams noted no formal chain of custody for sampling data; no quality

environmental surveillance program had been implemented; insufficient monitoring of 

radioactive airborne particulate releases created consistent difficulties in estimating worker 

dose; and failure to adequately monitor radionuclide emissions from point sources (including 

those from remedial actions) resulted in misleading and inappropriate worker dose estimations. 

Moreover, both research teams called attention to consistently poor filter changing-and-handling 

practices, which resulted in a loss of particulate matter and inaccurate radioactivity 

measurements. 

Both the Office of Environmental Audit and the Tiger Team noted Rockwell International’s 

unilateral decision to cease all manner of soil and water sampling by notifying the Nuclear 

Regulatory Commission (NRC) that future sampling would only occur on an “as needed” basis. 

Rockwell International failed to inform the NRC of legitimate reasoning behind ceasing the 

environmental sampling program, and did not provide a definition of its conditional approach to 

sampling going forward. It is interesting to note that the timing of this decision coincided with 

EPA’s observation that incorrect and insufficient practices were being used for soil sample 

analysis. 

DOE’s Office of Environmental Audit and the Tiger Team address DOE operations at SSFL that 

occurred outside Area IV; worker rotation site-wide; joint DOE-NASA projects that involved 

employees from both divisions of North American Aviation (Atomics International and 

Rocketdyne); and a degree of site operations verses a lack of adequate environmental and 

worker monitoring practices that call into question the validity of dose reconstruction outcome 

for SSFL workers. 

4.0 Nuclear Research and Development 

Nuclear R&D increased rapidly at SSFL from 1953 into the late 1960’s. Ten experimental 

nuclear reactors are typically acknowledged as having functioned at SSFL. However, NASA

operated a nuclear reactor in Building 4012 (SNAP Critical Test Facility). The HMRFSR 

operated from 1970-1972. It was excluded from the 2006 Site Profile, and is described in 

Section 6.1.2, pertaining to Building 4012. 

The 2006 Site Profile indicates that by 1988, all nuclear operations at SSFL had ceased. 

However, a document authored in 1990 indicates that Building 4009 (OMR/SGR Critical Facility 1

described in Section 7.1) was still being used for Rocketdyne projects requiring radioactive 

materials. In addition, the 2006 Site Profile omitted a second Hot Laboratory in Building 4009, 

as well as a second Van de Graaf Particle Accelerator at the same location. 

1 Internal Letter, Rockwell International, “DOE Authority for Release of Certain Facilities at SSFL,” to A.J. Adduci, 

Nuclear Energy Division, DOE. January 11, 1990. Document retrieved from publicly accessible historical archive 

courtesy of California Department of Toxic Substances Control (DTSC). File: HDMSe00413848.pdf 

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There were seven criticality test facilities at SSFL (i.e. facilities housing operations involving 

masses of fissionable material capable of sustaining a nuclear chain reaction). Reactors and 

criticality test facilities located at SSFL are listed below in the following tables. 

Other nuclear facilities in Area IV included the Radioactive Materials Handling / Disposal Facility 

(RMDF / RMHF). The 2006 Site Profile excluded several important buildings and facilities 

associated with this facility, including a low-level radioactive waste incinerator that functioned 

from 1964 into the 1980’s. This particular facility was among the most notable generators of 

radioactivity at SSFL. 

4.8.1 ISOTOPES OF CONCERN 

The 2006 Site Profile addresses potential radiological contaminants associated with specific 

nuclear operations at SSFL in terms of isotopes of concern. The guidance for the preparation of 

a Site Profile (ORAU 2005) requires for a table to include dose reconstruction parameters such 

as the solubility type and assumed particle size of the contaminants, although the same 

information is also required in Section 5, Internal Dosimetry. According to Section 2 of the 2006 

Site Description, the section is for informational use only and the data provided in Section 5 of 

the 2006 Site Description “should be used for dose reconstruction, regardless.” However, the 

2006 Site Description specifies that facility-specific solubility and particle size data for SSFL

“has not been found.” 

Given the number of radiological facilities that have been excluded from the 2006 Site Profile, 

the insufficient information provided about years of reactor operations, incidents, releases, and 

building operations, CORE Advocacy recommends careful evaluation of the Isotopes of 

Concern alongside the information provided in the following tables, which was sourced from 

EPA’s Area IV Study (2009). 

4.8.2 MAGNITUDE OF SITE ACTIVITY 

The 2006 Site Profile provides a table intended to “provide a perspective of the magnitude of the 

nuclear operations at SSFL,” in accordance with the Site Profile Preparation Guide (ORAU, 

2005). The guide requires that the magnitude of operations is expressed in radioactivity (Ci) of 

the isotopes of concern. Again, the same information is required in Section 5, Internal 

Dosimetry. The 2006 Site Profile recommends that the radioactivity values in Section 5 should 

be used for dose reconstruction, and instructs the use of “descriptive text” that gives the 

magnitude of nuclear operations in cases where numeric data is not available. 

Given the insufficient data provided that pertains to years of nuclear reactor operations, 

incidents and releases, building use and operations, and several important nuclear facilities that 

were excluded from the 2006 Site Profile, CORE Advocacy recommends careful evaluation of 

all radioactivity values provided, which should be compared to those derived by EPA during the 

Area IV Study (2009), referenced in this document. 

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Table 1: Research Reactors at the Santa Susana Field Laboratory1 

Reactor BLDG# Facility Name Power Level 

(kW)

Period of 

Operation

Power 

Generated 

(MWd)

Radioactivity at 

End of 

Operation(10

KEWB 4073 Kinetic Experiment 

Water Boiler / Neutron 

Pulse Test Facility

1 1956-1966 1 6

WBNS / 

L-85 / 

AE-6

4093 Water Boiler Neutron 

Source

L-85

3 1956-1980 2 18

SRE 4143 Sodium Reactor 

Experiment

20,000 1957-1964 6,700 120,000

SER 4010 SNAP Experimental 

Reactor Facility

50 1959-1960 13 300

S2DR 4024 SNAP Environmental 

Test Facility

65 4/1961-12/19

62

13 390

STR 4028 Shield Test Irradiation 

Facility

50 12/1961 - 

7/1964

1 300

S8ER 4010 SNAP 8 Experimental 

Reactor Test Facility

600 5/1963 - 

4/1965

215 3,600

STIR 4028 Shield Test Irradiation 

Facility

1,000 8/1964-1974 28 3,714

SNAP-1

0

FS3

4024 SNAP Environmental 

Test Facility

37 1/1965 - 

3/1966

16 6,000

S8DR 4059 SNAP 8 

Developmental 

Reactor

619 5/1968 - 

12/1969

182 220

HMRFSR4012 NASA Nuclear 

Reactor

SNAP Critical Facility

1970-1972

1 U.S. Environmental Protection Agency (EPA), Historical Site Assessment (HSA), Santa Susana Field Laboratory 

Site, Area IV Radiological Study, December 2012. Citation 1: 

1 Oldenkamp, R.D. and Mills, J. C., Nuclear Operations at Rockwell’s Santa Susana Field Laboratory – A

Factual Perspective, Rockwell International; Report No. N001ER000017, September 6, 1991, p. 23. 

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Table 2: Criticality Test Facilities Located at the Santa Susana Field Laboratory1 

Systems for Nuclear Auxiliary Power (SNAP): AEC/DOE - NASA - North American Aviation Program 

Facility Name Building # Period of 

Operation

Notes

SNAP Critical 

Test

4373 1957-1963 First SNAP-2 Criticality Tests

Organic 

Moderated 

Reactor

4009 1958-1967 Basic tests of reactor concept

Sodium Graphite 

Reactor

4009 1958-1967 Basic tests of reactor concept

SNAP Critical 

Experiment

4012 1961-1971 Later SNAP criticality tests

Fast Critical 

Experiment

4100 1961-1972 Started as Advanced Epithermal Thorium Reactor 

(AETR)

SNAP Flight 

Systems

4019 1962 SNAP Flight System Criticality

SNAP Transient 

Test

4024 1967-1969 SNAP Transient Response Tests

1 EPA HSA 5-A, SSFL. Citation 3: 

1 Oldenkamp, R.D. and Mills, J. C., Nuclear Operations at Rockwell’s Santa Susana Field Laboratory – A 

Factual Perspective, Rockwell International; Report No. N001ER000017, September 6, 1991, p. 23. 

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Table 3: SSFL FACILITIES: RADIONUCLIDES & BUILDING USE 

The table below provides a numerical listing of SSFL buildings and facilities that should be included in the 

Site Profile. Changes in numbering schemes over the years are provided. A partial list of Visitor Log Entry 

Codes, which correspond to worker dosimeter badges and can aid in proving a monitored worker’s 

location, are provided. A list of radionuclides of concern for each location is provided, with a special 

column indicating Tritium and/or Neptunium use. The number of incidents associated with each facility is 

indicated. In addition, locations where DOE-NASA or NASA operations occurred (which likely required the 

participation of Rocketdyne workers affiliated with Area I, II or III Time Clock Locations), are indicated. If 

the location was excluded from the 2006 Site Profile, it is indicated. The description of the building/facility 

operations, processes, materials, and incidents can be found on the page number provided. 

*Visitor Log Entry Code T034 may apply to any/all buildings at the RMHF, since “T034” or “Building 4034” was 

considered to be the primary Use Authorization Building, out of which all operations, use permits, etc. were 

designated and/or assigned. 

**SNAP was a joint project between DOE-NASA to research and develop nuclear space propulsion systems. SNAP’s 

research and development relied on collaboration between North American Aviation’s atomic energy and space 

divisions, Atomics International and Rocketdyne. Some Rocketdyne Space & Information Division employees 

frequently performed job duties in Area IV, as needed, after clock-in at an Area I, II or III location. 

Bldg. 

#

Visitor 

Log 

Entry 

Codes

Building Name Radionuclides of Concern H-3 Np INC.

RPTS

DOENASA

AREA

I,II,III

Time 

Clocks

Not in

2006 

Site 

Profile

Page

#

4003

003 

T003

ETB 

Hot Cave

Uranium, thorium, transuranic 

e l e m e n t s , m i x e d fi s s i o n 

products, and activation 

products. U-238, U-234, U-235, 

U - 2 3 6 , P u - 2 3 9 , P u - 2 4 0 , 

Pu-241, Pu-242, Am-241, 

Th-228, Th-232, Th-234, H-3,

C-14, Na-22, Na-24, Cr-51, 

Mn-54, Ni-59, Ni-63, Fe-59, 

Co-60, Kr-85, Sr-89, Sr-90, 

Sb-125, I-129, I-131, Cs-134, 

Cs-137, Ce-144, Ba-La-140, 

Nb-95, Ru-103, Ru-106, 

Xn-133, Xe-135, Pm-147, 

Sm-151, Eu-152, Ra-226, Ac228

6 226

4005

005

T005

UCFPP U-238, U-234, U-235, C-14, 

Mn-54, S-35, Th-231, P-32, 

Fe-59, Co-60, Pa-234m

8 X 239

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4006

006

T006

SCTI P r e d o m i n a n t l y a n o n -

radiological facility; records of 

minor uses of radioactive 

materials. Cs-137, Mn-54, H-3, 

and natural and U-234, U-235, 

U-238. Potential migration from 

SNAP 4010, 4012, and 4024 

include: Sb-125, Am- 241, 

C s - 1 3 4 , C s - 1 3 7 , C o - 6 0 , 

Eu-152, Eu-154, Pu-238, 

Pu-239, Pu-240, Pu-241, Sr-90, 

H-3, and U-234, U-235, U- 238.

X 1+ X 381

4009

009

T009

OMR Reactor

SGR Reactor

SGR Hot Lab

[2 HOT LABS ON 

SITE]

Van de Graaf 

Particle 

Accelerator

[2 ON SITE]

Eng. Dev. Fac.

HERF

Decay products from U-235, 

U-238, and Th-232 decay 

chains include Th-234, Th-228, 

a c t i n i u m - 2 2 8 ( A c - 2 2 8 ) , 

radium-226 (Ra-226), lead- 214 

(Pb-214), bismuth-214 (Bi-214), 

P b - 2 1 2 , B i 2 1 2 , a n d 

thallium-208 (Tl-208). In 

addition, H-3, Sr-90 and 

Cs-137, Na-24, Co-57, Co-60, 

Eu-152, and Am-243 would 

have been formed.

X 218

4010

010

T010 

SNAP

SER

S8ER

(DOE-NASA 

program)

Sb-125, Am-241, Cs-134, 

C s - 1 3 7 , C o - 6 0 , E u - 1 5 2 , 

Eu-154, Pu-238, Pu-239, 

Pu-240, Pu-241, Sr-90, H-3, 

U-234, U-235, U-238. Be-10 is 

a potential radionuclide of 

concern at Building 4010 

because beryllium reflectors 

were used in the SNAP reactor.

X 3 X 124

4011

011

T011

Rad Inst. Calib. 

Lab

Am-241, Cs-137, Co-60, 

E u - 1 5 2 , E u - 1 5 4 , I r - 1 9 2 , 

Pb-210, Pu-238, Pu-239, 

Pu-240, Pu-241, K-40, Ra-226, 

Sr-90, Ta-182, Tc-99, Th-230, 

Th-232, H-3, U-234, U-235, 

U-238.

X 8 X 396

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4012 

012 

T012

SNAP Critical 

Tests

Neutron Pulse

Generator

NASA Reactor:

 HMRFSR

Radiography & 

X-Ray (ETEC)

(DOE-NASA 

program)

Am-241, Be, Cs-137, H-3, PuBe, Pu-238, Pu-239, Pu-240, 

Pu- 241, Sr-90, U-234, U-235, 

U-238

X 41 X 173

4013

013

T013

SNAP 

Component 

Assembly & 

Performance 

Bldg.

X-Ray Booth

ETEC Thermal 

Transient Test 

Facility (TTF)

(DOE-NASA 

program)

Am-241, Cs-137, Pu-238, 

Pu-239, Pu-240, Pu-241, Sr-90, 

U-234, U-235 and U-238

X X 198

4019 

019 

T019

SNAP10AFS1

Fissile Material 

Storage

ETEC Const. 

Staging & Comp. 

Facility

(DOE-NASA 

program)

Am-241, Cs-137, H-3, Pu-238, 

Pu-239, Pu-240, Pu-241, Sr-90, 

U-234, U-235, U-238

X 3 X 179

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4020 

020 

T020

6012

T020

AIHL

6001F

6000C

Hot Lab

*Material 

transported to 

4023 included 

items with 

Neptunium.

U-235, U-238, and Th-232 

decay chains include Th-234, 

Th-228, Ac-228, Ra-226, 

Pb-214, Bi-214, Pb-212, 

Bi-212, and Tl- 208. In addition, 

H-3, Sr-90, Cs-137, Na-24, 

Co-57, Co-60, Eu-152, and 

Am-243. Solid reactor fuel 

materials (including but not 

limited to) U-Mo, UO2, U-ZrH, 

plutonium-bearing fuels, (in 

addition to cladding materials: 

aluminum, sintered aluminum 

powder, stainless steel, 

zirconium, Hastelloy alloys, 

Sodium, and NaK bonded 

capsules)

X X 129 252

4021

021 

T021

6017

T022

TO34

6000E

6001R

RMHF

1959 - present

157 Radionuclides were 

potentially present at this 

location. They include: Uranium 

Fuel Materials: U-234, U-235, 

U-238. Transuranic: Pu-238, 

Pu-239, Pu-240, Pu-242, 

Am-241. Fission: Cs-137, 

Sr-90, Kr-85, Pm-147. Thorium 

Breeder Material: Th-228, 

Th-232. Neutron Activation 

Products: Co-60, Eu-152, 

Eu-154, H-3, Fe-55, Ni-59, 

Ni-63, Mn-54, K-40, Na-22.

X 24 286

4022 

022 

T022

6017

T022

TO34

6000E

6001R

RMHF Uranium Fuel Materials: U-234, 

U-235, U-238. Transuranic: 

Pu-238, Pu-239, Pu-240, 

Pu-242, Am-241. Fission: 

Cs-137, Sr-90, Kr-85, Pm-147. 

Thorium Breeder Material: 

Th-228, Th-232. Neutron 

Activation Products: Co-60, 

Eu-152, Eu-154, H-3, Fe-55, 

Ni-59, Ni-63, Mn-54, K-40, 

Na-22.

X 22 298

4023 

023 

T023

SNAP 

CTL/STL

*Material was 

transported from 

4020 to 4023.

(DOE-NASA 

program)

Am-241, Pu-238, Pu-239, 

Pu-240, and Pu-241,Cs-134, 

Cs-137, Sr-90; Th-228, Th-232, 

U - 2 3 4 , U - 2 3 5 , U - 2 3 8 ; 

i s o t o p e s o f n e p t u n i u m 

(Np-237); Co-60, Mn-54, 

Eu-152, Eu-154, H-3, Ni-63, 

Fe-55,Ta-182

X X 2 X 359

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4024

024 

T024

SNAP2DR

SNAP 10

SNAP TTF

(DOE-NASA 

program)

Co-60, Eu-152, Cs-137, Sr-90, 

H-3, Eu-154, Fe-55, Ni-59, 

Ni-63, Mn-54, K-40, Na-22, 

(neutron activation products); 

Th-232, U-234, U-235, U-238 

(nuclear fuel material); Am-241, 

Pu-238, Pu-239, Pu-240, 

Pu-241, Pu-242 (transuranic 

elements)

X 3 X 28182

4025

025

T025

SNAP Remote 

Handling Mockup Bldg.

(DOE-NASA 

program)

Co-60 (Suggesting another 

sealed source in the building); 

Radionuclides associated with 

potential migration from the 

RMHF Building 4075 include 

isotopes of uranium, thorium, 

plutonium, and mixed fission 

p r o d u c t s . R a d i o n u c l i d e s 

associated with nearby SNAP 

Building 4024 include Am-241, 

C s - 1 3 7 , C o - 6 0 , E u - 1 5 2 , 

Eu-154, Pu-238, Pu-239, 

Pu-240, Pu-241, Sr-90, H-3, 

U-234, U-235, U-238

X 1 X X 200

4026

026

T026

SCTL Proximity to SNAP 4010, 4012, 

4019

X X 388

4027

027

T027

SNAP

(DOE-NASA 

program)

Proximity to RMHF results in 

higher ambient radiation 

conditions. U-234, U-235, 

U - 2 3 8 , P u - 2 3 8 , P u - 2 3 9 , 

Pu-240, Pu-242, Am-241, 

Cs-137, Sr-90, Kr-85, Pm-147, 

T h - 2 2 8 , T h - 2 3 2 , C o - 6 0 , 

Eu-152, Eu-154, H-3, Fe-55, 

Ni-59, Ni-63, Mn-54, K-40, 

Na-22, Y-90

X 1 X X 202

4028

028 

T028

STR

STIR

LMFBR

Radiography of 

S8DR Fuel 

Following Major 

Incident

(DOE-NASA 

program)

depleted uranium (U-234, 

U-235, U-238), Pu-Be, N-16, 

Po-210, Pu-238, Pu-239, 

Pu-240, Pu-241, Am-241, 

Cs-137, Co-60, Eu-152, 

Eu-154, H-3, Sr-90

X 9 X 143

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4029 

029 

TO29

RMF

HWMF

Co-60, Ra-226, Cs-137, Po-Be, 

Pu-Be

4 272

4030

030

T030

Van de Graaf 

Particle 

Accelerator

H-3, He3 - Primary concern is 

Tritium. Tritium contamination 

documented.

X 0 X 348

4032

032

T032

SNAP

LMFBR

Clinch River

ISSS 

(DOE-NASA 

program)

Co-62, Ir-192 5 X X 206

4034

034

T034

6017

TO34

6000E

6001R

RMHF

RMDF

U-234, U-235, U-238,Pu-238, 

Pu-239, Pu-240, Pu-241, 

Pu-242, Am-241, Cs-137, 

Sr-90, Th-228, Th-232, Co-60, 

Eu-152, Eu-154, H-3, Fe-55, 

Ni-59, Ni-63, Mn-54, K-40, 

Na-22.

X 0 305

4036

036

T036

SNAP 

(DOE-NASA 

program)

The EPA research team did not 

find evidence that high activity 

radioactive materials were used 

or stored within Building 4036. 

However, its proximity to RMHF 

may have resulted in direct 

radiation and sky-shine from 

the RMHF, and affected 

ambient radiation conditions in 

the area. Radionuclides of 

concern at the RMHF include 

all radionuclides addressed in 

the background study plus any 

a d d i t i o n a l r a d i o n u c l i d e s 

identified during the EPA HSA. 

EPA expressed concern about 

radioactive materials release 

and potential drainage from the 

R M H F, w h i c h m a y h a v e 

r e s u l t e d i n r e s i d u a l 

contamination at this location.

0 X X 210

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4041 

041 

T041

SRE

Component & R/

A Waste Storage 

Building

U-238, U-234, U-235, U- 236, 

Pu-239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-228, 

Th-232, H-3, Na-22, Na-24, 

Cr-51, Mn- 54, Fe-59, Co-60, 

Kr-85, Sr-89, Sr-90, Sb-125, 

I-129, I-131, Cs-134, Cs-137, 

Ce-141, Ba (La)- 140, Nb-95, 

Ru-103, Ru-106, Xn-133, 

Xe-135, Pm-147, Sm-151

X 0 95

4042

042

T042

SNAP

LMFBR

(DOE-NASA 

program)

X X 212

4044 

044

T044

6017

T022

TO34

6000E

6001R

RMHF Bi-210, Pb-210, Pu-239, Sr-90, 

Tc-99, and Th-230. An historical 

photo (c. 1990) shows two 

small containers labeled 

“Radioactive Material” and 

“Radioactive Material - 

Contaminated Waste” under a 

desk in Building 4044. Incident 

documents mixed fission 

products. Proximity to 4022 

raises concern about 

contaminant migration of 

U-234, U-235, U-238, Pu-238, 

Pu-239, Pu- 240, Pu-241, 

Pu-242, Am-241, Cs-137, 

Sr-90, Th-228, Th-232, Co-60, 

Eu-152, Eu-154, H-3, Fe-55, 

Ni-59, Ni-63, Mn-54, K-40, and 

Na-22.

X 1 307

4048

048

T048

PDU

Coal Gasification

Potential Radium 0 X X 372

4049

049

T049

Hydraulic Tests

Piqua Test Loop

PDU Coal Gas.

C-14, S-35, P-32, Fe-59, Co-60, 

Mn-54, N -59, Fe-55

4 X X 373

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4055

055

T055

T055

6013

NMDF The primary special nuclear 

materials handled in Bldg. 4055 

were plutonium and uranium. 

Accordingly the radionuclides 

of concern are uranium, 

Pu-239, and their decay and 

daughter products, primarily 

Am-241, decay products from 

U-235, U-238, and Th-232 

decay chains include Th-234, 

Th-228, Ac-228, Ra-226, 

Pb-214, Bi-214, Pb- 212, 

Bi212, and Tl-208. In addition, 

H-3, Sr-90, Cs-137, Na-24, 

Co-57, Co-60, Eu-152, and 

Am-243 would have been 

formed.

X 20 232

4057

057

T057

SNAP

(DOE-NASA 

program)

U-238, U-234, U-235, Pu-238, 

Pu-239, Pu-240, Pu-241, Am-241, 

Cs-137, Sr-90, H-3, Fe-55, Co-58, 

Co-60, N -63, Ba-133, Eu-152, 

Eu-154, Eu- 155, Pm-147, Ta-182

X X X 214

4059 

059

T059

S2DR

LLTR

MAJOR 

INCIDENT 1969

The S8DR contained 211 fuel/

m o d e r a t o r e l e m e n t s o f 

zirconium-uranium hydride ((ZrU)Hx

agents. Some parts of Building 

4059 became activated by 

neutrons produced by the 

reactors. Tritium was produced 

a n d w a s d e t e c t e d i n 

g r o u n d w a t e r. P o t e n t i a l 

radioactive contaminants 

include U-238, U-234, U-235, 

Pu-238, Pu-239, Pu-240, 

Pu-241, Am-241, Cs-137, 

Sr-90, H-3, Fe-55, Co-58, 

Co-60, Ni-63, Ba- 133, Eu-152, 

Eu-154, Eu-155, Pm-147, and 

Ta-182.

X 19 X 151

4064 

064

T064

FSF

Special Nuclear 

Materials 

Storage

U-233, U-234, U-235, U-236, 

U - 2 3 8 , T h - 2 2 8 , T h - 2 3 2 , 

Pu-238, Pu-239, Pu-240, 

Pu-241, Pu-242, Am-241, 

Ra-226, Cs-134, Cs-137, 

Sr-90, H-3, Na-22, K-40, 

Mn-54, Fe-55, Co-60, Ni-59, 

N i - 6 3 , E u - 1 5 2 , E u - 1 5 4 , 

Pm-147, Ta-182

X X 8 245

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4066

066

T066

Instrument 

Repair / 

Calibration Bldg.

U-238, U-234, U-235, Pu-238, 

Pu-239, Pu-240, Pu-241, 

Am-241, Cs-137, Sr-90, H-3, 

Fe-55, Co-58, Co-60, Ni-63, 

Ba-133, Eu-152, Eu-154, Eu155, Pm-147, Ta-182

X 2 X X 400

4073

073 

T073

KEWB Reactor Co-60, Cs-137, Eu-152, 

Eu-154, Eu-155, Mn-54, 

Pb-210, Ra-226, Sr-90, U-235, 

U-238

3 X 46

4074

074

T074

KEWB Proximity to RMHF results in 

higher ambient radiation 

conditions.

X 57

4075

075

T075

6017

T022

TO34

6000E

6001R

RMHF uranium, plutonium, thorium, 

mixed fission products, cobalt, 

europium.

1 310

4083

083

T083

KEWB Proximity to KEWB X 58

4093

093

T093

AE6 Reactor

L-85 Reactor

Xenon-135, Co-60, Cs-137, 

Eu-152, Eu-154, Sr-90, U-238, 

and U-235

4 52

4100

100

T100

AETR Reactor

FECL

LMFBR Reactor

NASA CT-SCAN

(SSME - ISS)

U-233, U-234, U-235, U-236, 

U - 2 3 8 , T h - 2 3 2 , N p - 2 3 7 , 

Pu-238, Pu-239, Pu-240, and 

Pu-241 in the AETR and in 

January 1972 for use in the 

FCEL. Decay products would 

include Th-228, Ra-228, 

Th-230, Ra-226, Pb-210, 

Pa-231, and Ac-227.

X X 2 X 223

4103

03

T03

KEWB

Support

Office / Control 

Instrumentation

Considered “Non-Radiological.” 

Proximity to RMHF and AE6/

L85 affected ambient radiation.

Max. ambient exposure rate 

found to be 23.1 μR/hr. Limit 

was listed at 19.7 μR/hr.

X

4123

123

T123

KEWB C o - 6 0 , C s - 1 3 7 , E u - 1 5 2 , 

Eu-154, Eu-155, Mn-54, 

Pb-210, Ra-226, Sr-90, U-235 

and U-238

X 59

4133 HWMF Cs-137 0 X 393

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4143

143

T143

6026 SRE Reactor

Major Incident 

1959

U-238, U-234, U-235, U-236, 

Pu-239, Pu-240, Pu-241, 

P u - 2 4 2 , A m - 2 4 1 , 

neptunium-237 (Np-237), 

Th-232, H-3, Na-22, Na-24, 

Cr-51, Mn-54, Fe-59, Co-60, 

Kr-85, Sr-89, Sr-90, Sb-125, 

I-129, I-131, Cs-134, Cs-137, 

C e - 1 4 1 , B a ( L a ) - 1 4 0 , 

technetium-99 (Tc-99), Nb-95, 

Ru-103, Ru-106, Xn-133, 

Xe-135, Pm-147, Sm-151.

X 41 64

4153

153

T153

SRE Support U-238, U-234, U-235, U-236, 

Pu-239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, 

Na-22, Na-24, Cr-51, Mn-54, 

Fe-59, Co-60, Kr-85, Sr-89, 

Sr-90, Sb-125, I-129, I-131, 

Cs-134, Cs-137, Ce-141, Ba 

(La)-140, Nb-95, Ru-103, 

Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151

X 6 X 111

4163

163

T163

SRE 

BOX SHOP

CERF

U-238, U-234, U-235, U-236, 

Pu-239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, 

Na-22, Na-24, Cr-51, Mn-54, 

Fe-59, Co-60, Kr-85, Sr-89, 

Sr-90, Sb-125, I-129, I-131, 

Cs-134, Cs-137, Ce-141, Ba, 

(La)-140, Nb-95, Ru-103, 

Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151.

X 6 98

4226

226

T226

SCTL Proximity to SNAP 4010, 4012, 

4019

X 388

4228

228

T228

SNAP/SCTI The HSA research team did not 

find evidence that radioactive 

materials were used in Building 

4228. However, radionuclides 

resulting from former reactor 

operations in Building 4012, 

which supported the northwest 

end of Building 4228, include 

Am-241, Cs-137, Pu-238, 

Pu-239, Pu-240, Pu-241, Sr-90, 

U-234, U-235, U-238. In 

addition, H-3 due to Building 

4012 operations.

X 216

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4273

273

T273

4283

283

T283

Radioactive 

Laundry

*Area II 

Drainage

U-238, U-234, U-235, U-236, 

Pu- 239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, 

Na-22, Na-24, Cr-51, Mn-54, 

Fe-59, Co- 60, Kr-85, Sr-89, 

Sr-90, Sb-125, I-129, I-131, 

Cs-134, Cs-137, Ce-141, Ba 

(La)-140, Nb-95, Ru-103, 

Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151

X 0 X 392

4355

355

T355

SCTI Proximity to SNAP 4010, 4012, 

4019

X 383

4356

356

T356

SCTI Sealed source use of Cs-137 

and Ir-192 are noted below in 

Radiological Use Authorizations 

and Radiological Incident 

R e p o r t s . T h i s s u g g e s t s 

potential use of other sealed 

sources as well. Radionuclides 

associated with potential 

migration from SNAP Buildings 

4010 and 4012 include: 

Sb-125, Am-241, Cs-134, 

Cs-137, Co-60, Eu-152, Eu154, Pu-238, Pu-239, Pu-240, 

Pu-241, Sr-90, tritiated titanium 

H-3, and natural and enriched 

U-234, U-235, U-238

X 4 X 378

4357

357

T357

SCTI Proximity to SNAP 4010, 4012, 

4019

X 385

4358

358

T358

SCTL Proximity to SNAP 4010, 4012, 

4019

X 388

4359

359

T359

SCTI Proximity to SNAP 4010, 4012, 

4019

X 377

4360

360

T361

SCTI Proximity to SNAP 4010, 4012, 

4019

X 377

4361

361

T361

SCTI Proximity to SNAP 4010, 4012, 

4019

X 377

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4373

373

T373

SNAP Critical 

Fac.

Rocket Fuel 

Storage

(DOE-NASA 

program)

U, Pu: U-235, U-238, Th-232, 

Th-234, Th-228, Ac-228, 

Ra-226, Pb-214, Bi-214, 

Pb-212, Bi212, Tl-208, H-3, 

Sr-90, Cs-137, Na-24, Co-57, 

Co-60, Eu-152, Am-243, 

Gd-64, Sm-153

X log X X 162

4392

392

T392

SCTI Proximity to SNAP 4010, 4012, 

4019

X 377

4453

453

T453

AE-6

L-85

C o - 6 0 , C s - 1 3 7 , E u - 1 5 2 , 

Eu-154, Sr-90, U-238 and 

U-235

0 X 62

4457

457

T457

SCTI Proximity to SNAP 4010, 4012, 

4019

X 377

4563

563

T563

6017

T022

TO34

6000E

6001R

RMHF uranium, plutonium, thorium, 

and mixed fission products

0 X 324

4621

621

T621

6017

T022

TO34

6000E

6001R

RMHF Co-60, Cs-137, Sr-90, Th-232, 

Am-241, Co-60, Cs-137, Ir-192, 

PuBe, and “Th-170” [sic, 

possibly Tm-170],” Bi-210, 

Cf-252, Cs-137, Cr-51, Co-56, 

Co-57, Co-58, and Co-60, 

Ir-192, Pb-210, Kr-85, Mn-52, 

and Mn-54, Ra-226, Sr-90, and 

Tm-170.

4 X 325

4622

622

T622

6017

T022

TO34

6000E

6001R

RMHF U-234, U-235, U-238, Pu-238, 

Pu-239, Pu-240, Pu-242, 

Am-241, Cs-137, Sr-90, Kr-85, 

Pm-147, Th-228, Th-232, 

Co-60, Eu-152, Eu-154, H-3, 

Fe-55, Ni-59, Ni-63, Mn-54, 

K-40, Na-22, Y-90

X 0 X 298

4641

641

T641

SHIPPING

RECEIVING

All known isotopes at SSFL; 

entrance and exit point for all 

Area IV arrivals and shipments.

X 399

4643

643

T643

KEWB

STACK

C o - 6 0 , C s - 1 3 7 , E u - 1 5 2 , 

Eu-154, Eu-155, Mn-54, 

Pb-210, Ra-226, Sr-90, U-235, 

and U-238

0 X 61

Bldg. 

#

Visitor 

Log 

Entry 

Codes

Building Name Radionuclides of Concern H-3 Np INC.

RPTS

DOENASA

AREA

I,II,III

Time 

Clocks

Not in

2006 

Site 

Profile

Page

#

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
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4653

653

T653

SRE R/A Waste 

Vault

U-238, U-234, U-235, U- 236, 

Pu-239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, 

Na-22, Na-24, Cr-51, Mn-54, 

Fe- 59, Co-60, Kr-85, Sr-89, 

Sr-90, Sb-125, I-129, I-131, 

Cs-134, Cs-137, Ce-141, Ba 

(La)-140, Nb- 95, Ru-103, 

Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151

X 1 104

4654

654

T654

Interim 

Radioactive 

Waste Storage 

Area

uranium, mixed fission and 

activation products

2 106

4656

656

T656

SCTI Proximity to SNAP 4010, 4012, 

4019

X 377

4658

658

T658

6017

TO34

6000E

6001R

RMHF All known to RMHF, due to 

location being “entrance / exit”

0 X 329

4663

663

T663

6017

TO34

6000E

6001R

RMHF uranium, plutonium, thorium, 

mixed fission products

1 X 330

4664

664

T664

6017

TO34

6000E

6001R

RMHF

R/A Waste

INCINERATOR

MAJOR 

SOURCE OF 

AIRBORNE 

PARTICULATE 

AND 

RADIOACTIVE 

EMISSIONS AT 

SSFL.

Isotopes of uranium, plutonium, 

and thorium, and mixed fission 

products.

4 X 333

4665

665

T665

6017

TO34

6000E

6001R

RMHF uranium, plutonium, thorium, 

mixed fission products, cobalt, 

europium.

2 317

4668

668

T068

6017

TO34

6000E

6001R

RMHF Isotopes of uranium, plutonium, 

and thorium, and mixed fission 

products

0 X 338

Bldg. 

#

Visitor 

Log 

Entry 

Codes

Building Name Radionuclides of Concern H-3 Np INC.

RPTS

DOENASA

AREA

I,II,III

Time 

Clocks

Not in

2006 

Site 

Profile

Page

#

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
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4756

756

T756

SCTI Proximity to SNAP 4010, 4012, 

4019

X 385

Leach

Field

6017

TO34

6000E

6001R

RMHF U-234, U-235, U-238, Pu-238, 

Pu-239, Pu-240, Pu-242, 

Am-241, Cs-137, Sr-90, Kr-85, 

Pm-147, Th-228, Th-232, 

Co-60, Eu-152, Eu-154, H-3, 

Fe-55, Ni-59, Ni-63, Mn-54, 

K-40, Na-22, Y-90

X 318

4686

686

T686

SRE Temp. Hot 

Waste Storage 

Building

U-238, U-234, U-235, U-236, 

Pu-239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, 

Na-22, Na-24, Cr-51, Mn-54, 

Fe-59, Co-60, Kr-85, Sr-89, 

Sr-90, Sb-125, I-129, I-131, 

Cs-134, Cs-137, Ce-141, Ba 

(La)-140, Nb-95, Ru-103, 

Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151

X 1 96

4689

689

T689

SRE Inter. 

Contaminated 

Storage Area

U-238, U-234, U-235, U-236, 

Pu- 239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, 

Na-22, Na-24, Cr-51, Mn-54, 

Fe-59, Co- 60, Kr-85, Sr-89, 

Sr-90, Sb-125, I-129, I-131, 

Cs-134, Cs-137, Ce-141, Ba 

(La)-140, Nb-95, Ru- 103, 

Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151

X 0 109

4695 

695 

T695

Cold Trap Vault U-238, U-234, U- 235, U-236, 

Pu-239, Pu-240, Pu-241, Pu-242, 

Am-241, Th-232, H-3, Na-22, 

Na-24, Cr-51, Mn-54, Fe-59, 

Co-60, Kr-85, Sr-89, Sr-90, 

Sb-125, I-129, I-131, Cs-134, 

Cs-137, Ce-141, Ba (La)-140, 

Nb-95, Ru-103, Ru-106, Xn-133, 

Xe-135, Pm-147, Sm-151

X 103

4723

723

T723

SRE Steam 

Cleaning Pad

*Area II 

Drainage

U-238, U-234, U-235, U-236, 

Pu- 239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, 

Na-22, Na-24, Cr-51, Mn-54, 

Fe-59, Co- 60, Kr-85, Sr-89, 

Sr-90, Sb-125, I-129, I-131, 

Cs-134, Cs-137, Ce-141, Ba 

(La)-140, Nb-95, Ru-103, 

Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151

X 1 X 115

Bldg. 

#

Visitor 

Log 

Entry 

Codes

Building Name Radionuclides of Concern H-3 Np INC.

RPTS

DOENASA

AREA

I,II,III

Time 

Clocks

Not in

2006 

Site 

Profile

Page

#

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
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4733

733

T733

SRE Sodium 

Cleaning Pad

*Area II 

Drainage

U-238, U-234, U-235, U-236, 

Pu- 239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, 

Na-22, Na-24, Cr-51, Mn-54, 

Fe-59, Co- 60, Kr-85, Sr-89, 

Sr-90, Sb-125, I-129, I-131, 

Cs-134, Cs-137, Ce-141, Ba 

(La)-140, Nb-95, Ru-103, 

Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151

X 1 X 117

4773

773

T773

Sodium Waste 

Pond

*Area II 

Drainage

U-238, U-234, U-235, U-236, 

Pu- 239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, 

Na-22, Na-24, Cr-51, Mn-54, 

Fe-59, Co- 60, Kr-85, Sr-89, 

Sr-90, Sb-125, I-129, I-131, 

Cs-134, Cs-137, Ce-141, Ba 

(La)-140, Nb-95, Ru-103, 

Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151

X 2 X 119

4886 Sodium Burn Pit U-238, U-234, U-235, U-236, 

Pu-239, Pu-240, Pu-241, Pu-242, 

Am-241, Th- 232, H-3, Na-22, 

Na-24, Cr-51, Mn-54, Fe-55, 

Fe-59, Co-60, Kr-85, Sr-89, Sr-90, 

Sb-125, I-131, Cs-134, Cs-137, 

Ce-141, Ce-144, Ba (La)-140, 

Nb-95, Ru-103, Ru-106, Xe-133, 

Xe-135, Pm- 147, Eu-152, Eu-154, 

and Sm-151

X 365

VTS I Coal Gasification Potential Radium X X 375

VTS 

II

Coal Gasification Potential Radium X X 375

VTS 

III

SABER

(SPTF Support)

Sealed Sources X X 375

386

B923 Civil Works / 

Structure

Sealed Sources X X 375

B924 Steam System

(Saber / SPTF)

Sealed Sources X X 386

B934 Control Room 

(Coal 

Gasification)

Sealed Sources X X 375

B901 Pre-Test N/A X X 375

LETF Laser 

Engineering Test 

Facility

Sealed Sources X X 390

Bldg. 

#

Visitor 

Log 

Entry 

Codes

Building Name Radionuclides of Concern H-3 Np INC.

RPTS

DOENASA

AREA

I,II,III

Time 

Clocks

Not in

2006 

Site 

Profile

Page

#

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
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1 Downey The AE6 was developed at 

Downey Facility and moved to 

SSFL in 1955-56.

Canoga / 

VanOwen

L-47 / L-77 Reactors - 

1954-1960

The L-47 was lost; never 

located.

104

004

HQ

6001 DeSoto L-77 Reactor 1960-1976

Bldg. 

#

Visitor 

Log 

Entry 

Codes

Building Name Radionuclides of Concern H-3 Np INC.

RPTS

DOENASA

AREA

I,II,III

Time 

Clocks

Not in

2006 

Site 

Profile

Page

#

*Visitor Log Entry Code T034 may apply to any/all buildings at the RMHF, since “T034” or “Building 4034” was 

considered to be the primary Use Authorization Building, out of which all operations, use permits, etc. were 

designated and/or assigned. 

**SNAP was a joint project between DOE-NASA to research and develop nuclear space propulsion systems. SNAP’s 

research and development relied on collaboration between North American Aviation’s atomic energy and space 

divisions, Atomics International and Rocketdyne. Some Rocketdyne Space & Information Division employees 

frequently performed job duties in Area IV, as needed, after clock-in at an Area I, II or III location.

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 44 of 418 

5.0 Development and Testing of Nuclear Reactors 

Between 1954 and 1980, several nuclear reactors were built, tested, and operated in Area IV. 

These included both nuclear reactors and critical test assemblies. Nuclear reactor programs 

focused on the research, development, experimentation and operation of homogeneous water 

boiler-type reactors, sodium-cooled graphite-moderated reactors, and uranium-zirconium 

hydride reactors. 

>> Interesting Fact: Some nuclear reactor development at SSFL involved both North American 

Aviation’s Atomics International and Rocketdyne. The AEC and NASA were jointly invested in 

the research, development, and success of the Systems Nuclear Auxiliary Power (SNAP) 

program; NASA operated the HMRFSR reactor in Area IV; and undertook Space Shuttle and 

International Space Station development in buildings and facilities operated by DOE, in Area IV. 

These activities routinely involved employees of Rocketdyne whose clock-in locations at Areas I, 

II and III have led to a determination of ineligibility to EEOICPA. More information on NASADOE projects can be found in Sections 5.5-6.2, pertaining to SNAP Reactors & Criticality 

Facilities. 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 

 

Page 45 of 418 

5.1 HOMOGENOUS WATER BOILER REACTORS 

The water boiler reactors were operated in Buildings 4073 and 4093, and relied on the support 

of several other buildings located throughout SSFL. These reactors used a 93% enriched uranyl 

sulfate solution held in a critical configuration in a spherical vessels. Rather than actually boil, 

the neutron and gamma flux caused radiolytic decomposition of water into hydrogen and oxygen 

in the form of tiny bubbles, which gave the impression of boiling. Two Homogenous Water Boiler 

Reactors were developed and tested at SSFL Area IV. A description of their operations, and the 

neighboring facilities and processes required to support these operations, are provided on the 

following pages. 

Several buildings and facilities associated with homogenous water boiler reactor operations 

were excluded from the 2006 SSFL Site Profile. Information about their operations and 

processes can be found in Section 5.2.

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 46 of 418 

Section 5.1.1 

Building Number: 4073

Building Alias: 073 / T073

Building Name: Kinetics Experiment Water Boiler (KEWB) - Neutron Burst Facility 

Building Function: Homogenous Water Boiler Reactor 

Supportive Buildings in the Vicinity: 

4123 (123 / T123): Waste Storage / Fuel Storage

4643 (643 / T643): Exhaust Blower Building

4793 (793 / T793): Electrical Control Building 

Note: North American Aviation Space & Information Division (Rocketdyne) participation.1 

Radionuclides of Concern: Reactor fuel for the KEWB reactor was U-235 dissolved as uranyl 

sulfate in solution. Based on radiological investigations and historical documents, the 

radionuclides of concern include Co-60, Cs-137, Eu-152, Eu-154, Eu-155, Mn-54, Pb-210, 

Ra-226, Sr-90, U-235, and U-238.2 

5.1.1.1 Description of Operations & Processes 

Building 4073 was constructed in 1955 and served as the KEWB reactor building. Other 

buildings within the vicinity of Building 4073 include Building 4123 (Waste and Fuel Storage), 

Building 4643 (Exhaust Blower Building), and Building 4793 (Electrical Control Building). The 

EPA HSA provides a detailed description of all associated building features. 

KEWB operations began in 1956. The KEWB was a small graphite-encased research reactor 

that used a water solution of uranyl sulfate as fuel. From the autumn of 1956 to the summer of 

1959, more than 900 transient tests were conducted under a variety of initial core conditions 

and for various reactivity inputs and injection rates. Following these tests, the “A” Core 

(spherical) was replaced by the “B” Core (cylindrical) that was used in a series of experiments 

utilizing the cylindrical core geometry. The reactor had a capacity of 50 kWt, but did not normally 

operate at full power; it has been reported that the majority of reactor operations were 

conducted at a power level of 1 kWt or less.3 

1 EPA/HGL, Final HSA 5-A, p. 63, citation 290:

290 Atomics International, Document NAA-SR-7300 Special, “Evaluation of the Atomics International Nuclear 

Development Field Laboratory as a Location for Reactor Facilities,” May 25, 1962. 

2 EPA/HGL, Final HSA 5-A, p. 73, citation 332:

332 Rocketdyne Report, N001ER000017, “Nuclear Operations at Rockwell’s Santa Susana Field Laboratory 

– A Factual Perspective,” September 1991. 

3 EPA/HGL, Final HSA 5-A, p. 64, citation 295:

295 Rocketdyne Report, N001ER000017, “Nuclear Operations at Rockwell’s Santa Susana Field Laboratory 

– A Factual Perspective,” September 1991. 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 47 of 418 

In May 1962, the KEWB was operating as a neutron burst facility with 30% of its operating time 

devoted to the investigation of void formation in organic reactor transients and the remaining 

operation time being used by outside organizations, such as the Space Technology Laboratory, 

University of California, and DOE-Contractor North American Aviation’s Space & Information 

Division (S&ID) Rocketdyne.1 

The KEWB building’s 1,000-gallon tank, originally used to retain reactor cooling water, drained 

to an open drainage channel to its eventual discharge in the Area II R-2A Pond. [The R-2A Pond 

functioned as part of the Site-Wide Water Reclaim System that provided all Rocketdyne teststands in Areas I, II and III with water used to conduct rocket engine testing and hose down 

work areas].2 

As of August 1966, inspection of Building 4073 indicated that activities at the building had been 

reduced to “caretaker status” and out-of-reactor fuel inventories had been reduced accordingly. 

Operations in the building halted in November, 1966. A June 4, 1969 nuclear safety analysis 

provided information for the removal of fuel from the KEWB reactor as part of the KEWB 

deactivation. According to the nuclear safety analysis, the nuclear fuel at the KEWB comprised 

18 liters of 93% enriched uranium, uranyl sulfate solution containing 1,800 grams of U-235. The 

analysis indicated that the fuel would be loaded in five 912 containers and stored temporarily at 

the “Santa Susana Vault” prior to being shipped for reprocessing. As of June 1969, the fuel was 

located in the storage tank of the KEWB reactor. According to a nuclear safety analysis, it was 

to be transferred to polyethylene bottles. The repackaging was to occur in the decontamination 

room of Building 4021 (RMHF).3 

The KEWB reactor facility (including Building 4073) was demolished in 1975. Contaminated or 

activated equipment and materials were removed and sent to the Radioactive Materials 

Handling Facility (RMHF) for decontamination and disposal for unrestricted use, or packaged for 

shipment to the Nuclear Engineering company burial site in Beatty, Nevada, for burial. The 

above-grade structures and roof of Building 4073 were demolished. The remaining concrete 

floors and walls were decontaminated to levels that were “as low as practicable,” and levels that 

were established as “acceptable” by Rockwell International. Beta-gammer emitter contamination 

limits were defined, totaled as 0.1 mead at 1 cm with 7 mg/cm2 absorber; removable 

1 EPA/HGL, Final HSA 5-A, p. 64, citation 296:

296 Atomics International, Document NAA-SR-7300 Special, “Evaluation of the Atomics International Nuclear 

Development Field Laboratory as a Location for Reactor Facilities,” May 25, 1962. 

2 U.S. DOE Office of Environmental Audit, “Environmental Survey Preliminary Report, U.S. DOE Activities at SSFL,” 

1989. 

3 EPA/HGL, Final HSA 5-A, p. 64, citation 298:

298 Atomics International, Document PP-704-990-002, Decontamination and Disposition of Facilities

Program Plan, January 23, 1975. 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 48 of 418 

contamination limits were set at 100 dpm/100 cm2. Alpha Emitter totals were 100 dpm/100cm2, 

and removable contamination limits were 20 dpm/100cm2.1 

During decontamination and decommissioning, six of the graphite logs located directly beneath 

the reactor vessel on the floor of the enclosures were contaminated with what appeared to be 

uranium salts. The source of the uranium salts could not be determined from the information 

provided in the report. The graphite logs were wrapped in plastic sheet, tagged as radioactive 

waste, and transferred to the RMHF.2 While smear surveys of the concrete pad directly below 

the reactor and the concrete wall adjacent to the reactor revealed no removable contamination, 

concrete samples from these areas did indicate induced radioactivity. All concrete contained 

detectable induced radioactivity and was removed, so the levels documented in final totals 

became the “natural radiation level” of the concrete.3 

The fuel handling room contained the process systems for controlling, mixing, monitoring and 

storing the reactor fuel. Reactor liquids had been drained during the reactor deactivation in 

1968, and during the D&D the systems were thought to be empty and dry. However, when 

cutting through a horizontal line to the gas recombiner, a dark brown liquid spilled out and 

contaminated the protective and personal clothing of a technician. The contamination measured 

40 mrad/hr. 

The EPA-HSA contains sequential details and a complete chronology of the D&D of the KEWB. 

However, it should be noted that upon completion of D&D processes, it was confirmed by a final 

survey that Building 4073 was within “prescribed limits.” The prescribed limits, as defined in the 

1974 dismantling plan, were 100 dpm/cm2 removable beta and 20 dpm/100 cm2 removable 

alpha. A total of 3,045 cubic feet of radioactive waste was removed during D&D efforts and the 

site was released for unrestricted use by ERDA on March 3, 1976. 

5.1.1.2 Building 4073 Radiological Incident Reports 

The 2006 Site Description indicates that “there have been no incidents of consequence 

involving the KEWB.” However, there have been several incidents associated with KEWB 

1 EPA/HGL, Final HSA 5-A, p. 64, citations 30-302:

300 Rockwell International Report, AI-ERDA-13159, “KEWB Facilities Decontamination and Disposition Final 

Report,” February 25, 1976.

301 Argonne National Laboratory Report, no document number, “Surplus Facilities Management Program, 

Interim Post Remedial Action Survey Report for Kinetic Experiment Water Boiler (KEWB) Facility, Santa 

Susana Field Laboratory, Rockwell International, Canoga Park, California,” May 1983.

302 Rockwell International Report, FDP-704-990-002, Dismantling Plan for KEW Facility (Bldgs 073, 123,

and 793), October 17, 1974. 

2 EPA/HGL, Final HSA 5-A, p. 65, citation 303:

303 Rockwell International Report, AI-ERDA-13159, “KEWB Facilities Decontamination and Disposition Final 

Report,” February 25, 1976. 

3 EPA/HGL, Final HSA 5-A, p. 66, citation 304:

304 Rockwell International Report, AI-ERDA-13159, “KEWB Facilities Decontamination and Disposition Final 

Report,” February 25, 1976. 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 49 of 418 

operations in Building 4073 that could have resulted in a release to the environment and worker 

exposure. The following table provides information presented in the Incidents Database 

provided by Boeing. Summaries of available incident reports are provided below. 

Additionally, it is worth noting that EPA felt it was appropriate to include several excerpts from 

worker interviews in their investigation of the KEWB. CORE Advocacy does not include excerpts 

from worker interviews in this document, but has included all worker interviews in their entirety 

on the accompanying disk, which contains the EPA-HSA in its totality. 

5.1.1.3 Building 4073 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0522 2/10/1958 KEWB None Identified KEWB reactor operators received 

exposures above weekly guidelines.

A0504 4/1/1961 SS KEWB Building None Identified KEWB reactor core experiments caused 

exposure above guidelines to employee.

A0291 9/26/1961 KEWB Reactor Mixed Fission 

Products*

Radiation exposure in excess of 

guidelines.

• On February 10, 1958 KEWB reactor operators received weekly exposures greater than 300 

mrem while performing core maintenance activities. Upon further investigation, it was 

concluded that the elevated levels were a result of dosimeter error, and that actual exposures 

were within permissible levels. (A0522).1

It should be noted that incident reports involving worker exposure frequently indicate that 

dosimeter badges were faulty. The frequency of this phenomenon suggests that SSFL

employees at large may have been routinely provided with unreliable dosimeter badges. This, 

in turn, raises questions about the reliability of dosimetry data provided for monitored workers.

• From April 1 to June 30, 1961 a research engineer conducting KEWB reactor core 

experiments received a quarterly exposure to gamma and neutron radiation at levels greater 

than 3 rem. These core experiments were required for successful termination of the KEWB 

Program and resulted in high radiation levels in the reactor room, and consequently caused 

the employee’s exposure. The employee was aware of his high cumulative exposure in early 

May; however due to the importance of the tests and lack of other qualified operators, he 

continued to conduct “unreflected” core experiments, and it is reported that he did not obtain 

the required prior approval to exceed the 3 rem quarterly limit (A0504 and A0291).

It should be noted that this report reflects an employee’s ability to exceed permissible 

exposure levels without prior approval. If employees were routinely motivated to subject 

themselves to greater-than-permissible exposure levels without prior permission or the 

knowledge of their superiors, questions are raised about the control of employee exposure 

and the company’s protocol in ensuring worker safety, or enforcing safety standards on site. 

1 EPA/HGL, Final HSA 5-A, p. 71, citation 314:

314 Gerber, L.A., Atomics International Inter-Office Letter Re: Personnel Radiation 

Exposures, Incident A0522, March 5, 1958. 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 50 of 418 

5.1.1.4 Building 4073 Undocumented Radiological Incidents 

Based on Health Physicist Log Book entries for the KEWB, there may have been other incidents 

associated with KEWB operations. On May 18, 1960, a log book entry at 1100 hours indicated 

that “oil from EG&G experiment leaked out - this area contamination level from 18 to 132 dpm. 

Area marked off to be cleaned A.S.A.P.” An entry on the same day at 1300 hours read, 

“contamination level in Reactor Room 3 100 dpm. Decontamination A.S.A.P.” At 1400 hours on 

the same day, an entry described as a smear of the “oil spot from dumping” indicated a 

contamination level of 3x104 dpm. At 1415 hours, the log book reports that equipment from the 

reactor room was covered with oil and smears of the equipment found a contamination level of 

28x104 dpm. “EG&G” personnel were checked at 1445 hours that day and were found to have 

contaminated hands and shoes. The log entry indicates that complete decontamination of hands 

and shoes were performed. The next entry in the log books was made on May 19, 1960 at 900 

hours, and indicates the building was opened and surveyed. The maximum reading was 

documented to have been 2 R/hr. 

On May 20, 1960, the Health Physicist entered the following into the log book: “Survey and 

smear all EG&G equipment used in test building. All equipment packed and sealed for shipment 

to EG&G Boston, Mass.” At 1330 hours on the same day, the test building was cleaned and 

decontaminated. The log book did not provide the results of any additional surveys in relation to 

this incident.1 

The EPA-HSA provides a chronology of Radiological Investigations and Decontamination / 

Cleanup of Releases that occurred at the KEWB. It is noteworthy that investigations conducted 

by Rockwell International did not include specific information about the level of contamination or 

sampling procedures that were used. Rockwell International only reported that survey results 

were below the 1976 acceptable limits and that it had found no levels of beta-gamma surface 

contamination above the measured background. In May 1983, Argonne National Laboratories 

performed a post-remediation radiological survey that included a surface scan to determine 

ambient gamma exposure rate and low-level radiation levels. Soil samples were taken and 

analyzed for gamma radiation and uranium. While the survey reportedly found no 

measurements above background, according to the 1983 report conducted by Argonne National 

Laboratories, background levels were relatively high (40 μR/hr and 8,000 cts/min) due to shine 

from the nearby RMHF Buildings 4021 and 4022. 

According to a 1975 site-wide D&D program plan, the KEWB was exempt from licensing 

requirements as a U.S. Energy Research and Development Administration (ERDA) facility. The 

plan noted; however, that in the event the facilities and associated ERDA-optioned land would 

be reverted to Atomics International, the regulations of the State of California Bureau of 

1 EPA/HGL, Final HSA 5-A, p. 72, citation 318 & 319:

318 Sessions, Health Physicist, KEWB and AE-6 Log Book, pg 44-50.

319 Additional information regarding this apparent incident could not be located. The research team has 

been unable to identify EG&G or determine the precise location of the oil spill or the contents of the oil. 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
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Radiological Health would apply to the site. As a result, the facilities included in the plan were to 

be decontaminated to a level that was “acceptable to the State of California.”1 

Radiological Use Authorization No. 84 was issued January 16, 1975 for the D&D of the KEWB 

facility. It listed enriched uranium, activation products, and mixed fission products of unknown 

quantity as being distributed throughout the facility.2 

The preliminary MARSSIM Classification for the Building 4073 area and its associated 

structures is Class 1 because of previous site use, incident reports, and radioactive materials 

used during building operations. 

1 EPA/HGL, Final HSA 5-A, p. 73, citation 328:

328 Atomics International, Document PP-704-990-002, Decontamination and Disposition of Facilities

Program Plan, January 23, 1975. 

2 EPA/HGL, Final HSA 5-A, p. 74, citation 329:

329 Heine, W., Authorization No. 84, January 16, 1975. 

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Section 5.1.2 

Building Number: 4093

Building Alias: 093 / T093

Building Name: Water Boiler Neutron Source (WBNS)

a.k.a. AE-6 Reactor 

a.k.a. L-85 Reactor 

Building Function: Homogenous Water Boiler Reactor 

Associated Buildings: 4643 (643/T643): Fuel Handling & Storage

4893 (893/T893): Pad

Sanitary Leach Field 

Building 4093 Radionuclides of Concern: Reactor fuel for the L-85/AE-6 reactor consisted of 

U-235 (93.11% enrichment), dissolved as uranyl sulfate in 12.5l of .35 molar H2SO solution. As 

a result the radionuclides of concern are xenon-135, Co-60, Cs-137, Eu-152, Eu-154, Sr-90, 

U-238, and U-235. 

5.1.2.1 Description of Operations & Processes 

Building 4093 was constructed in approximately 1958 to house the AE-6 Reactor. The AE-6 

Reactor was originally called the Water Boiler Neutron Source (WBNS) Reactor. Built at the 

Downey Facility (Downey, California) in 1952, the WBNS had a maximum power of .5 Wt. The 

WBNS was modified to produce a maximum power of 3 kWt and moved to SSFL, where it 

became referred to as the AE-6 Reactor. 

The AE-6 and its associated experimental facilities were specifically designed to provide a 

thermal neutron source for evaluating neutron behavior in subcritical exponential-type 

assemblies, and for irradiating foils and other materials. A 1959 Atomics International brochure 

also indicated the reactor was “utilized for the performance of exponential experiments, lattice, 

buckling, and other reactor physics studies.” In 1972 the name of the reactor was changed to 

“L-85.” The reactor operated one-and-off for 24 years, from November 1956 to February 1980. 

THE AE-6/L-85 installation consisted of a reactor building that housed the reactor and control 

room, and a separate building used for fuel handling and storage (Building 4643). 

The reactor bay had an open ceiling approximately 10 meters high with an overhead crane.1 

The EPA-HSA provides a detailed description of the building features. The neutron flux was 

available in various magnitudes for experimental purposes. In 1974, polyethylene shielding was 2

1 EPA/HGL, Final HSA 5-A, p. 83, citation 371:

371 Oak Ridge Associated Universities, Confirmatory Radiological Survey of the L-85 Reactor Facility, 

Rocketdyne Division, Rockwell International Corporation, Santa Susana, California, December 1986. 

2 EPA/HGL, Final HSA 5-A, p. 83, citation 372:

372 Atomics International, NAA-SR-7300, Evaluation of the Atomics International Nuclear Development Field 

Laboratory as a Location for Reactor Facilities, May 25, 1962. 

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added to the top of the reactor and along the east side wall shield. Following the addition of the 

shielding, it was determined the radiation levels for reactor operation at the 2.75 kW power level 

were within “acceptable limits;” however, the document does not indicate what the 1974 

acceptable limits were. The building did not include a stack. 1

The L-85 reactor operated as a commercial operation for central station power plant operator 

training and for neutron radiography inspection of precision forgings, castings, and electronic 

and explosive devices for manufacturing defects. 

The preliminary MARSSIM Classification for the Building 4093 area is Class 1 because of the 

building’s previous site use, known radiological use, and previous incident reports. 

5.1.2.2 Building 4093 Radiological Incident Reports 

According to the 2006 SSFL Site Description, only one incident occurred at Building 4093 on 

March 25, 1959, and there is no reference to worker exposure. However, the Incident Report 

related to that particular incident is referenced in the table below, along with several others that 

could have resulted in a release to the environment and worker exposure. The information is 

presented in an Incidents Database that was provided to NIOSH by Boeing in 2009. Summaries 

of all available incident reports are also provided below and in Attachment A of the HSA. 

5.1.2.3 Building 4093 Radiological Incident Report Summary - Data Provided by Boeing 

*Mixed fission products have been identified as typically including Cs-137 and Sr-90. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

AO275 3/25/1959 AE-6 Reactor Mixed Fission 

Products*

Fission gas release contaminated 

employees and part of the building.

A0512 10/6/1960 AE-6 Fuel Building None Identified High radiation alarm malfunctioned 

during fuel handling.

A0106 7/30/1982 L-85 Reactor Room U/Mixed 

Fission 

Products*

Purging of reactor core resulted in 

contamination of employee and floor

A0661 5/24/1995 Building 4093, 

SSFL

Cs-137 Radioactive material in a non-controlled 

area

• On March 25, 1959 one of the instruments controlling the power of the AE-6 malfunctioned 

and allowed the reactor to exceed the normal power level of 3kWt and approach 4kWt. The 

operation at the higher power level lasted approximately one minute. A release of fission gas 

contaminated part of the high bay and employees. Contamination levels were measured from 

7.5 mR/hr to 13 mR/hr. The maximum release estimated was 10 mCi of principally xenon-135 

(A0257). However, the Atomics International IOL Report by G.L. Blackshaw, who was present 

at the time of the event, specified that the primary contamination was with fission products, not 

1 EPA/HGL, Final HSA 5-A, p. 83, citation 373:

373 Felten, L.D. Internal Letter Re: Review of L-85 Reactor-Building 093, May 19, 1972. 

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xenon-135.1
 


 

Blackshaw detailed employee contamination as a result of personnel removing shielding 

blocks while radiation levels near the recombiner were still 25 R and replacing the shielding 

blocks without wearing gloves. Due to the heavy concrete dust in the region of the recombiner, 

employees’ hands faces and clothing became contaminated at 1.5-2.0 mr. The employees 

were sent to the ETB (Building 4003) for showers, where subsequent examination apparently 

showed that their contamination had been removed by showering.

Blackshaw described heavy air contamination in the control room, and Building 4093’s bay 

doors were opened to release it to the outdoors. A sample of vacuum pump oil revealed a 

count rate of 111,000 cpm that decreased to 500 cpm the following day, showing evidence of 

short-half-lived fission products. Moreover, Blackshaw described specific reactor functions that 

supported his assertions such a degree of contamination could have been caused by fission 

gas. According to Blackshaw’s description of events, the principle contamination at the AE-6 

was with fission products, not xenon-135.

• The report states that, “Due to the 25 R field in the immediate vicinity of the recombined (as 

compared to the usual value of 100-200 mr after an overnight wait), only 8 shielding blocks 

had been removed instead of the usual 16. This made it difficult to reach the exhaust tank in 

its usual position in a wooden container next to the reflector.” This suggests that the readings 

close to the recombined were still in excess of the agreed-upon levels for personnel in the 

work area.

• On October 6, 1960 an alarm malfunctioned as a result of an “electrical transient.” Two 

employees were handling AE-6 fuel at the time the alarm sounded; however, the nature of the 

alarm was discovered and the area was released (A0512).2

• On July 30, 1982 rinse water contaminated with 5 ml of U-235 was spilled during the fuel 

draining operation, contaminating an employee and an area of the high bay floor. According to 

the incident report, after all the L-85 fuel solution was drained from the core, and the rinse 

water was drained without incident, an inline absolute filter was connected to the drain line 

and nitrogen was flowered through the system to dry it out. More moisture was in the system 

than anticipated, and it got through the absolute filter and dripped off of the plastic covered 

shield block onto the floor. The employee present observed the spill, stopped the flow of 

nitrogen, and used “kimwipes” to move the absolute filter further back on the shield block. The 

incident report indicated that a check of the alpha air monitor located in the reactor room 

indicated no airborne release, while a smear survey of the reactor room floor spill area 

showed contamination levels of 930,000 beta, and dropped down to 1,000 dpm further from 

the spill/work area. According to the 2005 HSA, the area was partially decontaminated at the 

1 EPA/HGL, Final HSA 5-A, p. 88, citation 394 & 395:

394 Unknown, Building 4093 – L-85 (AE-6) Research Reactor, Unknown Date.

395 Blackshaw, G.L., NAA-SR-Memo 3757, “Release of Fission Gases from the AE-6 Reactor on March 25, 

1959,” April 15, 1959. 

2 EPA/HGL, Final HSA 5-A, p. 88, citation 396:

396 Sessions, S.D., Atomics International Inter-Office Letter Re: Radiological Safety Incident Report A0512, 

November 17, 1960. 

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time and fully decontaminated during facility decommissioning. The incident report does not 

document how the area was partially decontaminated (A0106).1

• On May 24, 1995 an encased radioactive high efficiency particulate air (HEPA) filter was found 

in a pile of debris during the demolition of buildings 4453 and 4093. While it was not obvious 

from which building the filter originated, the incident report indicated the filter most likely came 

from the roof of Building 4093. The incident report did not provide an estimated age of the 

filter. A survey of the inside of the filter using a Ludlum 12 GM frisker indicated an average of 

22 to 3000 cpm and a maximum of 500 cpm fixed beta. While below the facility release limits 

of 5,000 dpm/100 cm2 average and 15,000 dpm/100 cm2 maximum, it did exceed the free 

release limits of “no detectable activity” for fixed contamination. No detectable fixed beta 

radiation was observed on the outside of the encased unit. The incident report indicated that 

smears of the filter showed no removable alpha or beta contamination. The filter was taken to 

the RMHF for further evaluation. Scanning with a portable NOMAD gamma spectrometer 

indicated the presence of Cs-137. The report did not provide the activity levels of Cs-137. At 

the RMHF, the filter was packaged for disposal as low-level radioactive waste (A0661). 2

• Gamma radiation was found to be 1,000 mr/hr on the east side of the building directly 

opposite the core. In order to determine the need for access restrictions on the area 

surrounding Building 4093, a radiation survey of the perimeter of the AE-6 reactor was 

conducted August 13, 1970. The reactor was at a power level of 2 kW, and was operating with 

the shield doors on the east side open and a plastic Lucite rod, approximately six feet long, 

inserted in the one-inch diameter beam hole. The EPA research team has been unable to 

determine if the building operated with the shield doors open frequently.3 

The EPA HSA provides a detailed chronology of radiological investigations associated with 

Building 4093 and its surrounding area, as well as a sequential description of decontamination 

and decommissioning activities.

1 EPA/HGL, Final HSA 5-A, p. 88, citation 397 & 398:

397 . Wallace, J.H., Rockwell International Internal Letter Re: Radiological Safety Incident Report A0106, 

August 4, 1982.

398 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005. 

2 EPA/HGL, Final HSA 5-A, p. 88, citation 399: 

399 Rutherford, P.D., Rockwell Internal Letter Re: Radiological Incident Report A0661, June 7, 1995. 

3 EPA/HGL, Final HSA 5-A, p. 95, citation 426:

426 Johnson, B.I., Internal Letter Re: Radiation Survey of the AE-6 Reactor, Building 093, August 13, 1970. 

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Section 5.2 HWBR: Additional Locations / Operations 

Several buildings and facilities associated with Homogenous Water Boiler Reactor operations 

were excluded from the 2006 Site Profile. This section provides a partial list and a description of 

important facilities that should be included in the Site Profile. 

Some of the locations have documented processes that involved radiological materials. In some 

cases, the locations may have been designated as ‘non-radiological’ facilities despite their 

location within Area IV, involvement in radiological processes, and/or proximity to known 

sources of radiation exposure.

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Section 5.2.1 

Building Number:

Building Alias: 

4074 

074/T074 

Building Name: KEWB Reactor Support Building 

Building Function: Storage

Film Processing 

Note: Designated “non-radiological.” Proximity to RMHF affected ambient radiation 

levels. 

Radionuclides of Concern: 

5.2.1.1 Description of Operations & Processes 

Building 4074 was constructed in 1958 for use as a storage and film processing building where 

personnel processed photographic oscillograph paper for the KEWB. It featured a sampling 

station for measuring the ambient air radioactivity concentration on a quarterly basis. To date, 

no incident reports associated with Building 4074 operation have been located, and there are no 

indications that radioactive materials were handled or used at this location. 

The maximum ambient exposure rate presented in the survey report for Building 4074 included 

other buildings located in its vicinity - 4074, 4083, and 4453. The maximum ambient exposure 

rate was found to be 23.1 μR/hr. The limit was listed as being 19.7 μR/hr. It is important to note 

the report did not provide information to indicate how the ambient exposure rate was measured 

or how the limit was derived; however, the report did attribute ambient exposure rates over the 

limit to the nearby RMHF, and were reported not to represent residual contamination. 

The preliminary MARSSIM Classification for the Building 4074 area is Class 1 because of the 

building’s proximity to Building 4093 (Water Boiler Reactor) and the limited operational 

information for the building. 

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Section 5.2.2 

Building Number:

Building Alias: 

4083 / 4103

083/T083 / 103 / T103 

Building Name: KEWB Reactor Support Building 

Building Function: Office 

Control Building

Instrumentation / Recording of Transient Tests 

Note: Designated “non-radiological.” Proximity to AE-6, L-85 and RMHF affected ambient 

radiation levels. 

5.2.2.1 Description of Operations & Processes 

Building 4083 and Building 4103 were constructed in 1958 to serve as the office and control 

building for the KEWB reactor in Building 4073. These locations were included in the Nuclear 

Regulatory Commission (NRC) license, dated January 5, 1972 (R-118 DOCKET NO. 50-375) for 

the L-85 reactor (Building 4093); however, EPA’s research team was not able to locate any 

documentation indicating that Building 4083/4103 was included in the license for the L-85 

reactor located in Building 4093. 

The maximum ambient exposure rate presented in the survey report for Building 4074 included 

other buildings located in its vicinity - 4074, 4083, and 4453. The maximum ambient exposure 

rate was found to be 23.1 μR/hr. The limit was listed as being 19.7 μR/hr. It is important to note 

the report did not provide information to indicate how the ambient exposure rate was measured 

or how the limit was derived; however, the report did attribute ambient exposure rates over the 

limit to the nearby RMHF, and were reported not to represent residual contamination. 

The preliminary MARSSIM Classification for the Building 4074 area is Class 1 because of the 

building’s proximity to Building 4093 (L-85 Water Boiler Reactor) and the limited operational 

information for the building. 

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Section 5.2.3 

Building Number: 4123

Building Alias: 123/T123 

Building Name: KEWB Reactor Support Building 

Building Function: Radiological Waste Storage (KEWB, Bldg. 4073) 

Radionuclides of Concern: Reactor fuel for the KEWB reactor was U-235 dissolved as uranyl 

sulfate in solution. Based on radiological investigations and historical documents, the 

radionuclides of concern include Co-60, Cs-137, Eu-152, Eu-154, Eu-155, Mn-54, Pb-210, Ra226, Sr-90, U-235 and U-238.1 

5.2.3.1 Description of Operations & Processes 

Building 4123 was constructed sometime between 1957 and 1962. Surrounding buildings 

include Building 4073, 4643, and 4893. It was used for the temporary storage of radiological 

wastes material associated with the operations of building 4073, the KEWB Reactor. This 

included the storage of uranyl sulfate and mixed fission products.2 

EPA determined it was appropriate to include an excerpt from the employee interviews, which 

included a former employee’s assertion that upon the KEWB core’s replacement of the spherical 

core with the cylindrical core, the spherical core was stored at Building 4123. According to the 

employee, the workers tasked with the transfer of the spherical core to storage in room Building 

4123 resulted in “getting our yearly dose.” No incident reports were identified by the EPA

research team. 

EPA HSA provides a detailed chronology of radiological investigations and decontamination of 

Building 4123. In May, 1983 Argonne National Laboratories performed a post remediation 

radiological survey to verify that the site was free of radioactivity except for normal background. 

Argonne National Laboratories determined that background was relatively high (40 μR/hr and 

8,000 cts/min) due to the “shine” from nearby Buildings 4021 and 4022 (RMHF). The survey 

concluded the site could be released for unrestricted use. 

In August 1988, Rocketdyne performed a surface scan between the KEWB facilities and the 

RMHF measuring ambient gamma exposure rate to ensure no contamination existed as a result 

of radioactive materials movement. The survey results were below the 1988 acceptable limits. 

However, the findings are inconsistent with the findings of the Argonne National Laboratory’s 

report that determined background levels to be 40 μR/hr (the limit was 17.0 μR/hr). The August 

1 EPA/HGL, Final HSA 5-A, p. 99, citations 444:

444 

Rocketdyne Report, N001ER000017, “Nuclear Operations at Rockwell’s Santa Susana Field Laboratory 

– A Factual Perspective,” September 1991. 

2 EPA/HGL, Final HSA 5-A, p. 96, citations 429 & 430:

429 Author unknown, Vanowen and SSFL Operations 1959, Laboratory Status Reports, undated.

430 Rocketdyne Report, N001ER000017, “Nuclear Operations at Rockwell’s Santa Susana Field Laboratory 

– A Factual Perspective,” September 1991. 

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1988 report did not provide any comparison of findings with previous reports to provide any 

reasons for the discrepancy.1 

The preliminary MARSSIM Classification for the Building 4123 area is Class 1 because of the 

building’s purpose as a temporary radioactive waste storage facility, the location of the building 

within ETEC, and its proximity to Building 4073. 

1 EPA/HGL, Final HSA 5-A, p. 98, citations 440:

440 Argonne National Laboratory, “Surplus Facilities Management Program, Interim Post Remedial Action 

Survey Report for Kinetic Experiment Water Boiler (KEWB) Facility, Santa Susana Field Laboratory, 

Rockwell International, Canoga Park, California,” May 1983. 

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Section 5.2.4 

Building Number: 4643

Building Alias: 4643/T643 

Building Name: KEWB Reactor Support Building 

Building Function: Exhaust Stack / Mechanical Building 

Radionuclides of Concern: Reactor fuel for the KEWB reactor was U-235 dissolved as uranyl 

sulfate in solution. Based on radiological investigations and historical documents, the 

radionuclides of concern include Co-60, Cs-137, Eu-152, Eu-154, Eu-155, Mn-54, Pb-210, 

Ra-226, Sr-90, U-235, and U-238.1 

5.2.4.1 Description of Operations & Processes 

Building 4643 was a small mechanical building with a 60-foot exhaust stack that had a 2,000-

cubic foot per minute blower system. The building was connected to a 300-gallon underground 

storage tank that was connected to Building 4073 and held radioactive gases released from the 

core for a half-life period prior to expelling the exhaust to the atmosphere.2 

Constructed in the early 1960’s, Building 4643 was an exhaust building that provided ventilation 

for the KEWB reactor building. During D&D in 1975, a contamination survey of Building 4643 

and the associated equipment indicated the floor of the exhaust building was contaminated with 

removable beta contamination levels of up to 600 dpm/100cm2. The exhaust blower and filter 

plenum were also contaminated and were removed and sent to the RMHF for disposal. 

In August 1988, Rocketdyne performed a surface scan of the terrain between the KEWB 

facilities and the RMHF measuring ambient gamma exposure rates to ensure no contamination 

existed as a result of radioactive materials movement. Again, Rocketdyne’s findings were 

inconsistent with those of Argonne National Laboratories report that determined background 

levels to be 40 μR/hr. The August 1988 report did not provide any comparison of findings with 

previous reports, or any explanation for the discrepancy. The EPA-HSA provides a detailed 

chronology of facility radiological investigations and D&D. 

The EPA research team did not locate any documented radiological incidents associated with 

Building 4643.

1 EPA/HGL, Final HSA 5-A, p. 73, citation 332:

332 Rocketdyne Report, N001ER000017, “Nuclear Operations at Rockwell’s Santa Susana Field Laboratory 

– A Factual Perspective,” September 1991. 

2 EPA/HGL, Final HSA 5-A, p. 107, citations 463 & 464:

463 Rockwell International, AI-ERDA-13159, “KEWB Facilities Decontamination and Disposition Final

Report,” February 25, 1976.

464 Atomics International, Document NAA-SR-7300 Special, “Evaluation of the Atomics International Nuclear 

Development Field Laboratory as a Location for Reactor Facilities,” May 25, 1962. 

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Section 5.2.5 

Building Number: 4453

Building Alias: 453/T453 

Building Name: AE-6 / L-85 Reactor Support Building 

Building Function: Fuel Handling Facility

Measuring Background for Facilities Related to AE-6/L-85

Neutron Radiography Storage Building (1984) 

Notes: Despite technical designations as “non-radiological” facility, this location has 

documented use and storage of radiological material. 

Radionuclides of Concern: Co-60, Cs-137, Eu-152, Eu- 154, Sr-90, U-238 and U-235. 

5.2.5.1 Description of Operations & Processes 

Located approximately 165 feet southeast of Building 4093 (AE-6 / L-85 Reactor Building), 

Building 4453 was constructed in approximately 1958. It was used for fuel handling and for 

measuring background for facilities associated with AE-6 / L-85 reactor operations. The EPA

research team notes that the Use Authorization for Building 4093 allowed for the storage of 200 

grams of U-235 within a metal safe at Building 4093, but “Fissile materials, other than the fuel 

solution in the reactor and what is stored in the metal safe, must be stored in approved safegeometry containers in an array no more reactive than authorized for the containers.” While 

documents do not provide the location of these storage containers, EPA’s research team noted 

1 it is possible that the materials were stored in Building 4453. In 1984, Building 4453 is listed as 

a “Neutron Radiography Storage Building.” 2 The above-average ambient exposure rates for 

Building 4453, 4083 and 4074 were attributed to the nearby RMHF and were reported not to 

represent residual contamination. 

Building 4453 was located outside the restricted area surrounding the AE-6 / L-85 reactor and 

was reported to have had no history of non-sealed source radioactive materials use, and no 

documented radiological incidents. However, documents indicate the building served as the fuel 

handling facility for Building 4093.Because the building is stated to have served as a fuel 

handling facility for Building 4093, there is a possibility of elevated radionuclide concentrations 

in the soil. The preliminary MARSSIM Classification for Building 4453 is Class 1. 

The EPA research team did not locate any radiological incidents associated with Building 4453.

1 EPA/HGL, Final HSA 5-A, p. 101, citation 450:

450 Skoholt, Donald, North American Rockwell Corporation, Docket No. 50-375, Facility License No 

R-118, January 5, 1972. 

2 EPA/HGL, Final HSA 5-A, p. 101, citations 451:

451 Olson, P, et. al, GEN-ZR-0002, CERCLA Program Phase II – Site Characterization, May 29, 1987. 

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5.3 SODIUM-COOLED GRAPHITE-MODERATED REACTORS 

The 2006 Site Description provides an insufficient, incomplete and inaccurate description of 

operations associated with the Sodium Reactor Experiment (SRE). The 2006 Site Description 

indicates the reactor was shut down in 1964, missing nearly three years of reactor operations 

under the SRE-PEP (Power Expansion Program). 

In addition, the 2006 Site Description only references one incident at the SRE (July 12, 1959 - 

“Power Run 14”), and provides information about the event that conflicts with numerous 

historical reports authored by the AEC and Atomics International, as well as with the Incident 

Database. 

The Incident Database contains 45 incident reports associated with Building 4143 alone, which 

could have resulted in releases to the environment and worker exposure. 

The 2006 Site Profile provides an incomplete list of buildings/facilities associated with SRE 

operations, for which no descriptions of facility function or incidents are provided. CORE 

Advocacy provides a description of operations and incidents for these facilities in this section. 

Several buildings/facilities involved in SRE operations were excluded from the 2006 Site Profile. 

A partial list of the excluded buildings/facilities, with descriptions of operations and incident 

report summaries, is provided in Section 5.4. 

>> Interesting Fact: DOE maintained a facility in Area I that was used to support SRE 

operations in Area IV. The location was known as the Area I Atomics International MOC Fuel 

Rod Tower. It was constructed in the 1950’s and supported Area IV operations, until 1974. In 

2009, the California Department of Toxic Substances Control (DTSC) identified Area I as a 

“point of radiological interest” as a result of DOE operations and facilities in Area I.1 

1 California Department of Toxic Substances Control (DTSC) RFI Data Gap Work Plan, Boeing RFI Subarea 1A

North. File: 66074_1Anorth_PublicMeeting_051613_rev8_draft.pdf Page 21 & Page 24 

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Section 5.3.1 

Building Number:

Building Alias: 

4143

143 / T143

Building Name: Sodium Reactor Experiment (SRE) Reactor Building 

Building Function: 1957 - 1964: Sodium Graphite Reactor (SRE)

1965 - 1967: SRE PEP Program 

Note: Major Nuclear Incident (July 12, 1959).

Tritium 

Radionuclides of Concern: U-238, U-234, U-235, U-236, Pu-239, Pu-240, Pu-241, Pu-242, 

Am-241, neptunium-237 (Np-237), Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-59, Co-60, 

Kr-85, Sr-89, Sr-90, Sb-125, I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, technetium-99 

(Tc-99), Nb-95, Ru-103, Ru-106, Xn-133, Xe-135, Pm-147, Sm-151. 

5.3.1.1 Building 4143 Description of Operations & Processes 

The Sodium Reactor Experiment (SRE) was a nuclear reactor developed to demonstrate a 

sodium-cooled graphite moderated reactor for civilian use. It reached full power (20 MW) in 

May, 1958 and provided 37 gigawatt hours of power to the Southern California Edison Company 

(SCE), powering the City of Moorpark as part of its demonstration. The SRE experienced a 

major incident July 12, 1959 that is documented below. In addition, several other incidents 

occurred that involved the SRE, which are also documented below. The HSA did not identify any 

Radiological Use Authorizations issued for Building 4143 prior to 1975. 

The 2006 Site Profile indicates that the SRE was “shut down for the last time February 14, 1964 

and maintained in a safe shutdown condition until September, 1967.” However, the SRE 

operated from 1965-1967 as part of the SRE Power Expansion Program (PEP).1

Building 4143 was constructed between April 1955 and February 1957. According to the HSA, 

the SRE was a high-temperature reactor with a slightly enriched uranium metal fuel (Core I) and 

sodium-cooled, hexagonal zirconium-clad graphite moderator elements. Following an accident 

in 1959, the enriched uranium fuel was replaced with a 93% uranium-thorium metal alloy fuel 

(Core II).

Building 4143 was approximately 20,000 feet in area and included below-grade vaults and 

galleries for radioactive materials, a high bay, ground floor and mezzanine offices, rooms 

housing support equipment, a paved area surrounding the building, several out-buildings, and 

drainage channels. The HSA provides a detailed description of Building 4143’s features,

including an inventory of equipment and a description of equipment operations. The HSA

describes the SRE reactor core as a matrix of moderator elements containing the fuel elements, 

control rods, neutron source, and instruments for measuring temperature and sodium level. 

1 EPA/HGL, Final HSA 6, p. 69, citation 5:

5 Ureda, B. and Heine, W., Facilities Dismantling Plan for SRE, Atomics International Report No. 

FDP-704-990- 003, June 24, 1975, pp. 9, 13, 17-18. 

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According to the HSA, because the SRE was an “experimental reactor,” the core geometry 

changed as test elements changed.1 

SRE Reactor Operations 1965 - 1967: SRE-PEP Program 

From May 1965 to September 1967, the SRE and Building 4143 were used for the SRE-PEP

program. Atomics International operated the main primary sodium system for 4,386 hours at 

700 ̊F and for 13,196 hours at 350 ̊F. The objective of the SRE PEP was to raise the sodium 

operating temperatures to 1,200 ̊F and thermal power levels to 30 MW with a stainless-steelclad uranium carbide fuel loading. 

The primary sodium system was shut down in September 1967. Atomics International’s 

deactivation plan involved a “stored-in-place” configuration but nonessential equipment was 

removed, and the steam generator and uncontaminated support facilities were not maintained. 

Deactivation activities were completed in 1968. Decommissioning began in 1974 with an 

objective of removing significant radioactivity from the site and releasing the facility for 

unrestricted use. The SRE superstructure was retained to provide containment for airborne 

contamination released by decontamination operations. Once Building 4143 was released for 

unrestricted use, it was used for ETEC component storage. The reactor core was not removed 

until 1977. 2 The HSA provides substantive excerpts from SRE employee interviews, and a 

detailed chronology of DD&D operations related to the SRE and Building 4143, which was 

demolished in 1999.3 

The EPA research team was unable to locate information about air monitoring results at the 

stack, although some information is provided in the Incident Reports Database. CORE 

Advocacy located the Atomics International Quarterly Report of Activity Released to the 

Atmosphere from Building 4143, dated November 20, 1959. Documentation of the activity 

released to the atmosphere at the SRE begins July 1, 1959 and ceases July 25, 1959. There is 

no data provided between July 25, 1959 and September 16, 1959 due to “slow release rates,” 

according to the report.4 It should be noted that, between the dates referenced above, fuel 

removal operations began and high radiation levels were recorded in numerous technical 

reports authored by Atomics International, and the original Incident Reports contained in the 

Incident Report Database. It should also be noted that, according to Incident Reports in the 

Incident Report Database, the SRE stack only exhausted hot cell and vent system effluent. 

1 EPA/HGL, Final HSA 6, p. 30. 

2 EPA/HGL, Final HSA 6, p. 31, citations 3 &1:

3 Ureda, B. and Heine, W., Facilities Dismantling Plan for SRE, Atomics International Report No. 

FDP-704-990-003, June 26, 1975, pp. 9, 13, 17-18.

1 

Ureda, B. and Heine, W., Facilities Dismantling Plan for SRE, Atomics International Report No. 

FDP-704-990- 003, June 26, 1975, pp. 15-18. 

3 EPA/HGL, Final HSA 6, p. 32, Information from Interviewees. 

4 Atomics International Internal Memo, “Quarterly Report of Activity Released to Atmosphere,” November 20, 1959. 

File: HDMSe00581353.pdf 

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5.3.1.2 2006 Site Profile Overview of the SRE: Inaccurate / Incomplete 

The 2006 Site Description provides incomplete and inaccurate information about Building 4143 

SRE operations and incidents over the course of operation, as described above. In addition, the 

2006 Site Description provides incomplete and inaccurate information about a major incident 

that occurred July 12, 1959 during what is known as “Power Run 14.”

Atomics International technical reports describe reactor operations that occurred before, during 

and after Power Run 14.1 2 3 However, when these technical reports are compared to incident 

reports contained in the Incident Report Database and Atomics International internal documents 

generated during the event, there are noted inconsistencies. In addition, several technical 

reports authored by Atomics International reflect worker dose estimations that, when all details 

are considered and radiation measurements in the work area are evaluated, appear to be 

suspect if not downright implausible. 

During Power Run 14, the core scope was used and several courses of action describe lifting 

and/or rotating of the reactor face. These efforts were undertaken to assess the damage the 

SRE had suffered, and later in an effort to remove damaged fuel rods from within the reactor’s 

core. Many summary reports authored in later years make the assertion that only krypton and 

xenon noble gases were released during these operations, and workers were never at serious 

risk of exposure to radiation. However, workers were tasked with scrubbing walls, floors and 

equipment with absorbent materials and decontamination solution; a task that would not have 

been necessary if contamination had been limited to noble gases, and more closely associated 

with decontamination procedures involving mixed fission products. 

Several incident reports that are contained in the Incident Report Database, generated after 

Power Run 14, describe similar or identical operations involving use of the core-scope and/or 

the lifting and rotation of the reactor face, during what would be considered normal reactor 

conditions. In numerous instances, mixed fission products from the reactor core contaminated 

the work area, and the workers. Nasal smears document serious exposure, and in several 

incidents workers are tasked with similar decontamination procedures involving the use of 

absorbent materials and decontamination solutions to scrub the walls, floors, and equipment in 

the work area. 

Since most employees involved in Power Run 14 lack dosimeter data, it would be reasonable to 

compare the worker dose estimations between those involved in Power Run 14 to those 

involved in later incidents involving similar operations with the core-scope and reactor face 

rotation. Moreover, in the event employees involved in Power Run 14 must rely on coworker 

data for dose reconstruction, it may be appropriate to apply that of fellow SRE employees 

1 Atomics International, AEC Research and Development Report: SRE Fuel Element Damage - An Interim Report of 

the Atomics International Ad-Hoc Committee 1959. File: NAA-SR-4488.pdf 

2 Atomics International, Distribution of Fission Product Contamination in the SRE, by R.S. Hart, 1960.

File: NAA-SR-6890.pdf 

3 Atomics International, AEC Research and Development Report: SRE Fuel Element Damage - Final Report of the

Atomics International Ad-Hoc Committee 1961.

File: NAA-SR-4488_Suppl.pdf 

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involved in similar or identical job processes, whose exposures to mixed fission products are 

better documented.1 2 

While the 2006 Site Description only identified the one incident associated with Power Run 14 at 

Building 4143, the Incident Report Database provided by Boeing documents 45 incidents 

associated with this location that could have resulted in environmental release and worker 

exposure. 

The preliminary MARSSIM classification for the Building 4143 area is Class 1 due to its 

documented releases and former use as the SRE reactor building. 

5.3.1.3 Building 4143 Radiological Incident Reports 

There have been several incidents associated with Building 4143 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

5.3.1.4 Building 4143 Radiological Incident Report Summary - Data Provided by Boeing

* “MFP” is likely to reference “Mixed Fission Products.” 

Incident 

File Name

Date of 

Incident

Incident 

Occurred 

After 1964 

“Shutdown”

Location of 

Incident

Isotopes

MFP = Mixed 

Fission 

Products

Description of Incident

A0374 6/13/1958 SRE High Bay Employee did not wear prescribed 

respiratory protection.

A0478 2/6/1959 SRE High Bay MFP Two employees handled 

contamination equipment without 

proper use of protective gloves.

A0315 6/4/1959 - 

6/5/1959

Wash Cell MFP Explosion blew fuel element 

undergoing sodium cleaning out of 

the Wash Cell.

A0274

MAJOR 

INCIDENT

7/12/1959 - 

5/21/1960

Reactor Core MFP SRE Fuel Element Damage

A0351 9/10/1959 R/A Storage MFP Employees entered controlled area 

without appropriate authority.

1 EPA/HGL, Final HSA 6, p. 58, citation 4:

4 Galperin, A. and Shannon, J. W., Atomics International Internal Letter, re: Radiological Safety Incident

Report, SRE High Bay, 5/25/60, August 30, 1960. 

2 EPA/HGL, Final HSA 6, p. 59, citation 1:

1 

Lane, W. D., Atomics International Internal Letter, re: Radiological Incident Report, SRE High Bay, 6/9/60, 

July 20, 1960. 

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A0390 11/21/1959 SRE High Bay MFP Employee contaminated during 

modification of SRE Core Recovery 

Tool.

A0548 12/7/1959 SRE High Bay MFP Employee entered controlled area 

without prescribed protective 

clothing.

A0391 12/18/1959 SRE Hot Cell MFP Employee entered controlled area 

without required hood and exited 

without a survey.

A0510 4/11/1960 SRE High Bay MFP Leakage around viewing glass 

caused contamination of employee.

A0394 5/15/1960 SRE High Bay MFP SRE Sodium Fire in High Bay

A0393 5/26/1960 SRE High Bay MFP Employee contaminated by core 

cover gas escaping during core 

scope removal

A0460

Incident 

Report 

Missing

6/2/1960 SRE High Bay MFP Man slightly injured and 

contaminated while working on 

SRE Fuel Coffin.

A0005 6/9/1960 SRE High Bay MFP Core cover gas leak contaminated 

employees.

A0442 7/7/1960 SRE High Bay MFP Reactor cover gas escaped 

contaminating employee.

A0305 12/15/1960 Tent Cs/Sr Improperly sealed respirator 

caused high internal and external 

contamination.

A0435 5/17/1961 SRE High Bay MFP Protective boot leaked, 

contaminating shoes.

A0515 11/4/1961 SRE High Bay MFP Employee exposed to airborne 

activity while removing samples in 

High Bay

A0472 12/23/1961 SRE High Bay MFP Controlled area entered without 

proper authority, protection, and 

procedure.

A0534 2/6/1962 SRE HP Office MFP Unauthorized employee used HP 

equipment with faulty results.

A0015 4/4/1962 SRE High Bay MFP Airborne activity in SRE High Bay

A0373

Incident 

Report 

Missing

6/13/1962 SRE High Bay MFP Operator engulfed in fire during 

transfer of core heaters.

Incident 

File Name

Date of 

Incident

Incident 

Occurred 

After 1964 

“Shutdown”

Location of 

Incident

Isotopes

MFP = Mixed 

Fission 

Products

Description of Incident

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A0404 7/30/1962 SRE West Pad Sb-124 SRE source transfer operation 

caused employee exposure above 

guidelines.

A0379 10/20/1962 SRE West Pad 3 Employees size-reducing SRE 

Core Heaters became 

contaminated.

A0030 3/19/1964 X SRE & Surface 

Drainage

MFP R/A Liquid pumped onto ground as 

“clean.”

A0409 6/13/1964 X SRE FHM PIT MFP Airborne activity in fuel handling 

machine pit contaminated 

employees.

A0493 6/17/1964 X SRE FHM PIT MFP High airborne activity in fuel 

handling machine service pit 

contaminated two employees.

A0380 6/21/1964 X SRE High Bay MFP Employees and area contaminated 

from improper doffing.

A0382 918/1964 X Poor Man’s Cell Work in the Poor Man’s Hot Cell 

caused airborne exposure to 

employees.

A0545 12/1/1964 X SRE High Bay MFP Gamma survey of control-rod 

thimble contaminated probe and 

surveyor.

A0371 12/18/1964 X SRE High Bay Size reduction of reactor 

components caused exposure and 

contamination of employees.

A0296 1/14/1965 X SRE High Bay Movement of temperature probe 

caused exposure in excess of 

guidelines.

A0386 2/20/1965 X SRE High Bay Core light replacement 

contaminated employees when 

radioactive gas was released.

A0416 2/27/1965 X SRE High Bay MFP Servicing moderator cask resulted 

in employee contamination.

A0444 4/24/1965 X SRE High Bay MFP Employee drilled through finger 

during operation in SRE High Bay,

A0445 5/22/1965 X NA Pipe Gallery MFP Employee lacerated thumb during 

operation in SRE High Bay.

A0041 8/19/1966 X Tab Exposure 

Facility

MFP Personnel exposed to primary 

sodium in TAB Exposure Facility.

Incident 

File Name

Date of 

Incident

Incident 

Occurred 

After 1964 

“Shutdown”

Location of 

Incident

Isotopes

MFP = Mixed 

Fission 

Products

Description of Incident

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A0321 12/8/1967 X Maintenance 

Cell

MFP Radioactive water discovered in 8” 

pipes that penetrated maintenance 

cell floor.

A0289 10/23/1976 D&D

Period

SRE High Bay The bottom of a Cold Trap fell off 

during movement, contaminating 

floor.

A0059 8/1/1977 SRE High Bay MFP ACP Reactor vessel segments storage 

pit leaked water to the soil.

A0414 8/10/1977 SRE High Bay MFP Employee exposed to airborne 

activity during D&D operations.

A0578 6/24/1978 SRE Employee injured when he tripped 

on culvert.

Incident 

File Name

Date of 

Incident

Incident 

Occurred 

After 1964 

“Shutdown”

Location of 

Incident

Isotopes

MFP = Mixed 

Fission 

Products

Description of Incident

*Isotopes are written as they are presented in the Incident Database. The EPA research team believes that ‘MFP’ is 

an acronym for mixed fission products, and ‘ACP’ stands for activation products. 

• On June 13, 1958 three employees failed to wear the respirators recommended and supplied 

by the health physicist while working beneath the fuel handling coffin. The incident report 

notes that compliance with health physics recommendations is expected. No other information 

about the incident is provided (A0374).1

• On February 6, 1959 two Atomics International personnel handled contaminated equipment 

without the proper use of protective gloves. The men were replacing a television camera that 

had been used on the reactor loading face, with a movie camera. After handling the television 

camera and its component parts, the television camera was removed. One man asked for the 

SRE Polaroid camera to take a picture of the camera setup. He received the Polaroid camera 

with contaminated gloves and when he was not able to make the camera adjustments he 

wanted, he removed the gloves by placing the fingers of the gloves between his teeth and 

withdrawing his hands. After continuing to have difficulty, both men handled the camera, one 

with gloves on and one without gloves. The picture was taken and the camera was returned to 

the SRE operations department. Both men then handled the movie camera and mounted it on 

the old television camera brackets. Again, one man had gloved hands and the other did not. 

The men then walked off the potentially contaminated plastic sheeting on the floor before 

removing their shoe covers. Smears of the plastic sheeting showed less than 30 dpm betagamma and smears of the television camera ranged from 1 to 200 dpm beta-gamma. The 

Polaroid camera and movie camera were not significantly contaminated. When questioned, 

the men stated they had used the hand and foot counter before leaving the building. No 

personnel contamination was found. One of the men was found to be new and had little 

knowledge of the hazards of radiation contamination. [This observation may speak to the level 

of training among nuclear personnel, as to the known risks and hazards associated with their 

1 EPA/HGL, Final HSA 6, p. 46, citation 1:

1 Fisher, W.L., Internal Letter, re: Non-Compliance with Health Physics Recommendation, June 16, 1958. 

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job descriptions]. (A0478).1

• On June 4, 1959 a fuel element (R-56) was removed from the SRE reactor core for transport 

to Wash Cell B for sodium cleaning. The operator noted that the fuel element appeared “quite 

dirty.” After cooling, the operator felt the fuel rod “appeared better, however, a dark material 

was noted near the upper part of the hanger rod, near the lower end of the hanger rod, at [the] 

top of [the] fuel cluster and at the lower extremity of the fuel cluster.” The dark material was not 

definitively identified in the incident report; however, it is implied to be the hydrocarbon coolant 

and lubricant tetralin or one of its decomposition products.

The fuel element was lowered into Wash Cell B by the transfer cask. A vacuum was started on 

the cell and wash water was pumped into the cell. When the vacuum was decreased to about 

2 inches of mercury, the vent valve was opened. Some “pops” were heard, but that was typical 

of the sodium-water reaction that occurred in the cleaning process. At approximately 6:55 

a.m., an abnormal rise in pressure occurred leading to an explosion, which blew the fuel 

element shield plug and hanger rod out of the wash cell and onto the reactor room floor. The 

fuel element cluster itself remained in the wash cell. Flames and smoke erupted from the 

wash cell, the radiation air monitor sounded, and the building was evacuated.

A survey reading taken through a buckled door measured 125 mrad/hr approximately 10 feet 

from the hanger rod. Absolute filter banks on the SRE building roof were surveyed. Results 

were 250 mrad/h for the west filter bank and 150 mrad/h for the east filter bank. Personnel 

entered the high bay to collect samples. Liquid material, presumed to be sodium hydroxide, 

found on the floor measured 2.5 rad/h at 6 inches. The SRE control room, mazzanine area, 

and health physics office were surveyed, but only minor contamination was detected and 

these areas were quickly decontaminated. A high volume air sample was taken in the high bay 

near the wash cells at 2:00 p.m. and airborne activity was found to be 3 x 10 9 μCi/cm3 betagamma.

On June 5, 1959 the core plug and hanger rod were placed into a moderator can storage well. 

Measurements taken at that time found the core plug to be 1 rad/h at 6 inches, the hanger rod 

to be 2.5 rad/h at 6 inches, and the lid of the moderator can to be 30 mrem/h at the surface. A

high volume air sample was taken at 11:25 a.m. and results showed airborne activity to be 2 x 

10 9 beta-gamma. Decontamination and re-surveying of the high bay began.

An investigation board noted the most probable cause of the incident was a hydrogen-oxygen 

reaction. Oxygen from the wash cell vent pipe combined with hydrogen resulting from the 

sodium-water reaction that occurs in the cleaning process to create the pressure build-up and 

explosion. Additionally, a new wash cell procedure had just been put into place, and the “hold 

down” clips used to keep the fuel element components in place in the event of a pressure 

surge were not installed. Other areas for improvement were also noted in the incident report.

A smear survey of the high bay area found results ranging from 1,700 beta-gamma near the 

center of the high bay to 214,000 dpm beta-gamma at the north end of the building near the 

wash cells. A smear survey of equipment in the high bay found contamination ranging from 

1 EPA/HGL, Final HSA 6, p. 47, citation 2:

2 Young, C. L., Atomics International Internal Letter, re: Potential Contamination of Personnel, February 19,

1959. 

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719 dpm beta-gamma on an equipment panel to 469,836 dpm beta-gamma on the transfer 

cask vent hose.

A 1961 AI report titled, “SRE Fuel Element Damage: Final Report, NAA-SR-4488,” states that 

the maximum contamination level detected as a result of the incident was 700,000 dpm/100 

cm2 taken 24 hours after the incident and found on the fuel handling cask located 25 feet from 

Wash Cell B. An environmental survey of the area surrounding the SRE showed results 

ranging from 0.03 mrad/hr northwest of the building to 0.20 mrad/h at the southwest corner of 

the SRE perimeter. Normal background for the survey meter was between .03 and .05 mrad/

hr. Soil samples collected on June 5 ranged from 9.36 x 10 6 μCi/g outside the SSFL boundary 

line to 1.65 x 10 5 μCi/g inside the SRE perimeter fence. The June 5 soil samples were 

comparable to samples taken two months earlier.1 (A0315)

In 1969, Rockwell collected a 500-cm3 sample of groundwater for radiometric analysis from 

the SRE wash cell valve pit. Results showed a beta concentration of 7.4 x 10 6 μCi/cm3. In 

November, 1969 Rockwell conducted a smear survey and found removable contamination of 

between 21 and 1,326 dpm/100 cm2. Rockwell concluded that the soil surrounding the wash 

cells was contaminated, probably as a result of the explosion that occurred in Wash Cell B in 

1959.2

• July 12, 1959 - Major Incident

On July 12, 1959, power run 14 of the SRE began, which resulted in the melting of 13 fuel 

elements. However, circumstances involved in the July 12, 1959 incident extended over the 

entire period of time from November 29, 1958 to July 26, 1959. AI reports describe the following 

sequence of events related to power runs 8 through 14 at the SRE (Incident Report A0274):

Note: [During operations described below, the core scope was used repeatedly. Releases from 

the SRE core are noted. It does not appear that employees were provided with nasal smears to 

determine possible contamination. Incident Reports A0393 (May 25, 1960) and A005 (June 9. 

1960) describe removal of the core scope and release of core gas that was contaminated with 

mixed fission product; employee nasal smears tested positive for contamination. The data may 

apply to operations and potential contamination of employees involved in the July 12, 1959 

incident and subsequent fuel recovery efforts. In addition, the contamination of walls, floors and 

equipment should be noted; such contamination is not characteristic of Xe-133 and Kr-85. 

1 EPA/HGL, Final HSA 6, p. 48, citations 1-3:

1 Investigating Board, Atomics International Internal Letter, re: SRE Wash Cell Incident of June 4, 1959, 

June 19, 1959.

2 Borg, G., Wash Cell Incident at the Sodium Reactor Experiment, AI Memo 5155, Atomics International, 

April 6, 1960.

3 Ashley, R. L., Beeley, R. J., Fillmore, F. L., Hallett, W. J., Hayward, Jr., B. R., and Jarrett, A. A., SRE Fuel 

Element Damage, Final Report, NAA-SR-4488 (suppl.), Atomics International, 1961. 

4 Ibid. 122119551 

2 EPA/HGL, Final HSA 6, p. 70, citation 1:

1 North American Rockwell Corporation Internal Letter from R. K. Owen to W. F. Heine, re: Water Seepage 

Hole in the SRE Wash Cell Valve Pit, January 15, 1970. 

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Employee exposure to mixed fission products resulting from leaking core and cover gas may 

need to be reevaluated on review of the SRE Incident Report]. 

o Run 8 began on November 29, 1958 after a shutdown of two months in which considerable 

repairs and modifications were made to the primary sodium system. During this work, the 

primary sodium was pumped back and forth several times between the primary loop and the 

primary fill tank, which was known to contain a large amount of sodium oxide. A substantial 

amount of oxide was thus introduced into the primary sodium. At the beginning of run 8, an 

unusually large spread in the fuel channel exit temperature was detected and attributed to the 

high oxide content of the sodium. The reactor was shut down to reduce the oxide in the sodium 

by cold trapping. On December 12, 1958, two fuel elements that were excessively hot and had 

black foreign material on them were removed and washed. Washing the fuel elements proved to 

be the most successful to reduce the temperature spread. Reactor operations continued 

intermittently. Some improvement in fuel channel exit temperature spread was found in running 

the reactor at elevated temperatures as well as by jiggling the fuel element one inch or less. On 

December 23, 1959 the reactor was shut down to inspect the fuel elements. Fifteen fuel 

elements were washed and additional cold trapping was done to reduce exit temperature 

spreads. The reactor was started again and the spread in the fuel channel exit temperatures 

continued to improve until the end of run 8 on January 29, 1959. During the shutdown, the 

sodium was cold trapped to less than 5 parts per million of oxygen content. No reactivity 

anomalies were observed during the run, but a cover gas sample indicated that the hydrocarbon 

1 coolant and lubricant tetralin had entered the primary sodium. The presence of tetralin indicates 

that the system was not functioning properly. 

o Run 9 began on February 14, 1959, but continued difficulties with the fuel channel exit 

temperatures forced a shutdown four days later to wash fuel elements and do further cold 

trapping. Reactor operations resumed and the temperature spread improved. Reactivity 

increased 1⁄2 percent during this run. Two reactor scrams (automatic dropping of control rods 

into reactor core to shut down reactor) occurred because of an excessive temperature drop 

across a moderator can and several other scrams occurred because of power line transients. 

Run 9 ended February 26, 1959. A fuel element was removed for routine examination in the 

SRE hot cell and was found to have a thin black deposit. The element was washed and placed 

back in the reactor. 2

o Run 10 lasted from March 6 to March 7, 1959. This run was conducted as a temperature test 

on a uranium oxide fuel element. No unusual circumstances were noted. A thimble was replaced 

at the end of the run. Later examination showed that a loss in reactivity of 1⁄4 percent occurred 

at the start of run 11.3 

o Run 11 began on March 16, 1959. Difficulties with the fuel channel exit temperatures were still 

being experienced, but the reactor continued to operate up to 20 megawatts (MW). On March 

27, 1959, several reactor scrams occurred caused by fluctuations in the main primary sodium 

1 EPA/HGL, Final HSA 6, p. 49, citation 1: 

1 

Ashley, R. L., Beeley, R. J., Fillmore, F. L., Hallett, W. J., Hayward, Jr., B. R., and Jarrett, A. A., SRE Fuel 

Element Damage: An Interim Report, NAA-SR-4488, Atomics International, November 15, 1959. 

2 Ibid. 

3 Ibid. 

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flow due to helium leaking into the primary sodium. The reactor returned to power and ran at 19 

to 20 MW until the run ended on April 6, 1959. An increase in reactivity of 1 percent occurred 

during the run. A reduction in fuel channel exit temperatures was noted. Twenty-one fuel 

elements were examined by television camera in the fuel handling cask and found to be in good 

condition. Ten days after the run, the radiation level in the main gallery was higher than 

expected. This was not surprising because it was realized that some fission product

contamination had occurred in the primary sodium. A filter was installed in the primary system, 

which collected considerable carbon-containing (decomposed hydrocarbon coolant and 

lubricant tetralin) material.1 

o Run 12 lasted from May 14 to May 24, 1959. The reactor performed normally during this 

entire period. A fuel element was examined after shutdown and there was no measurable 

change in fuel dimensions.2 

o Run 13 lasted from May 27 to June 3, 1959. Prior to run 13, the core gas radioactivity was 

found to be 1.7 x 10 3 μCi/cm 3. It was assumed to be Xe-133 and considered normal. Xe activity

had been observed after reactor operations for many months and was attributed to small pin 

hole leaks in the cladding of a few elements or to uranium contamination from the outside of 

new fuel elements. The filters were designed to remove particulates from the air; they did not 

stop noble gases from being released from the reactor building. With the exception of a scram 

caused by abnormal sodium flow rate, run 13 proceeded smoothly until May 30, 1959. At this 

time, several abnormalities in temperature measurements occurred. By June 2, 1959, it was 

clear that something occurred to impair the heat transfer characteristic of the reactor system. A

tetralin leak was detected and the run was terminated on June 3, 1959. On June 4, 1959, an 

attempt was made to wash a fuel element from the core and that resulted in the incident 

described above. In order to remove the tetralin from the primary system, a nitrogen gas 

stripping operation was used. About 3 pints of tetralin and 1,500 cm of naphthalene crystals 

(assumed to be a breakdown product of tetralin) were removed from the primary sodium 

system.3 

o Run 14 lasted from July 12, 1959 to July 26, 1959. The SRE was brought to criticality on July 

12, 1959, at 6:50 a.m. with rod positions at 46 inches as expected. At 8:35 a.m., as the reactor 

was increasing power to 0.5 MW, large temperature fluctuations were noted among the fuel 

channels. Fluctuations were about 10 ̊F compared to less than 5 ̊F fluctuation under normal 

operating conditions at 20 MW. Fuel channel exit temperatures also started to diverge more 

than normal. Operation continued until 11:42 a.m. when a reactor scram occurred due to loss of 

auxiliary primary sodium flow. The reactor outlet temperature at the time of the scram was 

485 ̊F. Criticality was re-established at 12:15 p.m. and rod-positions were still at 46 inches, 

indicating no change in reactivity during the scram. Operation continued at slowly increasing 

power levels and temperatures. 

At 3:30 p.m. both reactor room air monitors detected a sharp increase in activity. To reduce the 

activity level, the reactor pressure was lowered from 2 to 1 pounds per square inch gauge 

(psig). A survey of the reactor loading face shield revealed excessive radiation over the sodium 

1 Ibid. 

2 Ibid. 

3 Ibid. 

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level coil thimble in core channel 7. The initial reading was 500 mrad/h. A high bay air sample 

showed activity of 3x10 7 μCi/cm3 after 15 minutes of decay, and 4.5x10 8 μCi/cm3 after 90 

minutes of decay. 

At 4:20 p.m., it was noted that the filter from the air sampler showed an activity level of 160,000 

cpm. At 5:00 p.m., a sharp increase in the stack activity to 1.5 x 10 4 μCi/cm3 was noted. This 

returned to normal by 10:00 p.m. [Note: According to Incident Report A0015 (April 4, 1962) the 

stack only exhausted the hot cell and vent system effluent]. 

Also at 5:00 p.m., the radiation level over core channel 7 reached 25 rad/h and a decision was 

made to shut the reactor down and replace the thimble in channel 7 with a standard plug. At 

8:57 p.m. the reactor was shut down. The sodium level was checked and found to be normal. 

The reactor was brought to criticality again at 4:40 p.m., on July 13, 1959. Rod positions were 

49.5 inches out. No significant activity was noted in the high bay. At 5:28 p.m., the reactor power 

was at 1.6 MW and a planned increase was started. The power level persisted in rising faster 

than expected and at 6:25 p.m., the reactor was manually scrammed. The peak power was 

about 14 MW.1 After the scram, rod positions were 52 inches out as compared with 49.5 inches 

prior to the scram, but it was decided that the power excursion had not affected the reactor 

adversely. The reactor was brought up to criticality. 

At 9:00 a.m. on July 14, radioactivity in the high bay increased to 14,000 cpm on the air monitor. 

Two hours later, the source of activity had been localized to channels 29 and 50 in the core 

loading face. Seal rings were placed on the channels and the contamination levels dropped. 

Another scram occurred at 1:00 p.m. when a short circuit was introduced into the demand circuit 

for the main primary pump. Rod positions were an average of 51.1 inches prior to the scram at a 

power level of 3.7 MW. The reactor was quickly brought back to criticality. Rod positions were an 

average of 51.9 inches at 3.5 MW after recovery from the scram. 

The 1961 SRE Fuel Element Damage Final Report2 notes that the 1959 SRE Fuel Element 

Damage Interim Report stated a peak power of 24 Mw was reached. However, this value was 

obtained from a linear extrapolation, which was not valid.3 

On July 15, 1959, the reactor was pressurized and vented to reduce the radioactivity level 

caused by Xe in the reactor cover gas. This appears to have been released to the atmosphere. 

The filters were designed to remove particulates from the air; they did not stop noble gases from 

being released from the reactor building. It was then decided to shut the reactor down because 

it would not be able to reach the maximum power level necessary to continue getting the Edison 

turbine generator online while circulating through the steam generator. During the shutdown, the 

Edison loop was drained, the main airblast heat exchanger was filled, the reactor sodium level 

coil was reinstalled in core channel 7 and a complete helium leak check was conducted on the 

core loading face. 

1 Ibid. 

2 Ibid. 

3 The 1961 SRE Fuel Element Damage Final Report notes that the 1959 SRE Fuel Damage Interim Report stated a

pak power of 24 Mw was reached. However, this value was obtained from a linear extrapolation, which was not valid. 

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On July 16, 1959, the reactor again achieved criticality. The average rod position was 57 inches 

out, which indicated a substantial loss in reactivity (about 1.2 percent in the four days since the 

start of run 14). Intermittent operation continued at low power (less than 2 MW) until July 20, 

1959. During this time, measurements of primary sodium plugging temperatures were made, the 

effects of core cover gas pressure on reactivity was studied, and the reactor outlet temperature 

was gradually raised to see if this improved reactor operating conditions (as it did on run 8). 

On July 17, 1959 the following letter was written to R.E. Durand (Atomics International) by R.K. 

Owen, of the SSFL Health Physics Department: 

July 17, 1959

To: R.E. Durand

From: R.K. Owen

Subject: Airborne Radioactive Contamination in SRE High Bay 

During Reactor Operation. 

It is hereby recommended that the SRE be shut down until the 


sources of the airborne radioactive contamination in the High Bay 
 


are located and repaired. 


Intermittent airborne activity at SRE has long been a problem, 


but the primary reason for this recommendation was the condition 


occurring July 12, 1959 during power run #14 while the reactor 


was operating at about 1Mwt. At this time, the High Bay 


atmosphere became contaminated. A high volume air sample taken in 
 


the High Bay showed an airborne activity of 3x10-7 uc/cc. This 


concentration is 300 times the maximum permissible concentration 


in air for unidentified beta gamma emitters (MPC+10-9 uc/cc). An 


attempt to locate the cause of the release showed that reactor 


channel R7 containing a sodium level probe was the primary source 
 


of leakage. 


On subsequent days during this power run, other leaks were found 


in reactor channels R29 and R50 and possibly in the Cerrobend 


seal. 


Attempts at identifying this activity have to date been 


unsuccessful, however, the airborne contamination levels have in 


every case varied directly with reactor power. Even if 


identification of the activity showed it to have an MPC greater 


than 3x10-7, repair of the leaks would be desirable in order to 


prevent the release of more toxic materials in the event of a 


serious fuel rupture. 


/s/ R.K. Owen 


Health Physics Dept. 


On July 21, 1959 at 2:10 a.m., a scram was caused by a fast period indication. It is worth noting 

that “the reactor ha[d] a history of spurious scrams due to apparent period transients....Many socalled period scrams have been traced to voltage and frequency instability in the power 

suppl[y].”1 

The reactor was critical at 2:25 a.m. at an average rod position of 49 inches. Operation 

continued at 2.5 MW. At 6:45 a.m., radioactivity in the reactor started building up as indicated by 

1 Ibid. 

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continuous air monitoring system installed to protect personnel. At 10:00 a.m., two air monitors 

were reading 15,000 and 18,000 cpm. By 2:00 p.m., a high bay air sample found activity at 

2x10 9 μCi/cm3. 

On July 21, 1959 at 9:45 a.m., a reactor scram was manually initiated when flow was lost in the 

main secondary loop. This loss-of-flow-scram was caused by a low sodium level in the 

secondary expansion tank resulting from a faulty level coil. After the scram, fuel channel exit 

temperatures dropped. When the main secondary loop was restored to service, temperature 

swings were noted in the reactor cold leg. At 11:30 a.m., the reactor was again critical at an 

average rod position of 54 inches. The spread of fuel channel temperatures was still present, 

but AI’s safety limits were not exceeded. 

On July 22, 1959, the fuel temperature recorder on the element in core channel 55 showed 

fluctuating temperatures from 1,100 to 1,200 ̊F. This channel was composed of various 

experimental fuels and the point recording instrument found channel temperatures higher than 

recording instruments in the control room. The point recording instrument was felt by AI staff to 

be unreliable so no attempt was made to reduce the temperature. An instrument recorded a 

maximum temperature of 1,465 ̊F for the experimental fuel element on July 23, 1959. Operation 

continued at power levels up to 4.5 MW with sodium flow rates up to 1,500 gallons per minute 

(gpm), and reactor outlet temperatures up to 790 ̊F. 

On July 23, 1959, it was decided to shut the reactor down in view of the high fuel temperature 

for the element in channel 55 and because the fuel channel exit temperature spread was not 

improving noticeably. Reactor outlet temperature was kept between 700 and 800 ̊F, although a 

few fuel channel exit temperatures reached 900 to 1,000 ̊F. At 9:50 a.m. on July 23, 1959, a 

reactor scram occurred due to a fast period indication, but the reactor was critical again at 10:15 

a.m. Between midnight and 8:00 a.m. on July 24, fuel elements were jiggled in an attempt to 

dislodge any foreign material and lower the fuel channel outlet temperature. During this time it 

was noted that elements in core channels 10, 12, 35, and 76 were stuck in place. Another scram 

occurred at 12:50 p.m. on July 24, 1959. The reactor was critical again at 1:14 p.m. At 3:40 

p.m., another scram was caused by loss of auxiliary primary flow. Criticality was reestablished at 

3:56 p.m. Rod positions were an average of 54 inches out. 

On July 26, 1959, it was noted that elements in channels 12 and 35 were no longer stuck and 

the element in channel 76 was somewhat freer. On July 26, 1959, at 11:20 a.m., the reactor was 

shut down after a total of 16 megawatt-days (MWd) on run 14. The shutdown was made to 

examine each fuel element that had been running hot and clear away any obstructions in the 

channels. 

The first damaged fuel element was observed at 5:15 p.m. on July 26, 1959. By August 2, 1959, 

six parted fuel elements had been removed from the reactor and taken to the fuel storage facility 

before a parted fuel element from core channel 12 became lodged in the fuel handling cask 

forcing the suspension of operations. After it was determined that “a considerable amount of 

radiation exposure and contamination in the high bay area was resulting from the attempts to 

dislodge the jammed fuel element,” the plan shifted to modifying a moderator handling cask for 

fuel removal instead. The removal of the 19 remaining fuel elements began on October 8 and 

ended on October 19, 1959. The status of the fuel elements are shown in the Table 2.1 below. 

This is AI’s summary of events as of November 15, 1959. 

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AI did not issue a press release regarding the fuel element damage until August 29, 1959.1 

The HSA provides a detailed description of each fuel rod, its date of removal, and its 

characteristics upon examination. Fuel element examination showed that 13 of the 43 fuel 

elements in the SRE during power run 14 had sustained substantial damage. 

• Reactivity changes in power run 14 were also attributable to the presence of tetralin 

decomposition products in the reactor. The reactivity loss of 1.2 percent during the first four 

days of run 14 was caused by sodium entering the R-10 and R-42 moderator cans likely 

caused by abnormal conditions in the core. The reactivity increase of 0.3%, which produced 

an excursion to about 70% full power, was likely the result of sodium expulsion from several 

fuel channels in quick succession. The general conclusion of the 1961 SRE Fuel Element 

Damage report is that “no important change in [the SRE’s] physical condition [has occurred] as 

a result of the fuel element damage to the first core.”2

• Following removal of the fuel elements, inspection of the core began. The radiation level at the 

bottom of the loading face shield was measured at 43 rad/h. The fuel slugs and cladding 

fragments remaining in the core were picked up by articulated grapples, scanned, and then 

stored with other damaged fuel. Removal of damaged moderator cans showed that sodium 

had leaked into cans R-10 and R-42. Later sodium leak checks found cans R-32, R-44 and 

R-45 also contained leaks. A total of 16 cans were replaced. As a result of the fuel cladding 

failures during power run 14, an estimated 5,000 to 10,000 Ci of fission products were 

released to the primary coolant system.3

Although conducted after the end of the major incident, a gamma spectrometer scan of an air 

sample identified Xe-133 and Kr-85 as the principle remaining contaminants. AI extrapolated 

the decay of these two isotopes back to the date of the SRE shutdown on July 26, 1959 and 

estimated contamination levels of 7.4 μCi/cm3 for Xe-133 and .016 μCi/cm3 for Kr-85 by 

September 29, 1959. Fixed contamination of the primary sodium piping was found. Isotopes 

detected were Sr-89, Sr-90, zirconium-niobium-95 (Zr-Nb-95), Ce-144, and Cs-137.

The cold trap (vapor condenser) was also found to collect radioactivity with dose rates ranging 

from 50 to 80 rad/h at 2 inches from the cold trap surface. 4 There are many caveats to 

1 EPA/HGL, Final HSA 6, p. 56, citation 1:

1 Hart, R. S., Distribution of Fission Product Contamination in the SRE, NAA-SR-6890, Atomics International, 

March 1, 1962. 

2 EPA/HGL, Final HSA 6, p. 53, citation 2:

2 Atomics International News Release, During Inspection of Fuel Elements – SRE – Parted Fuel Element...,

August 29, 1959. 

3 EPA/HGL, Final HSA 6, p. 53, citation 2:

2 Atomics International News Release, During Inspection of Fuel Elements – SRE – Parted Fuel Element...,

August 29, 1959. 

4 EPA/HGL, Final HSA 6, p. 56, citation 1:

1 Hart, R. S., Distribution of Fission Product Contamination in the SRE, NAA-SR-6890, Atomics International, 

March 1, 1962. 

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interpreting sample results following the SRE fuel cladding failures. Limited sampling

opportunities and reactor recovery efforts imposed restrictions on the quantitative value of 

data collected.

• Constant monitoring for airborne contamination in the reactor room occurred during recovery 

efforts. The highest level recorded was reportedly 3x10 8 μCi/cm3 during core inspection on 

May 21, 1960. Operations had to be suspended for one hour. General contamination on the 

walls, floors, tools, etc. in the reactor room ranged from 100 to 1,500 dpm/cm2 of beta-gamma 

activity. Occasional high counts of 150,000 dpm/100 cm2 were detected and reportedly 

cleaned up.

Several instances of minor contamination of the asphalt blacktop were reported just outside 

the SRE access door. These areas were reportedly cleaned up. According to the AI 1961 SRE 

Fuel Element Damage report, with the exception of the inert gases Xe-133 and Kr-85, all of 

the fission fragments remained in the sodium or were absorbed by the metal surfaces of the 

cladding or the piping. The principle fission products found in SRE primary piping were Sr-90, 

Ce-144 and Cs-137. During SRE recovery operations, several modification were made to the 

reactor system. Tetralin was eliminated as a coolant in favor of NaK, nitrogen gas, and 

kerosene. The sodium heat transfer system was modified. Instrument monitors were added to 

the system and the fuel element geometry was altered to provide more clearance between the 

fuel element and the channel.1

• Incidents After the July 12, 1959 Major Incident

• On September 10, 1959 two AI employees entered a radioactive storage area west of the 

SRE without permission. The area contained highly radioactive, contaminated equipment and 

was normally barricaded with yellow and black rope and signs stating, “No Entry Without 

Tagged Entry Permit,” “Contact Health Physics Before Entering,” and “Contaminated 

Equipment.” Two AI health physicists discovered the barricade had been removed, a large 

trailer truck was parked 50 feet inside the barricade, and three outside contractors were 

loading radioactive waste onto the truck. These contractors were wearing protective clothing. 

In the southwest section of the area, three people (two AI employees and a contractor) were 

examining material without wearing any protective clothing. One of the health physicists 

requested that the men leave the area and one of the AI men responded that they were busy. 

When the men eventually started to leave, the health physicist requested that they remain for 

a contamination check and the men refused. One of the AI men stated they were late for a 

meeting and that the health physicist and his supervisor had better not interfere with the 

operation. As a result of this violation, the AI men and the contractor, subjected themselves to 

a high radiation field (A0351).2

1 EPA/HGL, Final HSA 6, p. 57, citation 1:

1 

Ashley, R. L., Beeley, R. J., Fillmore, F. L., Hallett, W. J., Hayward, Jr., B. R., and Jarrett, A. A., SRE Fuel 

Element Damage, Final Report, NAA-SR-4488 (suppl.), Atomics International, 1961. 

2 EPA/HGL, Final HSA 6, p. 57, citation 2:

2 Borg, G., Atomics International Internal Letter, re: Account of Incident on September 10, 1959 at about 10 

a.m. in the Radioactive Waste Storage Area West of the SRE as per G. Borg, September 11, 1959. 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
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• On November 21, 1959, an employee was modifying an SRE core recovery tool when an exit 

survey found contamination on his nose, upper chest, and undershorts. The undershorts had 

to be discarded as contaminated waste. Nose swipes indicated contamination up to 3,600 

dpm beta-gamma. Smears of the face and upper chest found contamination at 400 and 5,600 

dpm beta-gamma, respectively. The employee’s dosimeter worn at the time of the incident 

gave a reading of 50 mR. After a shower and repeated cleaning of the nostrils, no significant 

contamination was found. A positive urine specimen did come back and the employee was 

restricted from tagged areas until additional results were received (A0390).1

• On December 7, 1959 an employee violated standard operating procedures and did not notify 

the health physics department that he was entering the high bay area of the SRE, a red 

tagged area. He also failed to wear the proper clothing required for this high radioactive 

contamination area. The incident report notes the employee was reminded of proper 

procedure (A0548).2

• On December 18, 1959 an employee was found working in the SRE Hot Cell Service Area, a 

red tagged area highly contaminated with beta-gamma activity, without a protective hood. A

health physicist verbally informed the employee that he was in violation of his entry permit and 

he needed to leave the area. The employee stated that no hoods were available and he 

continued to work. When the employee finally left the area, he did not check himself for 

radioactive contamination as required by his entry permit. The employee was informed that it 

was his responsibility to follow the requirements of his work permit (A0391).3

• On April 11, 1960 an employee’s nose became contaminated up to 190 dpm beta-gamma 

when the viewing glass of the remote cell he was working in started leaking. He was 

immediately taken out of the area and decontaminated (A0510).4

• On May 15, 1960 a leak occurred in the tank used for containing sodium, removed from the 

reactor. The sodium ignited upon contact with the atmosphere causing a large volume of 

smoke. Approximately 200 g of sodium containing fission products was burned. A 30-minute 

air sample was taken about 10 minutes after the fire occurred. Beta-gamma activity was 

1.3x10 10 μCi/cm3 on immediate counting. After 1 hour the activity level decreased to 5.0 x 

1 EPA/HGL, Final HSA 6, p. 57, citation 3:

3 Young, C. L., Atomics International Internal Letter, re: Radiological Safety Incident Report, Sodium Reactor 

Experiment, 11/21/59, December 7, 1959. 

2 EPA/HGL, Final HSA 6, p. 57, citation 4:

4 Bell, C. E., Atomics International Internal Letter, re: Violation of Health Physics Practices, December 11, 

1959. 

3 EPA/HGL, Final HSA 6, p. 58, citation 1:

1 Lang, J. C., Atomics International Internal Letter, re: Notice of Rule Infraction, December 31, 1959. 

4 EPA/HGL, Final HSA 6, p. 58, citation 2:

2 Shannon, J. W., Atomics International Internal Letter, re: Radiological Safety Incident Report, SRE Hi Bay 

Area, 4/11/60, April 11, 1960. 

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10 11 μCi/cm3. (A0394).1

• On May 25, 1960 a core scope from the SRE core was being removed and came out of the 

gas lock because there were no markings to identify the upper limit of the equipment when it is 

in the gas lock. When the core scope came out, core gas contaminated with mixed fission 

products escaped for a short period. Two employees received nasal contamination ranging 

from 47 to 392 dpm beta-gamma. The employees were decontaminated and a 

recommendation was made to mark all equipment going through a gas lock with an upper limit 

(A0393).2

• On June 2, 1960 an employee was slightly injured and contaminated while working on an SRE 

fuel coffin. The incident report has not been located (A0460). 3

• On June 9, 1960 a core scope from the SRE reactor core was being removed through a 6-

inch gas lock when it was discovered that the valve on the gas lock had not completely closed. 

Immediate attempts were made to close the valve, but in the brief time it was open, gas 

containing mixed fission products escaped from the SRE reactor core into the high bay. Nasal 

smears were done and employees were found to be contaminated between 18,000 and 

26,000 dpm beta-gamma. At least one employee had hair and facial contamination of 5 mrad/

h. The contaminated employees were immediately removed from the high bay area and 

decontaminated. At least one employee was restricted from further exposure until bioassay 

results were obtained (A005).4

• On July 6, 1960 gas contaminated with mixed fission products escaped into the high bay 

atmosphere. An employee was measuring a moderator can through a window in the gas lock 

that had a plastic bag covering it and the employee’s arm. As the employee removed his arm 

from the plastic bag, the gas escaped and set off the continuous air monitor. The plastic bag 

was quickly sealed and the employee was advised to leave the high bay for nasal smears. 

Fifteen minutes later, the employee had not reported to health physics and claims his chief 

operator did not deem it necessary to leave. Eventually three men were given nasal smears 

as a precaution. Results ranged from 0 to 3,168 dpm beta-gamma. The employee with the 

highest reading was decontaminated down to 22 dpm beta-gamma for the right nostril and 10 

1 EPA/HGL, Final HSA 6, p. 58, citation 3:

3 Galperin, A., Atomics International Internal Letter, re: Radiological Safety Incident Report, SRE High Bay, 

5/15/60, May 26, 1960. 

2 EPA/HGL, Final HSA 6, p. 58, citation 4:

4 Galperin, A. and Shannon, J. W., Atomics International Internal Letter, re: Radiological Safety Incident

Report, SRE High Bay, 5/25/60, August 30, 1960. 

3 EPA/HGL, Final HSA 6, p. 58, citation 5:

5 Boeing Radiation Incident Database, 2010. 

4 EPA/HGL, Final HSA 6, p. 59, citation 1:

1 

Lane, W. D., Atomics International Internal Letter, re: Radiological Incident Report, SRE High Bay, 6/9/60, 

July 20, 1960. 122119552 

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dpm beta-gamma for the left nostril (A0442).1

• On December 15, 1960 two employees became contaminated during a sandblasting

operation. A sodium pump was decontaminated by sandblasting in a plastic tent fabricated for 

this purpose. The pump was contaminated with mixed fission products and a maximum 

radiation level was measured at approximately 5 rad/h. One employee was working in full 

protective clothing and face mast inside the tent, while the other employee was operating the 

sandblasting compressor outside the tent. After approximately 3 minutes, the employee in the 

tent realized dust was leaking into his face mask. He motioned to have the compressor shut 

off and stepped out of the tent where he was helped out of his protective gear. Nasal swipes of 

both men revealed beta-gamma contamination of 75,000 dpm (employee working in tent) and 

420 dpm (employee outside tent). A survey meter check found one employee’s clothing 

contaminated up to .2 mrad/h. The clothing was disposed and the men were decontaminated 

(A0305).2

• On May 17, 1961 an employee working in the SRE high bay discovered contamination on his 

shoes upon exiting for lunch. Investigation found that the rubber boots he had been wearing 

as required health and safety dress had developed leaks. After unsuccessful attempts to 

decontaminate the shoes, they were disposed as radioactive waste (A0435).3

• On November 4, 1961 an employee working on samples from “hot trap (heater) A” in the SRE 

high bay became contaminated while grinding a sample container away to expose the sample. 

The external surface of the sample container was less than 30 dpm beta-gamma, but the 

sample itself was 3 rad/h. The employee was wearing a hard hat with a face shield for grinding 

operations and was told that a full respirator would be required when he got close to exposing 

the sample. The employee stated that he had done this type of work before and had never 

been contaminated. Upon further observation from a health physicist, the sample had indeed 

been exposed by grinding operations. A smear survey was taken and indicated high internal 

contamination. The employee further stated that the air flow through the high bay door was 

enough to protect him from the exposure. The health physicist went to get a face mask for the 

employee, but none were available. Nasal smears indicated the employee’s right nostril was 

contaminated to 145 dpm beta-gamma and the left nostril was contaminated to 372 dpm betagamma. Contamination had also filtered through the employee’s cotton gloves and his hands 

1 EPA/HGL, Final HSA 6, p. 59, citation 2:

2 Galperin, A., Atomics International Internal Letter, re: Radiological Safety Incident Report, SRE High Bay, 

7/7/60, August 31, 1960. 

2 EPA/HGL, Final HSA 6, p. 59, citations 3 & 4:

3 Klostermann, J. P., Atomics International Internal Letter, re: Radiological Safety Incident Report, R/A

Material Storage Area – SRE, 12/15/60, December 19, 1960 

3 EPA/HGL, Final HSA 6, p. 59, citation 5:

5 DuBois, P. R., Atomics International Internal Letter, re: Radiological Safety Incident Report, SRE High Bay, 

May 17, 1961, May 31, 1961. 

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Page 83 of 418 

were contaminated to 2 mrad/h. The employee was decontaminated and released (A0515). 1

• On December 23, 1961 an employee entered a red tagged area without a tagged entry permit 

and manipulated high contaminated core lights without protective gear or without following 

proper procedure for working with radioactive materials (A0472).2

• On February 6, 1962 an employee was observed in the Health and Safety Counting Room 

using counting equipment to measure radioactivity on a smear because he didn’t believe the 

results of a health physicist’s survey of the plastic cover on a storage cell. The employee failed 

to detect the radioactivity (3x105 dpm / 100 cm2) observed before and after by health physics 

analysts. Not only did the employee delay decontamination of the object, but he violated rules 

by using radiation detection equipment he was not trained to use (A0534).3

• On April 4, 1962 two sets of air samples for radioactive particulates were taken on the roof of 

the SRE near the exhaust fans and filters. At least 90% of the high bay particulate activity was 

rubidium-88 (Rb88). One sample set was taken with the filters bypassed and the other sample 

set was taken with air exhausted through the filters. The reactor was at 5 MW and only the 

east fan was on. Air samples from the high bay were taken at the same time as samples on 

the roof. These results were compared to sample results on March 1, 1962 with the reactor 

critical.

Additionally, a roof sample had been taken on March 3, but it could not be correlated to high 

bay activity because it was taken in the stack, which only exhausts the hot cell and vent 

system effluent. Results were as follows: 

• Air Sample Data at the SRE April 4, 1962: 90% rubidium-88 (Rb88)

-11 3 

o April 4, 1962, 9:30 a.m. filtered exhaust – 5.5 x 10 μCi/cm beta-gamma

o April 4, 1962, 9:30 a.m. high bay activity – 9.4 x 10-9 

μCi/cm3 beta-gamma

o April 4, 1962, 1:30 p.m. nonfiltered exhaust – 5.9 x 10-9 

μCi/cm3 beta-gamma

o April 4, 1962, 1:30 p.m. high bay activity – 9.9 x 10-9 

μCi/cm3 beta-gamma
 

-11 3
 o March 3, 1962, stack sample – 2.04 x 10 μCi/cm beta-gamma

1 EPA/HGL, Final HSA 6, p. 60, citations 1 & 2:

1 

Yarrow, A. R., Atomics International Internal Letter, re: Radiological Safety Incident Report, Bldg. 143 (High

Bay), 11/4/61, November 7, 1961.

2 Health and Safety Department, Atomics International Internal Letter, re: Notice of Health and Safety Rule

Infraction, November 8, 1961. 

2 EPA/HGL, Final HSA 6, p. 60, citation 3:

3 Health and Safety Department, Atomics International Internal Letter, re: Notice of Health and Safety Rule

Infraction, January 4, 1962. 

3 EPA/HGL, Final HSA 6, p. 60, citation 4:

4 Health and Safety Department, Atomics International Internal Letter, re: Notice of Health and Safety Rule

Infraction, February 16, 1962. 

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-10 3 

o March 1, 1962, roof – 1.5 x 10 μCi/cm beta-gamma
 

-10 3
 o March 1, 1962, high bay – 1.1 x 10 μCi/cm beta-gamma (A0015).1

• On June 13, 1962 an operator was engulfed in fire during the transfer of core heaters.2

• On July 30, 1962 an employee received an exposure exceeding the weekly limit during a 

source transfer operation. A 200 Ci Sb-124 source was being transferred from one storage 

cask to another using a bottom loading cask suspended from a crane over the cask containing 

the source. A cable was then used to pull the source up into the suspended cask. As the 

source was being drawn up, the suspended cask began to oscillate. The employee 

approached the cask to steady it so the source could be drawn into it and by so doing 

exposed himself to the unshielded source. The employee’s film badge indicated a dose of 

2,150 mrem gamma, but approximately 400 mrem of the film badge exposure was received 

prior to the incident. The employee was put on restriction not to exceed 100 mrem per week 

exposure for the remainder of the quarter (A0404).3

• On October 20, 1962 a three-man crew cutting up obsolete core heaters in the northeast 

corner of the SRE yard became contaminated. The eight heaters were monitored before the 

operation. All heaters were decontaminated to less than 150 mrad/h at the surface. The men 

were dressed in redline coveralls, shoe covers, and gloves for the operation. At lunch time, the 

men went to the hot change room to change out of their protective gear and contamination 

was found on their clothing and bodies. Nasal contamination ranged from 70 to 1,200 dpm, 

hand contamination ranged from 3 to 5 mrad/h, and radiation exposure ranged from 60 to 70 

mrad. All men had scattered face and body contamination and two men had contamination on 

all of their clothes. The men were given continuous showers until contamination levels were 

brought to background. Further testing was prescribed for internal contamination. The 

contaminated employees reportedly contaminated most doorknobs in the lower level of the 

SRE building (up to 300 dpm/100 cm2) and a large area of the floor (up to 600 dpm/100 cm2).

These areas were decontaminated to 30 dpm/100cm2 (A0379).4

• On March 19, 1964 approximately 3,550 gallons of water were dumped from two new 

radioactive liquid waste storage tanks located on the north side of the SRE as a part of the 

1 EPA/HGL, Final HSA 6, p. 60, citation 1:

1 

Badger, F. H., Atomics International Internal Letter, re: Air-borne Activity of SRE High Bay, April 6, 1962. 

2 EPA/HGL, Final HSA 6, p. 60, citation 2:

2 

Boeing Radiological Incident Database, 2010. 

3 EPA/HGL, Final HSA 6, p. 61, citation 3:

3 Raper, R. R., Atomics International Internal Letter, re: Radiological Safety Incident Report, SRE West Pad, 

7/30/62, September 6, 1962. 

4 EPA/HGL, Final HSA 6, p. 61, citations 4 & 5:

4 Unknown Author, Atomics International Internal Letter, re: Incident Report, Contamination of Personnel 

Working on Disposal of Outworn Core Heaters, October 20, 1962.

5 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. G-16-17. 

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pre-operation testing for the PEP program. 24-hours later it was discovered that the water was 

contaminated with an estimated 5.8 x 104 μCi of irradiated corrosion products. A new 

radioactive liquid waste storage system was installed outside the north end of the SRE 

building. The system included one used tank (T-1) and two new storage tanks (T-2 and T-3). 

T-1 was a 150-gallon tank. T-2 was a 350-gallon tank installed alongside T-1 in the sump pit 

40 feet below ground level. T-3 was a 3,200-gallon tank installed at ground level. Piping was 

installed to allow gravity flow of radioactive water to either T-1 or T-2. The waste waster was 

then pumped to ground level where it could be sampled. Activity level permitting, the waste 

water could then be pumped to T-3 for storage and disposal. T-1 and T-2 were vented to the 

existing radioactive gas compressor vent system serving the fuel wash cells.

The old liquid lines to T-1 were closed and T-2 and T-3 were filled with clean water from a 

nearby fire hydrant to check level indicators and pumps. The next day the water was emptied 

from the tanks so that the system could go through final checkout. When a health and safety 

representative came to monitor the circulation lines during final testing, he discovered that two 

lines from T-2 were radioactive, while the lines to T-1 were at what was characterized at the 

time as “background.” Apparently, the radioactive gas vent lines had contaminated 

condensate. A portion of the vent system had been drained of condensate the day before and 

valved off, but the remainder of the vent line was left open to T-1 and T-2. Condensation 

accumulated in the open vent line and entered T-2 by gravity flow contaminating the clean 

hydrant water in the tank. The water was later analyzed and found to have approximately 18% 

Co-60, 13% Co-58, 62% Cr-51, and 7% Zr-95-Nb-95.

Once the contamination was discovered the SRE area was secured and a Class 1 Emergency 

was proclaimed. A 150 cubic foot air sample taken in the spill area found airborne activity at 

8.3 x 10 11 μCi/cm3 beta-gamma. Survey of personnel in the area found low-level radioactive 

contamination on shoes only. The maximum contamination on shoes was 2,000 dpm betagamma. Vehicles in the area were surveyed and smeared. The maximum contamination 

detected inside any vehicle was 1,800 dpm/100 cm2 beta-gamma and the maximum 

contamination outside any vehicle was 4 mrad/h. Contamination was also noted at the north 

door of the SRE building, where 800 dpm/100 cm2 beta-gamma was detected.

According to the incident report, the major spread of contamination, however, was to the 

watercourse with low-level contamination spread to no more than 500 feet from the point of 

dumping on the asphalt. After people and vehicles were decontaminated to less than 30 dpm/

100 cm2 they were released from the area. A water sample was taken from the 750,000-gallon 

SRE retention pond and the drain pump for the pond was secured to prevent release. The 

pond had not been dumped since the incident. Pond water measured 7.1x10-6 μCi/cm3 . The 

asphalt area where the water was first released measured 10 mrad/h with about 3 mrad/h 

gamma contamination. The SRE ditch measured a maximum 60 mrad/h. The asphalt was 

hosed down with water and contamination levels dropped to 4mrad/h, but the maximum 

contamination in the ditch increased to 80 mrad/h.

Operations decided to abandon efforts to flush the asphalt and brought in a crew to clean out 

the ditch, and use a floor scrubber and vacuum on the asphalt to decontaminate it. The 

asphalt was decontaminated to a maximum of 0.7 mrad/h. Four 55-gallon drums of mud and 

debris from the ditch were removed, which reduced contamination from 80 mrad/h to 15 mrad/

h. Further decontamination reduced the ditch contamination to less than 2 mrad/h, except for 

a small area reading 10 mrad/h that was to be removed later. Any areas indicating up to 2 

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mrad/h were to be fixed with an additional layer of asphalt.
 


 

Rain occurred intermittently for two days during cleanup operations and runoff water filled the 

SRE pond to within 6 inches of the spillway leading to Brandeis Canyon and Simi Valley on 

March 23, 1964. The SRE pond was drained to Rocketdyne’s Area II ponds to prevent 

overflow. [The Rocketdyne Area II ponds served the North American Aviation Saturn V Lunar 

Team, and provided water for personnel to hose down the work area following rocket engine 

testing. The Area II ponds were part of the Site Wide Water Reclaim System].

A sample of the SRE pond when this began indicated 1.8 x 10 6 μCi/cm3. Six hours later a 

sample indicated 4.1 x 10 7 μCi/cm3. Samples taken at the inlet of the first Area II pond and 

midpoint of the second Area II pond 9 hours into the drainage process found 1.4 x 10 7 μCi/ 

cm3 and 6.7 x 10 8 μCi/cm3. This compares to an average of 6.8 x 10 8 μCi/cm3 for a 3-month 

period the previous year. Ultimately, 421,000 gallons of water from the SRE pond was drained 

to the Area II ponds. It was calculated that the maximum release of contamination from the 

incident was 5.8 x 104 μCi/cm3, with 5.2 x 103 μCi/cm3 released in the SRE pond.

According to the incident report, of the major isotopes involved in contamination, the most 

restrictive maximum permissible concentration at the time for water was 6x10 5 μCi/cm3. With 

a total release restriction of 1 Ci per year and a release of approximately 3 mCi, it appeared 

that the standards established by the AEC for liquid disposal were met. The incident report 

notes that the possibility of significant contamination offsite “appears remote.” Failure to detect 

the release until 24 hours after it happened was attributed to failure of personnel involved to 

request monitoring of the water, failure of personnel to check out on the hand-and-foot 

counter, failure of routine surveys of the building floors to detect increased contamination 

levels, and lack of frequency in outside surveys (A0030).1

• On June 15, 1964 an employee vacuuming contamination in the fuel handling machine service 

pit became contaminated. Nose swipes found 90 and 450 dpm beta-gamma in his nostrils and 

fingers on his right hand were contaminated to 300 dpm beta-gamma. The employee was fully 

suited and wearing a full face mask while he was working. A few hours later, another 

employee wearing the same gear exhibited similar results after being in the service pit for 20 

minutes. His nostrils were contaminated to 120 and 4,000 dpm beta-gamma. It was 

determined that the air filter equipment was not rated high enough for the contamination levels 

in the service pit and supplied air should have been used. The airborne radioactive 

contamination was mixed fission products from the MK-1 fuel handling machine (A0409).2

• On June 17, 1964 two employees working in the fuel handling machine service pit were found 

to be contaminated. One employee was contaminated to a maximum of 0.3 mrad/h on the 

front of his neck and had a nasal smear of 1,525 dpm beta-gamma. The other employee was 

contaminated to a maximum of 0.4 mrad/h on the back of his neck and had a nasal smear of 

3,675 dpm beta-gamma. Both men were decontaminated, but internal contamination was still 

1 EPA/HGL, Final HSA 6, p. 63, citation 1:

1 

Badger, F. H., North American Aviation Internal Letter, re: SRE R/A Liquid Release, April 17, 1964. 

2 EPA/HGL, Final HSA 6, p. 63, citation 2:

2 

Bergstrom, W. H., Atomics International Internal Letter, re: Incident Report, Fuel Handling Machine Service

Pit, June 15, 1964, June 24, 1964. 

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present in sputum, spit and breath. The men were wearing full face masks with supplied air, 

double coveralls, hoods, two pair of plastic and one pair of canvas shoe covers, and two pair 

of surgeon’s gloves while working. Smears of one of the masks revealed 13,000 dpm betagamma inside and 84,000 dpm beta-gamma outside. The incident report notes that there was 

extremely high surface and airborne contamination beneath the MK-1 fuel handling machine 

and the incident occurred during removal of Core 1 fuel. No immediate cause was found for 

the mechanical means of by which the employees inhaled contamination through their 

protective gear (A0493).1

• On June 21, 1964 an SRE shift supervisor having been in the MK-1 fuel handling machine 

service pit was checked out and found to be contaminated. He had been in the pit for 

inspection prior to component disassembly. A count rate meter indicated approximately 300 

cpm on his left forearm and 500 cpm on his shoes. Another employee then exited and was 

also found to be contaminated to 300 cpm. Nasal smears for the men were 30 and 280 dpm 

beta-gamma. Apparently, a component from the fuel handling machine had been dropped and 

caused a cloud of contamination in the tented area. A check of the area indicated 1,000 to 

150,000 dpm/100 cm2 beta-gamma contamination all over the west end of the high bay. An air 

monitor on the east side of the plastic screen dividing the east and west high bay revealed a 

200 cpm increase in airborne contamination. Spots smears of the area revealed less than 30 

dpm beta-gamma. Personnel had reportedly taken off the full face mask in a portable change 

room set up adjacent to the service pit during a time when airborne contamination was still 

significant. Recommendations included providing filtered exhaust to the tented areas of work 

and use proper donning and doffing of protective gear, including removing face masks after 

outer garment (A00380).2

• On September 18, 1964 a crew began work in the SRE to remove Core I fuel, cladding and 

debris from the “Poor Man’s Hot Cell.” Full protective clothing was worn with the exception of 

full face respirators, as the plan was to set up equipment first and begin operations later in the 

shift. Grappling tools and a bucket were lowered into the pit. After tedious manipulation with 

the grappling tool, one 18-inch piece of cladding, possibly containing NaK and up to three fuel 

slugs, was precariously placed in the bucket. Full face respirators should have been donned at 

this point, but difficulty was encountered in removing the rod from the bucket and the 

immediate action of all three personnel was necessary to prevent the rod from falling back into 

the pit and starting a fire. Upon securing the rod on a ledge, work ceased and all personnel left 

the area. An immediate contamination survey of two crew members revealed no 

contamination; however nose swipes revealed nasal contamination of 1,380 and 9,000 betagamma. A third crew member had a nose swipe of 400 dpm beta-gamma after his shower, but 

the rest of his body was clean. The men were all decontaminated and placed on restriction 

pending bioassay analysis. The contamination was thought to result from moving the rod to 

1 EPA/HGL, Final HSA 6, p. 64, citation 1:

1 

Unknown Author, Atomics International Internal Letter, re: Incident Report, FHM Service Pit,

6-17-64, June 24, 1964. 

2 EPA/HGL, Final HSA 6, p. 64, citation 2:

2 

Badger, F. H., Atomics International Internal Letter, re: Incident Report, SRE High Bay West, June 21, 

1964, July 1, 1964. 

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the ledge and creating airborne contamination (A0382).1

•	 On December 1, 1964 a probe was wrapped and lowered into a control rod thimble in storage 

to obtain a radiation measurement. Plastic sheeting was laid on the floor and the employee 

conducting the survey wore shoe covers and gloves. After getting the measurement, the probe 

was carefully unwrapped to avoid contamination spread to the probe or the immediate area. A

contamination survey of the probe revealed a significant reading on the count rate meter 

despite precautions. The employee received 0.3 mrad/h contamination on his hands, between 

2,000 and 6,000 dpm of contamination over his face, neck, and the front of his clothing, and 

1,300 dpm of contamination on a nasal smear. A smear survey of the west high bay area after 

operation revealed contamination levels from 1,500 to 1,800 dpm/100 cm2 . Decontamination 

reduced levels to non-detect (A0545).2

• On December 18, 1964 size reduction of two control rods led to contamination of the 

surrounding area, equipment, and personnel. A tagged area entry permit was submitted for the 

transfer, cutting and canning of control rods in the SRE high bay. Prior to operations, the west 

high bay floor was covered with plastic and plastic barriers were set up around the storage 

cells. The work area was roped off and designated a red tagged area. Two control rod lower 

thimbles were transferred from the canisters in moderator storage cell 2 to the fuel handling 

machine. They were transferred to two fuel cans, one in storage cell 25 and the other in 

storage cell 74. The lower 93 inches of each thimble was cut off and remained in the storage 

cell. The upper 129 inches of each thimble was transferred from the fuel handling machine to 

the canister in moderator storage cell 2. The radiation levels encountered during the cutting 

and transfer operations were measured. At 3 feet from the thimble surface, activity was 

measured at 30 and 36 r/h (units may be incorrectly presented in the source document) for the 

two thimbles. At the surface of the cutting point, activity was measured at 11.5 and 12 r/h 

respectively.

Personnel exposures for three employees at this time were 45, 225 and 240 mr (units may be 

incorrectly presented in the source document). Follow-up smear samples of the area found 

3,000 dpm/100 cm2 beta-gamma on plastic outside the work area, 800 dpm/100 cm2 betagamma on a new full-face mask located south of the storage cells, 300 dpm/100 cm2 beta 

gamma on an uncovered portion of the floor south of the storage cells, and 100 dpm/100 cm2 

on the floor between the hot change room and the door leading to the high bay. One worker 

had a nasal smear indicating 600 dpm beta-gamma.

A cleanup of the contaminated areas began and another smear survey was conducted. 

Contamination remained on the hot change room floor (120 dpm/100 cm2), floor between the 

hot cell stairs and high bay door (120 dpm/100 cm2), and high bay floor just inside the hot 

change room doow (180 dpm/100 cm2). Other surveys found up to 2,400 dpm/100 cm2 betagamma on items in the welder’s work area, which prompted immediate roping off of the area 

1 EPA/HGL, Final HSA 6, p. 64, citation 3:

3 Tworek, D. D., Atomics International Internal Letter, re: Incident Report, “Poor Man’s Hot Cell,” 9-18-64, 

October 6, 1964. 

2 EPA/HGL, Final HSA 6, p. 65, citation 1:

1 

Unknown Author, Atomics International Internal Letter, re: Incident Report, SRE High Bay, 12-1-64, 

December 2, 1964. 

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and decontamination efforts. It was not clear at the writing of the incident report what caused 

the contamination of the one employee and the surrounding area. The procedure had followed 

established guidelines. It was recommended that the open moderator storage cell be covered 

with a lid or plastic between transfers and that work periodically cease to allow for health and 

safety monitoring in the future (A0371).1

• On January 14, 1965, the movement of a temperature probe caused radiation exposure in 

excess of guidelines for three personnel. Reevaluation of the personnel exposures suggested 

they were lower than reported because the film badge supplied used a linear formula to 

calculate exposures instead of a more accurate calibration curve. The corrected exposures 

showed that the three personnel involved in the incident received lifetime exposures well 

below the permissible level (A0296).2

• On February 20, 1965 an SRE core light was removed and replaced with a short plug. 

Personnel performed the operation manually with 0.5-inch water pressure in the core. The 

core light was brought up into a bag, which was tied off and cut. An extension of the core 

channel was removed and the plug was readied for insertion when particles of ignited sodium 

and contamination were released. The plug was dropped in place almost immediately, 

stopping the release of contamination. The spread of contamination was negligible but 

personnel were contaminated. Two workers had contamination on their necks (180 and 300 

dpm beta-gamma) and two workers had positive nasal smears (180 and 2,400 dpm). The 

health physics department was not given advanced notice of the operations and was not 

aware that the core would be open. Personnel indicated that this activity had been performed 

in the past without incident. No masks were worn during the operation (A0386).3

• On February 27, 1965 an employee was contaminated through two pairs of coveralls as he 

worked to change a door seal inside a moderator cask. The employee was contaminated to 

about 500 cpm on his neck. Only his head and shoulders were inside the cask and he was 

wearing a full face mask. The employee was observed the entire time by the health and safety 

department and no unusual occurrences happened while performing the operation (A0416).4

• On April 24, 1965 an employee was injured in a non-radiological operation in the SRE high 

bay. The employee was drilling an electrical circuit box and the drill passed through the tip of 

1 EPA/HGL, Final HSA 6, p. 66, citation 1:

1 Townsend, R. I., Atomics International Internal Letter, re: Radiological Safety Incident Report, High Bay 

Area of SRE, Building 143, Santa Susana, December 18, 1964. 

2 EPA/HGL, Final HSA 6, p. 66, citations 2 & 3:

2 

Boeing Incident Database, Reviewed 2011.

3 Correspondence from Remley, M. E., Atomics International, to Levy, J., U.S. Atomic Energy Commission, 

re: Type B Radiation Exposures, April 2, 1965. 

3 EPA/HGL, Final HSA 6, p. 66, citation 4:

4 

Bergstrom, W. H., Atomics International Internal Letter, re: SRE High Bay, 2-20-65, February 24, 1965. 

4 EPA/HGL, Final HSA 6, p. 66, citation 5:

5 Owens, D. E., Atomics International Internal Letter, re: Incident Report, SRE High Bay, 2-27-65, March 2, 

1965. 

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his third finger. He was taken to the West Valley Baptist Hospital for treatment and later 

returned to work. The incident was not radiological in nature, other than it occurred in the SRE 

high bay (A0444).1

• On March 22, 1965 an employee cut the tip of his thumb while attempting to adjust the 

position of a temporary ring in a sodium line prior to welding. A small area of skin on his right 

wrist was found to be contaminated to 60 dpm beta-gamma. This contamination was easily 

removed by flushing with water. All other survey results showed background activity (A0445). 2

• On August 19, 1966 an investigation took place to identify the cause of a recent incident at the 

SRE Tab Exposure Facility (TEF) resulting in exposure of operating personnel to primary 

sodium. At the time of the incident, personnel had inserted the tabs into the line and were 

preparing to expose them to primary sodium at nominal operating temperatures when the 

primary sodium started filling the piping and gas lock used for tab insertion and removal. A

freeze stem temperature alarm went off and an operating crew member found drops of sodium 

on the floor below the gas lock. The TEF inlet and outset valves were closed and the heaters 

were turned off so that the TEF was isolated from the primary sodium line. The incident 

occurred because the temperature alarm on the TEF was out of calibration and alarmed at a 

higher temperature, preventing cooling operations from being implemented to keep the freeze 

stem frozen. The occurrence could have been much worse during reactor operations, so 

standard operating procedures were changed to include adding a lead shield to access ports 

when the reactor was operating, helium leak testing the TEF prior to melting sodium in the 

piping and exposure site, and freezing adjacent piping while sodium was in the exposure site 

among other changes (A0041).3

• On December 8, 1967 radioactive water was discovered in two 8-inch pipes that penetrate the 

floor of the maintenance cell at the SRE. Samples indicated radioactivity of 6.10 4 μCi/cm3,

which was principally Cs-137 with small amounts of Sr-90, Y-90, and Co-60. According to the 

incident report, the radioactivity found in the water samples were comparable to standards for 

occupational exposure and were no more than a factor of 10 greater than the 1967 standards 

for exposures in uncontrolled areas, with the exception of Sr-90 in both cases. However, 

further analysis of the liquid indicated that only about 10 to 15% of the total activity in the 

water was Sr-90. The total activity in the water together with the loose or removable 

contamination in the pipes was determined to be something less than 0.5 mCi. The incident 

report states that there is no potential hazard of safety problem associated with the 

contaminated water, but does note that the accumulation of water is an engineering concern 

1 EPA/HGL, Final HSA 6, p. 66, citation 6:

6 Moore, J. D., Atomics International Internal Letter, re: Incident Report, SRE High Bay, 4-24-65, May 6,

1965. 

2 EPA/HGL, Final HSA 6, p. 67, citation 1:

1 

Moore, J. D., Atomics International Internal Letter, re: Incident Report, Building 143 Primary Na Pipe 

Gallery, Mar. 22, 1965, May 22, 1965. 

3 EPA/HGL, Final HSA 6, p. 67, citation 2:

2 Southward, B. G., Atomics International Internal Letter, re: Investigation of Incident at SRE Tab Exposure 

Facility, August 19, 1966. 

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(A0321).1

• On October 23, 1976 radioactive gas escaped during the removal and replacement of the 

R-34 instrumentation probe from the core top at the SRE. A health physicist took an air sample 

2 inches from inside the plug hole where the probe was located. On his way back to evaluate 

the sample, he discovered his shoes were contaminated. Results of the air sample were 2.0 x 

10 7 μCi/cm3 for the immediate count and 2.0 x 10 8 μCi/cm3 for the delayed count. The health 

physicist reported the suspected contamination around the core and a possible airborne 

release to the operating personnel. Operating personnel were almost finished reinserting the 

probe so a 3-minute air sample was taken as the probe was completely replaced. Results of 

this air sample were 6.3 x 10 9 μCi/cm3 for the immediate count and 1.7 x 10 11 μCi/cm3 for the 

delayed count. The stationary air monitor 5 feet from the core face indicated no increase in 

radioactivity above background. Four personnel were contaminated between 400 and 500 

counts above background. No contamination was found after showering. Nasal smears were 

below 20 dpm beta for personnel. A smear survey found maximum removable contamination 

at 10,000 dpm / 100 cm2 beta. Decontamination began to remove the gross contamination as 

much as possible before the end of the work shift. The SRE high bay was roped off until 

additional smears and decontamination were completed. The cause of the rease was 

attributed to the core pressure being zero and not negative. Operations involving the removal 

of plugs from the core were to have verification of negative pressure. Three days later the 

SRE high bay returned to routine working status (A0289).2

• On August 1, 1977 the bottom of a cold trap (vapor condenser) at the SRE fell off during 

movement and contaminated the floor. The cold trap was lifted to a vertical position when the 

bottom fell off. The top portion of the cold trap was placed in a 55-gallon plastic bag. The 

bottom portion of the cold trap was hoisted up and placed in a plastic bag inside a 55-gallon 

drum. Steel baffle plates inside the cold trap had to be pulled loose and some sodium spilled 

onto the plastic. Contamination was noted on one of the workers’ shoes. Nose smears were 

taken and ranged from background to 251 cpm. Surveys of personnel clothing revealed 

contamination ranging from 1,500 cpm to 20,000 cpm. Twenty spot smear samples were taken 

in the high bay area with a maximum count of 50,000 dpm. Additional smears were taken in 

the health physics office, change room, hallways, shop area and office area. Two smears 

came back over 50 dpm and those areas were cleaned (A0059).3

• On August 10, 1977 a water-filled storage pit used at the SRE for temporary storage of 

activated material removed from the reactor vessel during decontamination and disposition 

activities was found to be leaking. The loss of water was detected the day after explosive 

cutting was done in the reactor vessel near a vault penetration pit that had been welded 

1 EPA/HGL, Final HSA 6, p. 67, citation 3:

3 Correspondence from Remley, M. E., Atomics International, to Levy, J., U.S. Atomic Energy Commission, 

re: Contaminated Water at the SRE, December 12, 1967. 

2 EPA/HGL, Final HSA 6, p. 68, citation 1:

1 Radiation and Nuclear Safety, Rockwell International Internal Letter, re: Incident Report – Radioactive

Airborne Release at SRE, October 28, 1976. 

3 EPA/HGL, Final HSA 6, p. 68, citation 2:

2 Health and Safety, Rockwell International Internal Letter, re: Radiation Incident While Moving the Cold Trap 

into the Alcohol Passivation Tank at SRE, August 1, 1977. 

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closed. The thought was that the weld had cracked, allowing water to leak into the reactor 

cavity. Steps were taken to discontinue work that could aggravate existing cracks; monitor the 

storage pit water level frequently; and develop plans for locating the leak, storing radioactive 

components out of the pit, and pumping contaminated water from the pit. The water in the 

storage pit was contaminated to 7.2 x 10 2 μCi/cm3 gross beta activity with a mixture of Na-22, 

Co-60, Sr-90, Cs-137 and Pm-147.

Water samples were collected from existing wells and three additional monitoring wells were 

drilled. Analysis of the water samples from existing wells indicated that contamination did not 

spread outside of the immediate site, but could be in the surrounding soil, adjacent vaults, or 

to the reactor guard vessel annulus. During drilling of one of the new wells outside the east 

wall of the SRE, soil was found to be contaminated to 1,300 pCi/g. According to Rockwell, 

normal subsurface soil activity was 10 to 50 pCi/g. The water in the well was much less 

radioactive than the soil (39 μCi/cm3). Preliminary water samples taken from nine well sites 

ranged from 2.1 x 10 8 to 4.9 x 10 5 μCi/cm3 where activity was found. The highest 

contamination was found in the well with the highest soil activity. Thus, the leak appeared to 

be in the east end of the storage pit. An estimated 2,200 gallons of water was leaked from 

August 9, 1977 when the pit was filled to August 22, 1977. As a result of the loss of water, 

radiation levels in the working areas at the pit were now 20 to 50 mR/h. In October 1977, 

water samples were collected again and contamination ranged from 2.7 x 10 8 to 9.1 x 10 6 

μCi/cm3, with the highest level of contamination at the wash cell valve pit presumably a result 

of activities that agitated the sediment in the area. Two steel tanks were installed for 

containment of water in the storage pit and all water in the pit was pumped to these tanks. 

These tanks also served as the storage location for radioactive components (A0414).1

• On September 23, 1977 a release of airborne radioactivity occurred in the SRE high bay when 

a sodium passivation vessel was being unloaded by crane. No unusual occurrences or visible 

releases were noted in the operation but contamination was discovered when a health 

physicist checked his shoes. One employee had a nasal swipe with 66.5 cpm measured over 

10 minutes (Rockwell’s nasal background sample showed 20.5 cpm measured over 10 

minutes) indicating an airborne release; however, an air sampler 7 feet away did not detect 

any rise in activity. The contaminated particles were felt by Rockwell to be so large that they 

fell to the floor in the immediate area and thus were tracked out of the area on shoes (A0458 

and A0683). 2

• On June 24, 1978 an employee was carrying a stainless steel tray and tripped over a piece of 

culvert piping that protruded 4 inches above ground. The worker suffered an injury to his leg 

and was taken to West Hills Hospital for treatment. The area of the incident was smear 

surveyed for radioactive contamination. An area of the protruding piping was found to have 

1 EPA/HGL, Final HSA 6, p. 69, citations 1&2:

1 Correspondence from Tuttle, R. J., Rockwell International, to Jackson, C., Energy Research and 

Development Administration, re: Leakage of Radioactively Contaminated Water at SRE, August 23, 1977.

2 Correspondence from Tuttle, R. J., Rockwell International, to Jackson, C., Energy Research and 

Development Administration, re: Leakage of Radioactively Contaminated Water at SRE, October 25, 1977. 

2 EPA/HGL, Final HSA 6, p. 69, citation 3:

3 Owens, D. E., Rockwell International Internal Letter, re: Incident Report, SRE – September 23, 1977, 

Undated. 

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50.7 dpm of contamination. No other detectable contamination was found in the area 

(A0578).1 

5.3.1.5 Previous Radiological Investigation(s) / Decontamination & Cleanup of Releases

The HSA provides a detailed chronology of all radiological investigations and decontamination / 

cleanup of releases associated with Building 4143. 

As noted, although the 2006 Site Profile indicates that the SRE was “shut down for the last time 

in 1964,” the HSA references documentation showing that the SRE was shut down in 1967 

following the SRE PEP program. The AEC approved a plan for the deactivation of the SRE, and 

its implementation resulted in a “stored in place” configuration.2 

In October 1969, Rockwell collected a 500-cm3 sample of groundwater for radiometric analysis 

from the SRE wash cell valve. Results showed a beta concentration of 7.4 x 10 6 μCi/cm3. 

Groundwater seeped into the pit overnight and a second sample collected the next day 

indicated a beta concentration of 2.5 x 10 6 μCi/cm3. In November, 1969 Rockwell conducted a 

smeaer survey and found removable contamination of between 21 and 1,326 dpm/cm2. 

Rockwell concluded that the soil surrounding the wash cells was contaminated, probably as a 

result of the explosion that occurred in 1959 (Incident Report A00315).3 

In October 1972, Rockwell reported that several areas of the SRE, accessible to personnel, had 

radiation levels in excess of 100 mR/h. These areas constituted “high radiation areas” and were 

required to be equipped with alarms that would be automatically actuated in the event of 

personnel entry, or were required to be locked with positive control on entry. Rockwell reported 

that extensive internal contamination was present in the fuel handling machines and moderator 

handling machines. Also, there was a “significant” amount of combustible material in wooden 

and cardboard boxes, which posed a fire hazard.4

Between November 1979 and September 1982, ANL conducted a series of radiological 

measurements and analyses of the SRE. Activities included conducting instrument and smear 

surveys for radioactive contamination and collecting and analyzing material samples. During the 

survey in the summer of 1982, 46 locations with elevated radioactivity were detected, of which 

27 indicated contamination in excess of ANL’s acceptable limits. Elevated concentrations of 

1 EPA/HGL, Final HSA 6, p. 69, citation 4:

4 Owen, R. K., Rockwell International Internal Letter, re: Personal Injury Incident of June 24, 1978 at SRE, 

July 7, 1978. 

2 EPA/HGL, Final HSA 6, p. 69, citation 5:

5 Ureda, B. and Heine, W., Facilities Dismantling Plan for SRE, Atomics International Report No. 

FDP-704-990- 003, June 24, 1975, pp. 9, 13, 17-18. 

3 EPA/HGL, Final HSA 6, p. 70, citation 1:

1 North American Rockwell Corporation Internal Letter from R. K. Owen to W. F. Heine, re: Water Seepage 

Hole in the SRE Wash Cell Valve Pit, January 15, 1970. 

4 EPA/HGL, Final HSA 6, p. 70, citation 2:

2 North American Rockwell Internal Letter from W. F.eine to R. W. Hartzler, re: Use of SRE Facility for

Storage, October 30, 1972. 

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Cs-137 were detected in three samples from the storm sewer, and Sr-90, Cs-137 and U were 

detected in the samples from the sanitary sewer as it exited Building 4143. ANL concluded that 

the effluent from the outfall of the storm and sanitary sewer lines was potentially contaminated.1 

On July 23, 1985 the DOE released the SRE complex for unrestricted use and removed it from 

2 the DOE’s radiologically contaminated Surplus Facilities Program. A more detailed description 

and chronology of the Radiological Investigations of Building 4143 is provided in the HSA.

1 EPA/HGL, Final HSA 6, p. 72, citation 1:

1 Argonne National Laboratory, Post Remedial Action Survey Report for the Sodium Reactor Experiment 

(SRE) Facility, Santa Susana Field Laboratories, Rockwell International, Ventura County, California, DOEEV-0005-46, ANL-OHS/HP-84-101, February 1984, pp. iii-iv, 15. 

2 EPA/HGL, Final HSA 6, p. 72, citation 6:

6 Voigt, W. R., Jr., Certification for Unrestricted Use of the Sodium Reactor Experiment (SRE) Complex and 

the Hot Cave Facility (Building 003), U.S. Department of Energy, July 23, 1985. 

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Section 5.3.2 

Building Number:

Building Alias: 

4041

041 / T0431

Building Name: SRE Component Storage Building 

Associated Buildings:	 Building 4687 - Loading Dock

Building 4896 - Concrete Pad 

Building Function:	 Storage of Components & Radioactive Waste 

Notes: Tritium 

Radionuclides of Concern: Possible radionuclides include U-238, U-234, U-235, U- 236, 

Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Th-228, Th-232, H-3, Na-22, Na-24, Cr-51, Mn- 54, 

Fe-59, Co-60, Kr-85, Sr-89, Sr-90, Sb-125, I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)- 140, 

Nb-95, Ru-103, Ru-106, Xn-133, Xe-135, Pm-147, Sm-151. 

5.3.2.1 Description of Operations & Processes 

Building 4041 was constructed in 1958. It was the SRE Component Storage Facility. From 1959 

until 1964, the north portion of Building 4041 was used for interim storage of radioactive waste 

prior to shipment for disposal. The south portion was used for storage of controlled items. In 

May, 1962 Atomics International reported that approximately 2/3 of Building 4041 was filled with 

spare parts and radioactively contaminated components from the SRE that required cleaning 

before reuse. 1 In the early 1980’s, during decommissioning of the SRE, Building 4041 was 

initially decontaminated. In 1985 it was released for unrestricted use, after which it was used for 

non-radioactive Energy Technology Engineering Center (ETEC) equipment storage. However, in 

1982, ANL conducted a post-remedial-action survey. Four areas of Building 4041 were found to 

exceed ANL’s acceptable limits of 20 dpm / 100 cm2 for removable alpha contamination, and 

200 dpm/100 cm2 for removable beta-gamma contamination. Two areas of Pad 4678 were 

found to have activity above these limits. ANL surveyed the area again following Rockwell’s 

subsequent decontamination of the areas, and found it to be below the limits specified in the 

draft American National Standards Institute (ANSI) Standard N13.12 and NRC guidelines dated 

1982. 

The preliminary MARSSIM Classification for the Building 4041 area is Class 1, primarily due to 

the proximity with SRE reactor Building 4143 and because radioactive waste from this reactor 

was stored in Building 4041. 

5.2.2.2 Building 4041 Radiological Incident Reports 

There HSA did not find any radiological incident reports associated with Building 4041. 

1 EPA/HGL, Final HSA 6, p. 25, citation 4:

4 Atomics International letter from J. F. Trevillyan to J. V. Levy, U.S. AEC, re: Sodium Components Cleaning

Facility, May 28, 1962. 

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Section 5.3.3 

Building Number:

Building Alias: 

4686

686 / T686

Building Name: Temporary Hot Waste Storage Building 

Notes: Tritium 

Radionuclides of Concern: U-238, U-234, U-235, U-236, Pu-239, Pu-240, Pu-241, Pu-242, 

Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-59, Co-60, Kr-85, Sr-89, Sr-90, Sb-125, 

I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, Nb-95, Ru-103, Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151. 

5.3.3.1 Description of Operations & Processes 

Building 4686 was constructed in about 1957 for use as a temporary “hot” radioactive waste 

storage facility where solid radioactive waste, such as irradiated core components, moderator 

cans, and fuel elements were stored. The principal radioactive material stored in this area was 

plastic wrapped failed fuel elements containing induced radioactivity and mixed fission products. 

Aerial photographs indicate this facility was used for storage until at least 1972. 

The preliminary MARSSIM Classification for the Site 4686 area is Class 1 because of the former 

use in the SRE complex. 

5.3.3.2 Building 4686 Radiological Incident Reports 

There has been one incident associated with Building 4686 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing. Summaries of all available incident reports are 

provided in Attachment A. 

5.3.3.3 Building 4686 Radiological Incident Report Summary - Data Provided by Boeing 

*Isotopes are written as they are presented in the incident database. The HSA research team believes that MFP is an 

acronym for mixed fission products. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0396 6/17/1960 SRE Waste Area MFP Radioactive storage area was entered 

without approval or protective clothing.

• On June 17, 1960, three employees entered the SRE Temporary Waste Storage Area without 

contacting the health physics department, despite a sign on the entrance gate requesting this 

action. The employees were not wearing any protective clothing. They removed a highly 

contaminated grapple (a type of bucket) from the area and left it unattended. One of the 

employees later requested a survey of his clothing and found a spot of contamination on his 

pants. The other employees were checked, but no contamination was found. Investigation 

found that the SRE Temporary Waste Storage Area gate was usually unlocked and that 

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personnel freely moved in and out of the area without contacting the health physics 

department. There was also no specific information on whether a tagged area entry permit or 

protective clothing was required (A0396).1 

1 EPA/HGL, Final HSA 6, p. 110, citation 1:

1 Health Physics, Atomics International Internal Letter, re: Notice of Radiological or Industrial Safety Rule 

Infraction, August 12, 1960. 

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Section 5.3.4 

Building Number:

Building Alias: 

4163

163 / T163

Building Name: The Box Shop *1 

Component Equipment Repair Facility (CERF) 

Building Function: Radioactive and Chemical Materials Packaging Facility

Notes: Not all workers were monitored for radiation at this location. 

Tritium and Neptunium use / contamination. 

Radionuclides of Concern: U-238, U-234, U-235, U-236, Pu-239, Pu-240, Pu-241, Pu-242, 

Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-59, Co-60, Kr-85, Sr-89, Sr-90, Sb-125, 

I-129, I-131, Cs-134, Cs-137, Ce-141, Ba, (La)-140, Nb-95, Ru-103, Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151. EPA notes no analytical method is available for Sm-151. 

5.3.4.1 Description of Operations & Processes 

Building 4163, known as the Component Equipment Repair Facility (CERF) and Box Shop, is 

referenced in the 2006 Site Description as a location that was “directly involved” with the 

handling of radiological materials and operations at the SRE (Sodium Reactor Experiment) 

area. Building 4163 was located about 50-feet east of the SRE reactor building (4143). The 

building was rectangular in shape and built in two sections. The west side (closest to the SRE) 

comprised a steel prefabricated Butler building structure approximately 40’x40’. It was 

separated from the east side by a floor-to-ceiling sheetrock wall. 

The west side contained a change room, a 1-ton crane, sprinklers, and an exhaust system. The 

east side was a 3,200 square feet box shop, which contained a 5.5-foot deep SRE operations 

mock-up pit with concrete walls and a soil bottom. 

There are indications that during the early years of facility operations, it was determined that the 

west side of the building would be used for the repair of radiologically contaminated equipment, 

while the east side (containing a pipe shop and machine shop) would be used for constructing 

wooden shipping containers and for non-nuclear support work. This may have resulted in a 

designation of the east side of Building 4163 as “non-radiological,” a designation that may not 

have been appropriately updated as the use and function of the building changed. It should be 

noted that the wall separating the two portions of the facility was reportedly removed in the 

mid-1970’s, which would necessitate both parts of the building being considered “radiological.” 

1 The Box Shop is identified in the 2006 SSFL Site Description as a radiological facility. Available documentation 

shows use of this location for radiological processes. However, a “Crater/Packer” employed at the Box Shop received 

a response to a Radiation Exposure History Request from Boeing, wherein Boeing stated the worker was “never” 

employed at a radiological facility, never handled radioactive materials, and never exposed to radiation (and 

therefore, was never monitored). Boeing’s assertions about the employee’s job duties and work location conflict with 

the Site Profile and Employment Summary provided for the employee, who is depicted in photos alongside monitored 

workers in the Box Shop. The Box Shop provides an excellent example of the problematic discrepancies between the 

Site Profile, Boeing’s Employment Summaries, actual facility records, and actual worker records.

Case File and details provided to NIOSH / ABRWH upon request. 

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Tasks associated with Component Equipment Repair often required machines and equipment 

that were located in the Box Shop-side of the building, where the SRE operations mock-up pit 

and machine shop were located. Based on documented facility operations, the proximity of the 

building to the SRE, and incident reports involving contamination of both sides of the building, 

there is no reasonable explanation for photographs of employees of Building 4163 to show that 

some workers were monitored for radiation while others were not.

Box Shop employees built wooden boxes to ship radioactive/chemical materials. They

frequently constructed these items to “spec” based around the dimensions of the containers 

holding materials, which needed to be shipped. It was not unusual for the containers radiological 

materials to be brought to the Box Shop for packaging, or for Box Shop employees to retrieve 

used used shipping containers (that had not yet been decontaminated) from various radiological 

facilities located around SSFL, so that the containers could be reused/repurposed. In 1978 

Rockwell conducted a radiological survey at the east end of Building 4163 (Box Shop) to see 

whether contamination had migrated from the west end of the building. Rockwell did this in 

preparation for removing the wall between the two parts of Building 4163. The results showed 

contamination that was reportedly below Rockwell’s applicable limits. 

The preliminary MARSSIM Classification for the Building 4163 area is Class 1 due to its 

documented releases and its former use as an SRE support building. 

5.3.4.2 Building 4163 Radiological Incident Reports 

There have been several incidents associated with Building 4163 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

5.3.4.3 Building 4163 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0424 2/13/1960 SRE Pipe Shop ACP Employees unwrapped contaminated 

pump for repair in the clean pipe shop.

A0381 8/5/1964 SRE Area Improper handling of sodium 

components contaminated personnel 

and area.

A0542 9/15/1964 SRE Bldg. 163 MFP Drum containing primary piping from 

SRE was found smoking in CERF bldg.

A0461 12/7/1964 SRE CERF ACP Arc cutting plugged filters causing high 

airborne activity.

A0385 12/21/1964 T163 Work in the contaminated equipment 

repair facility contaminated 3 employees.

A0600 11/11966 CERF T163 Bldg. MFP Primary sodium valve cleaning resulted 

in explosion and contamination spread.

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*Isotopes are written as they are presented in the Incident Database. The EPA research team believed that “MFP” is 

an acronym for mixed fission products and “ACP” is an acronym for activation products. 

• On February 13, 1960 an employee removed a contaminated pump impeller (126 dpm/100 

cm2 beta-gamma) from its protective wrapping in a clean area of the SRE Pipe Shop. A health 

physicist questioned why this was done and the employee noted there was no other area 

available to perform the repair operation on the part. The health physicist noted that if the 

health physics department had been notified arrangements would have been made for a 

controlled work area. The impeller was rewrapped until it could be decontaminated. The 

incident report notes that the supervisor involved showed poor management and the 

employee has constantly shown a lack of respect for established radiological safety practices 

(A0424).1

• On August 5, 1964, personnel and equipment were contaminated in the Sodium Component 

Cleaning Area. An employee discovered his hands and feet were contaminated on a routine 

checkout survey. Smears of the forklift he had been driving showed 200 to 300 dpm/100 cm2 

beta gamma. Two bags of radioactive waste on the forklift were 700 dpm/100 cm2 betagamma on the outside surface. Other members of the crew were alerted to the contamination 

and brought in for monitoring. Contamination of the efour men involved ranged from 7 to 18 

mrad/h on the shoes, 2 to 6 mrad/h on the hands, and nasal smears ranging from 30 to 2,100 

dpm beta-gamma. The rest of their bodies showed nothing above background levels. The men 

were decontaminated and released, although two men were placed on restriction until 

bioassay results were received.

An extensive smear survey was then conducted in all areas and building where the men had 

been working as well as all equipment that had been used by the men. Areas found to have 

contamination levels exceeding “permissible limits” included: door handles between the SRE 

lobby and hallway (370 dpm beta-gamma), door handles in the CERF room (1,200 dpm betagamma), the ground in front of the CERF door (120 dpm beta-gamma), and the cherry picker 

seat/steering wheel (300 dpm beta-gamma), and running board (110 dpm beta-gamma). 

Additional smear surveys found maximum levels of contamination as follows: crane hook on 

steam pad (3.4 x 104 dpm beta-gamma), and floor of building containing oil-bath tank (3.6x103 

dpm beta-gamma). Both concrete pads and most of the items on the pads were contaminated 

to some level, but none to the levels described above. Smears of the asphalt between the two 

pads were below 30 dpm beta-gamma.

The incident appears to have resulted from handling of a contaminated pipe during various 

stages of the component cleaning operation. Nasal contamination may have resulted from the 

initial unwrapping of the pipe where airborne release may have occurred, or from personnel 

touching their noses with contaminated hands. The incident report suggests temporarily 

suspending operations until the exact cause of contamination is found and corrective 

1 EPA/HGL, Final HSA 6, p. 84, citation 1:

1 

Health Physics Department, Atomics International Internal Letter, re: Health Physics Notice of Rule

Infraction, February 13, 1960. 

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measures can be implemented (A0381).1

• On September 15, 1964 an employee noticed smoke coming from under the lid of a 55-gallon 

waste disposal drum at the CERF. The drum was approximately half full of pipes containing 

sodium contaminated with fission products. The fire department was called and they arrived in 

full protective gear. They opened the drum and poured a bag of calcium carbonate in to cover 

all of the sodium. The drum was closed. No contamination to personnel or equipment resulted 

from the incident (A0542).2

• On December 7, 1964 an arc welder cutting through a contaminated steel and sodium 

component had to stop because the smoke inside the glove box he was working in obscured 

his vision. Smoke then started coming out of various openings in the glove box so the 

operation was shut down and personnel were requested to leave the area. A high volume air 

sampler that was started at the beginning of the cutting operation was shut off after a 10-

minute period and indicated activity of 1.76 x 10 8 μCi/cm3. Two employees had nasal smears 

that revealed contamination of 72 and 150 dpm and a fireman responding to the incident had 

a nasal smear of 18 dpm. The men were restricted from radioactive areas until bioassay 

results were reviewed. The incident report notes that the health physics department should be 

given a copy of any written procedure for work involving radioactive material with sufficient 

time prior to the start of the job to fully evaluate hazards involved (A0461).3

• On December 21, 1964 a tagged entry permit was approved for cleanup of the CERF Building 

4163. The area had been used for the disposition of the SRE main intermediate heat 

exchange (MIHE). When a health physicist entered the work area to provide assistance, he 

noticed that an operation was being performed that was not approved in the permit. An 

employee was using a skill saw [located in the Box Shop] to cut wood used as cribbing for the 

SRE MIHE. The wood cribbing was saturated with cutting oil and material from within the heat 

exchanger. The air in the CERF was filled with sawdust and no one was wearing respirators. 

The employee was told to stop sawing, but he stated he was almost finished and that the 

health physicist could swipe his nose if he wanted. A nasal smear indicated background 

activity, but it was suspected the employee blew his nose prior to the swipe.

A 10-minute air sample indicated airborne activity at 1.83 x 10 9 μCi/cm3. Later, the whole 

work crew was surveyed for contamination and results were negative on all but three people. 

Nasal smears ranged from 120 dpm to 300 dpm beta-gamma. No clothing or skin 

contamination was detected as proper protective clothing was worn. A second 10-minute air 

sample taken after the dust settled showed a reduction in activity to 7.1 x 10 11 μCi/cm3. The 

incident occurred due to haste and lack of foresight in the cutting operation (A0385).

1 EPA/HGL, Final HSA 6, p. 85, citation 1:

1 Townsend, R. E., and C. L. Young, Atomics International Internal Letter, re: Radiological Safety Incident

Report, August 23, 1964. 

2 EPA/HGL, Final HSA 6, p. 85, citation 2:

2 Young, C. L., Atomics International Internal Letter, re: Incident Report, Building 163 (CERF), 9-15-64, 

September 17, 1964. 

3 EPA/HGL, Final HSA 6, p. 86, citation 1:

1 Galperin, A., Atomics International Internal Letter, re: Incident Report, CERF, 12-7-64, January 7, 1965. 

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• On November 1, 1966 a sodium explosion occurred in the CERF (Building 4163). The incident 

occurred during cleaning of a valve used in the SRE primary sodium service system. Although 

the valve had been soaking in Dowanol solvent for approximately 6 weeks prior to the 

cleaning operation, some unreacted sodium remained in the valve and reacted violently with 

the cleaning agent. Smear surveys of the area following the explosion indicated contamination 

levels from 726 to 57,762 dpm/100 cm2 beta-gamma. A high volume air sample obtained in 

the CERF change room indicated airborne concentrations of 2x10 9 μCi/cm3. A recount of the 

air sample 24 hours later indicated a concentration of 1.2x10 9 μCi/cm3. Filter paper from the 

air sample was analyzed for isotope identification. Results showed the presence of Cs-137, 

Na-22, Sr-90, and Y-90. The CERF room was decontaminated and SRE employees were 

reminded of the hazards involved in using solvents to react sodium (A0600).1 

1 EPA/HGL, Final HSA 6, p. 86, citation 3:

3 Unknown Author, Atomics International, re: Radiation Safety Progress for November 1966, December 9, 

1966. 

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Section 5.3.5 

Building Number:

Building Alias: 

4695 

695 / T695 

Building Name: Cold Trap Vault 

Notes: Tritium 

Radionuclides of Concern: U-238, U-234, U- 235, U-236, Pu-239, Pu-240, Pu-241, Pu-242, 

Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-59, Co-60, Kr-85, Sr-89, Sr-90, Sb-125, 

I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, Nb-95, Ru-103, Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151 

5.3.5.1 Description of Operations & Processes 

Building 4695 was constructed in about 1957 as the cold trap (vapor condenser vault). It was a 

below-ground vault for the SRE that contained a cold trap (vapor condenser) and two hot traps 

1 (heaters). Building 4695 was demolished in 1978. All below-grade structures were removed 

and the area was backfilled and paved. There were no radiological use authorizations for this 

facility, however, the DOE released the area for unrestricted use on July 23, 1985. 

The preliminary MARSSIM classification for Building 4695 is Class 1 due to its former use as an 

SRE support building. 

5.3.5.2 Building 4695 Radiological Incident Reports 

During the course of reactor operations, several primary sodium leaks and fires occurred within 

the vault. No additional information has been found about these incidents.2 

The HSA provides a detailed chronology of previous radiological investigations and 

decontamination/cleanup of releases at this location. The DOE released this location for 

unrestricted use July 23, 1985. 

1 EPA/HGL, Final HSA 6, p. 115, citation 2:

2 Owen, R. K., Radiological Survey Plan, Support of D & D Program Operations at T-143 (SRE), Rockwell 

International Report No. N704TP990008, September 15, 1981, p. 3. 

2 EPA/HGL, Final HSA 6, p. 115, citation 4:

4 Owen, R. K., Radiological Survey Plan, Support of D & D Program Operations at T-143 (SRE), Rockwell 

International Report No. N704TP990008, September 15, 1981, p. 3. 

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Section 5.3.6 

Building Number:

Building Alias: 

4653 

653 / T653 

Building Name: Interim Radioactive Waste Vault / Liquid Radioactive Waste Vault 

Notes: Tritium 

Radionuclides of Concern: The main source of radioactivity was from mixed fission products, 

and activation products. Possible radionuclides include U-238, U-234, U-235, U- 236, Pu-239, 

Pu-240, Pu-241, Pu-242, Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe- 59, Co-60, 

Kr-85, Sr-89, Sr-90, Sb-125, I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, Nb- 95, 

Ru-103, Ru-106, Xn-133, Xe-135, Pm-147, Sm-151. 

5.3.6.1 Description of Operations & Processes 

Building 4653 was constructed in about 1957 as the Interim Radioactive Waste Vault.1 Building

4653, located on the hillside above Building 4143 (SRE), consisted of several vaults one of 

which contained 10 50-gallon tanks and interconnected piping. A compressor vault contained 

compressors, controls and piping. A suction tank was buried behind and above Building 4653. 

Six decay tanks, four for gas and two for liquid, were buried in front of and below Building 4563. 

Building 4653 was connected to the SRE by piping.2 

Liquid wastes generated in the SRE were directed to one of the two 5,000-gallon liquid holdup 

tanks before disposal. Liquids were pumped from floor drains in the hot cell area, hot sinks in 

the controlled areas, and the fuel element wash station located in the high bay area of the 

reactor room. Most of the radioactivity was from mixed fission products. There were two gas 

storage systems; a low volume system for the core cover gas, and a larger volume system for 

all other areas of the SRE. Gas was pumped to one of four gas holdup tanks where short-lived 

radioactivity decayed. Activation products were the primary isotopes detected in the gas holdup 

system.3 

The preliminary MARSSIM classification for Building 4653 is Class 1 due to its former use as an 

SRE support building. 

1 EPA/HGL, Final HSA 6, p. 102, citation 1:

1 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. 

2 EPA/HGL, Final HSA 6, p. 102, citations 2 & 3:

2 Gallegos, A. N., Disposal of Radioactive Waste Systems at Bldg 653 and Bldg 143, Rockwell International 

Report No. N704-DWP-990-054, April 7, 1977, pp. 3, 9-10.

3 Owens, D. E., Radiological Survey Results – Release to Unrestricted Use, SRE Region V (Gas Storage

Vault), Rockwell International Report No. N704TI990031, November 2, 1978, pp. 3-9. 

3 EPA/HGL, Final HSA 6, p. 102, citation 3:

3 Owens, D. E., Radiological Survey Results – Release to Unrestricted Use, SRE Region V (Gas Storage

Vault), Rockwell International Report No. N704TI990031, November 2, 1978, pp. 3-9. 

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5.3.6.2 Building 4653 Radiological Incident Reports 

There has been one incident report located in association with Building 4653 that could have 

resulted in a release to the environment and worker exposure. The following table provides 

information presented in an incidents database provided by Boeing. Summaries of all available 

incident reports are provided in Attachment A. 

5.3.6.3 Building 4653 Radiological Incident Report Summary - Data Provided by Boeing 

*Isotopes are written as they are presented in the incident database. The HSA research team believes that MFP is an 

acronym for mixed fission products. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0062 11/14/1977 North of High Bay MFP Overfilled liquid transfer tank spilled 

radioactive liquid outside.

• On November 14, 1977 the 500-gallon radioactive liquid transfer tank, one of two on the 

hillside, was overfilled while performing a backflush of the SRE Culligan filter. An estimated 25 

gallons of water overflowed around the tank and to the surrounding asphalt. Activity was 

estimated at 0.12 μCi/cm3 for a total of 11 mCi. Principal isotopes were Cs-137, Sr-90, and 

Co-60. Absorbent material was placed on the spill and efforts were made to cover the areas 

with plastic and plywood to prevent further spreading of contamination in the high wind 

conditions at the time. The incident occurred because the operator was not continuously 

watching the tank sight gage and the back-flush operations proved difficult. Contaminated soil 

did not appear to have been removed at the time the HSA was conducted. (A0062). 1

The HSA provides a detailed chronology of previous radiological investigations and 

decontamination/cleanup of releases at this location. The DOE released this location for 

unrestricted use July 23, 1985. 

1 EPA/HGL, Final HSA 6, p. 103, citation 1:

1 Brengle, R.G., Rockwell International Internal Letter, re: Spill of Radioactively Contaminated Water During 

SRE Backflush Operations, November 15, 1977. 

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Section 5.3.7 

Building Number:

Building Alias: 

4654 

654 / T654 

Building Name: Interim Storage Facility (ISF)

Waste Storage for SRE, OMRE, SNAP 

Radionuclides of Concern: Uranium and mixed fission and activation products. 

5.3.7.1 Description of Operations & Processes 

Building 4654 was constructed in 1958 to support the SRE. It was originally used to store 

dummy and spent fuel elements, shipping and storage casks, and hot waste generated at the 

SRE. After the SRE stopped operating, Building 4654 was also used to store waste from the 

Organic Moderated Reactor Experiment (OMRE) and the Systems for Nuclear Auxiliary Power 

(SNAP). Building 4654 was taken out of service and did not support an active program from 

1964 through 1984. 

Building 4654 consisted of a concrete structure in the ground that anchored the topes of eight 

storage tubes. The tubes extended into large holes drilled in the bedrock, and were embedded 

with drilling mud. A paved pad was adjacent to the inground structure and provided a fenced 

storage area. 

Because low-level radiation was released from the storage of containers at this facility it was 

kept in surveillance and maintenance mode to contain the contamination until decommissioning 

began in 1984. 1 It is worth noting that Building 4654 was open to the environment providing the 

potential for surface water to seep into the storage tubes and possibly corrode the steel pipes. 

This building design increased the potential for contaminated water to leak into the subsurface 

below the storage tubes. 

A 1964 inspection tour of SSFL fuel storage areas notes that one of the SRE Core I fuel 

elements removed from the Building 4654 storage pits contained in an unwelded tube, which 

separated just as it was being fitted into the RMHF storage thimble. It was recommended that 

the tube be weld-sealed for contamination control. The inspection report notes that the building 

4654 storage area was also temporarily being used to store classified scrap materials.2

1 EPA/HGL, Final HSA 7-3, p. 184, citations 7 & 8:

7 Johnson, R.P. and Speed, D.L., Interim Storage Facility Decommissioning Final Report, ESG-DOE-13507, 

Rockwell International, Rocketdyne Division, March 15, 1985, p. 1.

8 Shah, Satish, N., Final Report Decontamination & Decommissioning of Interim Storage Facility 4654,

EID-04364, The Boeing Company, May 20, 1999, p. 12. 

2 EPA/HGL, Final HSA 7-3, p. 184, citations 7 & 8:

7 Johnson, R.P. and Speed, D.L., Interim Storage Facility Decommissioning Final Report, ESG-DOE-13507, 

Rockwell International, Rocketdyne Division, March 15, 1985, p. 1.

8 Shah, Satish, N., Final Report Decontamination & Decommissioning of Interim Storage Facility 4654,

EID-04364, The Boeing Company, May 20, 1999, p. 12. 

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The preliminary MARSSIM classification for Building 4654 is Class 1 due to the building design, 

its former use, radiological incidents, and previous investigations 

5.3.7.2 Building 4654 Radiological Incident Reports 

There has been one incident report located in association with Building 4653 that could have 

resulted in a release to the environment and worker exposure. The following table provides 

information presented in an incidents database provided by Boeing. Summaries of all available 

incident reports are provided in Attachment A. 

5.3.7.3 Building 4654 Radiological Incident Report Summary - Data Provided by Boeing 

*Isotopes are written as they are presented in the incident database. The HSA research team believes that MFP is an 

acronym for mixed fission products. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0014 1/23/1962 T654 R/A Storage MFP Loss of containment of R/A equipment 

stored outside.

A0079 7/5/1979 654 R/A Yard MFP Contamination found on pavement and 

outside storage yard fence.

• On January 23, 1962 a radiation survey of Building 4654 was conducted and revealed 

contamination had spread from the inside fenced area to the asphalt outside the fence. At that 

time, the contamination level outside the fenced area of Building 4654 was between 2 and 3 

millirads per hour (mrad/hr). The contamination on exposed equipment surfaces inside the 

fence was as high as 17 rads per hour (rad/hr). This information was written in the SRE 

operations logbook. The incident report indicated that no action was taken to decontaminate 

the area or prevent further spread of contamination. On February 13, 1962 a similar survey 

was conducted and indicated an increased spread of contamination due to rain. The 

contamination on the pavement was between 2 and 17 mrad/hr outside the enclosure. 

Exposed equipment surfaces inside the enclosure showed loss or displacement of 5 rad/hr of 

contamination as the maximum detectable level was 12 rad/hr. Four soil samples were taken 

along drainage pathways. Soil contamination ranged from 20 to 43 “uuc/gm” (presumably μCi/

g). Additional information regarding any cleanup of the area, or whether contamination 

continued to spread, was not provided in the incident summary (A0014).1

• During a preliminary survey of the storage yard on July 5, 1979, contaminated shipping casks 

stored in the area were found to be emitting high levels of radiation. These high levels were 

noted in the northeast corner of the fenced area. An area of asphalt measured greater than 

50,000 dpm with PUG-1 used. 2 A smear of a 3-gallon shipping cask found radioactive 

1 EPA/HGL, Final HSA 7-3, p. 186, citation 1:

1 

Badger, F.H., Atomics International Internal Letter, Re: Spread of Contamination in Area # 654, February

18, 1962. 

2 EPA/HGL, Final HSA 7-3, p. 186, citation 2:

2 A PUG is a radiation survey instrument made by Technical Associates, which has been developing 

radiation detection equipment since 1946. 

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contamination of 12,000 dpm and an instrument survey indicated 35 mrad/hr at the cask. 

Some of the other casks surveyed indicated radioactive intensities of 40 mrad/hr. A preliminary 

survey of the soil outside the fenced area indicated possible soil contamination along the 

northeast fence line along the ease side ranging from 200 to 400 cpm. The incident report did 

not include information to indicate any cleanup operations of the area (A0079).1

• Seals and packing on some casks and equipment stored at the ISF deteriorated from 

exposure to the elements to such an extent that low-level contamination was released. This 

release contaminated the asphalt surface near the casks and soil just outside the ISF fence. 

The casks and other sources of potential contamination were sent to an offsite disposal facility. 

Radioactive core components placed in the storage tubes contaminated the internal storage 

2 cells. The exact location where the casks and equipment were stored inside the ISF fenced 

storage area is unknown.

• During decontamination and decommissioning (D&D) excavation activities, the hydraulic 

hammer mounted on the end of the backhoe punctured storage tube #7. The area was 

surveyed for contamination, but none was found.3 

The HSA provides a detailed chronology of previous radiological investigations and 

decontamination/cleanup of releases at this location. The DOE released this location for 

unrestricted use February 1, 2005. 

1 EPA/HGL, Final HSA 7-3, p. 186, citation 3:

3 Owen, R.K., Handwritten Note Re: SRE Interim Storage Yard, July 5, 1979. 

2 EPA/HGL, Final HSA 7-3, p. 186, citation 4:

4 Johnson, R.P. and Speed, D.L., Interim Storage Facility Decommissioning Final Report, ESG-DOE-13507, 

Rockwell International, Rocketdyne Division, March 15, 1985, p. 1. 

3 EPA/HGL, Final HSA 7-3, p. 186, citation 5:

5 Johnson, R.P. and Speed, D.L., Interim Storage Facility Decommissioning Final Report, ESG-DOE-13507, 

Rockwell International, Rocketdyne Division, March 15, 1985, pgs. 9, 13. 

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Section 5.3.8 

Building Number:

Building Alias: 

4689 

689 / T689 

Building Name: Intermediate Contaminated Storage Area 

Note: Possible Tritium 

Radionuclides of Concern: U-238, U-234, U-235, U-236, Pu- 239, Pu-240, Pu-241, Pu-242, 

Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-59, Co- 60, Kr-85, Sr-89, Sr-90, Sb-125, 

I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, Nb-95, Ru- 103, Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151. 

5.3.8.1 Description of Operations & Processes 

Building 4689 was constructed in about 1959 as an intermediate storage area for contaminated 

items from the SRE. This is assumed to be an area where intermediate-level radioactive waste 

was stored. Aerial photographs indicate that this facility was used for storage until at least 1965. 

Building 4689 was removed prior to 1967. 

The preliminary MARSSIM classification of Building 4689 is Class 1 due to its former use as an 

SRE support building. 

5.3.8.2 Building 4689 Radiological Incident Reports 

The HSA research team did not located any radiological incident reports associated with this 

location. 

5.3.8.3 Radiological Investigations and Decontamination/Cleanup of Releases 

The HSA provides a detailed chronology of radiological investigations and cleanup activities at 

this location. However, this area appears to have become contaminated from the radioactive 

waste stored there. In 1978, Rockwell conducted a radiological survey of this location (then 

designated as “SRE Region V”). Survey results indicated a maximum reading of 0.08 ± 0.05 

mrad/h for surface gamma radiation, and soil samples were found to have less than 50 pCi/g of 

gross detectable beta activity. All survey results were below Rockwell’s applicable limits for 

release for unrestricted use.1 

In September 1982 ANL completed a verification survey of the SRE complex and found that it 

had been decontaminated to below limits specified in the Draft ANSI Standard No. N13.12 and 

NRC 1982 guidelines. 

1 EPA/HGL, Final HSA 6, p. 112, citation 3:

3 Owens, D. E., Radiological Survey Results – Release to Unrestricted Use, SRE Region V, Rockwell 

International Report No. N704TI990031, May 26, 1978, pp. 3-9. 

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5.4 SRE: ADDITIONAL LOCATIONS / OPERATIONS 

Several buildings and facilities associated with Sodium Graphite Reactor operations (SRE) were 

excluded from the 2006 Site Profile. This section provides a partial list and description of 

important facilities that should be included in the Site Profile. 

The majority of the following locations are classified as MARSSIM Class 1 status as a result of 

their use and proximity to other radiological buildings. Some have documented radiological 

processes, and/or documented reports of releases contained in the Incident Database. In some 

cases, support buildings were designated as “non-radiological” facilities despite their location in 

Area IV, documented use of radiological materials, incidents, and/or proximity to known sources 

of radiation exposure. Tritium has been listed among the radionuclides of concern at several 

locations. 

>> Interesting Fact: The SRE area drained directly to a ditch that transported run-off, spills, 

industrial effluent, and reactor blowdown to the R-2A Pond, located in Area II. As part of the 

Site-Wide Water Reclaim System, the ditch was constructed by AEC and Atomics International 

during the 1950’s - 60’s to keep Area IV from flooding by directing water into a system that 

drained to ponds located throughout Areas I, II and III. These ponds provided Rocketdyne 

personnel with water to fill fire extinguishers, hose down work areas, and conduct rocket engine 

testing. Each rocket engine test required, on average, up to 640,000 gallons of water. Over 

30,000 rocket engine tests were conducted at SSFL. 

On March 19, 1964 approximately 419,000 gallons of radioactively contaminated water from the 

SRE Sodium Pond was released to the Area II R-2A pond (Incident Report #A0030), where the 1

Saturn V Lunar Team conducted rocket engine tests for the Apollo missions. For a full summary 

of this incident, please see Section 5.3.1.4 Radiological Incident Report Summary for the SRE.

1 EPA/HGL, Final HSA 6, p. 63, citation 1:

1 

Badger, F. H., North American Aviation Internal Letter, re: SRE R/A Liquid Release, April 17, 1964. 

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Section 5.4.1 

Building Number:

Building Alias: 

4153 

153/T153 

Building Name/Function: SRE Service Building / Sodium Service Vault 

Notes: Tritium 

Radionuclides of Concern: Building 4153 was part of the SRE reactor complex and was 

located on the north side of Building 4143. Possible radionuclides include U-238, U-234, U-235, 

U-236, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, 

Fe-59, Co-60, Kr-85, Sr-89, Sr-90, Sb-125, I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, 

Nb-95, Ru-103, Ru-106, Xn-133, Xe-135, Pm-147, Sm-151. 

5.4.1.1 Description of Operations & Processes 

Building 4153 was constructed in approximately 1957 as a service building for the SRE. There 1

have been several radiological incident reports associated with Building 4153, which contained 

an underground vault, a 2,620-gallon secondary fill tank, a diffusion cold trap (vapor condensor) 

attached to the bottom of the secondary fill tank, an 80-gallon transfer tank, a sodium melt 

station, piping, valves, a freeze trap, a vapor trap, electrical controls, and switch gear.2 

The preliminary MARSSIM classification for the Building 4153 area is Class 1 due to its former 

use as a service building for the SRE reactor. 

5.4.1.2 Building 4153 Radiological Incident Reports 

There have been several incidents associated with Building 4153 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

5.4.1.3 Building 4153 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0340 3/12/1960 NA SERVICE 

VAULT

Sodium service vault spill cleanup 

resulted in fire in sodium container.

1 EPA/HGL, Final HSA 6, p. 77, citation 1: 

1 Ureda, B. and Heine, W, Facilities Dismantling Plan for SRE, Atomics International Report No. 

FDP-704-990- 003, June 26, 1975, pp. 76-77. 

2 EPA/HGL, Final HSA 6, p. 77, citation 2: 

2 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. 

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*Isotopes are written as they are presented in the incident database. The HSA research team believed that “MFP” is 

an acronym for mixed fission products and “ACP” is an acronym for activation products. 

A0426 4/4/1960 NA SERVICE 

VAULT

ACP SRE sodium service vault modification 

ignited tetralin during welding.

A0010 4/10/1961 NA SERVICE 

VAULT

MFP Sodium fire in sodium service vault.

A0532 1/29/1962 SRE NA Vault Employee reported injured and 

contaminated leg on the next day.

A0013 1/30/1962 SRE NA Vault MFP Employee moved contaminated item into 

clean area without HP approval.

A0533 3/6/1962 NA SERVICE 

VAULT

MFP Worker in sodium service vault without 

required permit or controls.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

• On March 12, 1960 a crew of men was going to work on a filter and cold trap (vapor 

condenser) in the Sodium Service Vault when one of the men noticed sodium leaking from a 

valve. The sodium had leaked over some of the piping and onto the floor of the vault. The fire 

department was notified and the sodium was removed from the vault by putting it in a 55-

gallon drum filled with calcium carbonate. The sodium was measured at 1,300 dpm/g and the 

radiation field the men were working in was 400 mrad/h. At approximately 10:45 a.m., a fire 

started in one of the 55-gallon drums, but it was quickly put out with an argon purge. A few 

minutes later, a fire was noticed in the sodium service vault. The vault workers donned 

contained breathing apparatus and some assisted the firemen in trying to put out the fire. 

Eventually, the vault had to be sealed and purged with argon to put the fire out. During the fire, 

a high volume air sample was taken and measured 1.64x10 10 μCi/cm3. All personnel and 

equipment in the immediate area were smeared and found free of contamination (A0340).1 

• On April 4, 1960 a fire started in the Sodium Service Vault when welding sparks ignited tetralin 

on the floor of the vault. Approximately 2 quarts of residual tetralin was on the floor when 

tetralin lines had been removed for modification. The area and personnel involved were 

monitored and no contamination was found. A high volume air sample was also taken and 

revealed no radioactive airborne contamination (A0426).2 

• On April 10, 1961 a contaminated sodium fire erupted in the Sodium Service Vault area. 

Sodium that had leaked onto the floor of the vault was being removed by being placed in 35-

gallon drums with calcium carbonate. Once all the sodium from the floor was drummed, 

sodium-covered insulation from the cold trap flange was placed in the drum. An explosion 

occurred and an employee nearby started pouring calcium carbonate on the contents of the 

drum. Another explosion occurred, accompanied by flames, and the drum was capped with a 

1 EPA/HGL, Final HSA 6, p. 78, citation 1: 

1 

Hetzler, D.K., Atomics International Internal Letter, re: Radiological Safety Incident Report, SRE Sodium

Service Vault, 3/12/60, March 30, 1960. 

2 EPA/HGL, Final HSA 6, p. 78, citation 2: 

2 

Marcotte, E.J., Atomics International Internal Letter, re: Radiological Safety Incident Report, Sodium

Service Vault SRE, 4/4/60, April 19, 1960. 

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lid. At this time, an employee in full face mask carried the drum out of the area so fire fighting 

could be effectively employed. After the smoking drum was removed another bag of calcium 

carbonate was emptied into the drum and an argon purge was started under the lid of the 

drum. Calcium carbonate was banked along the bottom of the drum to prevent any sodium 

from leaking out. The drum started to cool off and it was placed in a 55-gallon drum as 

secondary containment with additional calcium carbonate. Personnel involved received nose 

smears, but all were negative. A high volume air sample taken in the Sodium Service Vault 

showed airborne activity at 6.3 x 10 10 μCi/cm3, and smears taken in the surrounding area 

were all less than 30 dpm. The cause of the fire appeared to be the use of damp calcium 

carbonate (A0010).1 

• On January 29, 1962 an employee scratched his leg while working in the Sodium Service 

Vault. He did not report the injury or have it checked out at the time. The following day he 

became contaminated again. The front portion of the employee’s leg was contaminated to 1.3 

mrad/h. The wound itself was free from detectable contamination. The Sodium Service Vault 

was noted as a red tagged area. The employee had been handling waste from this area with a 

radiation field of 250 mrad/h before the contamination was detected (A0532).2 

• On January 30, 1962 a tagged area entry permit for work in the Sodium Service Vault was 

approved for the first shift of workers. The permit was specified to expire at 4:00 p.m. A tour of 

the SRE facility at 6:30 p.m. revealed that four men were working in the area. Two men were 

in the vault welding the hot trap (heater) and two men were packaging waste. All four men had 

initialed the entry permit after it expired. One of the employees requested the health physicist 

to monitor some radioactive waste. The waste was found to be 230 mrad/h, an abnormally 

high reading for waste from the Sodium Service Vault. A check of personnel was immediately 

conducted. One employee was found to have contamination on his hands (1.2 mrad/h), leg 

(1.3 mrad/h), shoe (1.2 mrad/h), socks (0.7 mrad/h), and nose (nasal smear 374 dpm). 

Another employee had 0.5 mrad/hr contamination on his hand because he did not have rubber 

gloves on under his canvas work gloves. Both men were successfully decontaminated. In an 

effort to pinpoint the cause of the contamination, an employee brought a flashlight into the 

Health and Safety Office, an untagged area, for a smear survey. The flashlight was found to 

be contaminated to 120 dpm beta-gamma. Eventually, a new tagged area entry permit was 

requested to complete the work in the Sodium Service Vault. Numerous radiological safety 

rules and practices were violated in this incident and rule infraction notices were issued 

(A0013).3

1 EPA/HGL, Final HSA 6, p. 79, citation 1: 

1 

Lane, W.D., Atomics International Internal Letter, re: Radiological Safety Incident Report, SRE Sodium

Service Vault, 4/10/61, April 28, 1961. 

2 EPA/HGL, Final HSA 6, p. 79, citation 2: 

2 Health and Safety Department, Atomics International Letter, re: Notice of Health and Safety Rule Infraction, 

February 7, 1962. 

3 EPA/HGL, Final HSA 6, p. 79, citation 3: 

3 Denham, R.S., Atomics International Internal Letter, re: Unusual Incident – SRE – January 30, 1962, 

February 5, 1962 

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• On March 6, 1962 an employee removed tape over a weld in the SETF high bay and NaK ran 

out onto his hand and leg. He “flipped” the NaK off his hands but his coveralls started to burn. 

He took the coveralls off and headed for the shower. The Fire Department cleaned up the 

NaK. Health and Safety monitored the area and found no detectable beta-gamma 

contamination. The employee was found free of contamination. He did not realize there was 

NaK in the system and was not dressed in proper protective gear (A0533).1 

1 EPA/HGL, Final HSA 6, p. 80, citation 1: 

1 Sessions, S.D., Atomics International Internal Letter, re: Radiological Safety Incident Report, SETF High

Bay, Bldg 24, 3/6/62, March 21, 1962. 

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Section 5.4.2 

Building Number:

Building Alias: 

4723 

723/T723 

Building Name/Function: SRE Steam Cleaning Pad 

Notes: Tritium

Drainage to Area II R-2 Ponds. 

Radionuclides of Concern: U-238, U-234, U-235, U-236, Pu- 239, Pu-240, Pu-241, Pu-242, 

Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-59, Co- 60, Kr-85, Sr-89, Sr-90, Sb-125, 

I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, Nb-95, Ru-103, Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151 

5.4.2.1 Description of Operations & Processes: 

Building 4723 was used as a steam cleaning pad for equipment and materials in the SRE. Prior 

to an incident in March, 1960 the pad was used to clean low-level radioactivity from items of 

equipment. After the incident, Atomics International decided not to use the pad for radioactive 

decontamination.1 

In 1959, a 6-foot diameter overflow pipe and a pumped sump were installed at the confluence of 

two main drainage channels upstream of the pond. Water was then pumped through a 4-inch 

diameter overland pipe to a channel connecting to the Area II ponds that serviced the Coca 

rocket engine test stand area used by Saturn V S-II employees. The overflow from these ponds 

is into Bell Canyon and thence to the Los Angeles River.2 

The preliminary MARSSIM classification for the Site 4723 area is Class 1 because of its former 

use within the SRE complex. 

5.4.2.2 Building 4723 Radiological Incident Reports 

There has been one incident associated with Building 4723 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing. Summaries of all available incident reports are 

provided in Attachment A. 

1 EPA/HGL, Final HSA 6, p. 123, citation 4: 

4 Inter-office letter from J. Borg and E. J. Marcotte to W. L. Fisher, re: Steam Clean Pad Incident, June 2, 

1960. 

2 EPA/HGL, Final HSA 6, p. 125, citation 5: 

5 Stelle, A. M., SRE Activity Requirement No. 27, D & D of Building 143 Retention Pond and Sanitary Sewer, 

Rockwell International Report No. N704ACR990024, September 14, 1981, p. 6. 

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5.4.2.3 Building 4723 Radiological Incident Report Summary - Data Provided by Boeing 

*Isotopes are written as they are presented in the incident database. The HSA research team believes that “MFP” is 

an acronym for Mixed Fission Product. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0004 3/19/1960 SRE Steam Pad MFP Steam cleaned radioactive sodium pipe 

at clean pad, contaminating area.

• On March 19, 1960 employees were decontaminating a 2-inch valve containing radioactively 

contaminated sodium at the SRE steam cleaning pad. The valve was placed in an oil bath and 

then removed for steam cleaning. The residual sodium was blasted from the valve and spread 

over the greater portion of the concrete pad. The employees were not wearing protective 

clothing and their shoes were contaminated as a result. The pad was hosed down, washing 

contaminated sodium and water onto the soil.

Two of the men had contamination on their shoes that ranged from 1.5 to 2.0 mrad/h betagamma activity. Surveys from the south, east and west edges of the pad to the center ranged 

from 0.5 to 30 mrad/h; however, the north edge of the pad to the center ranged from 30 to 100 

mrad/h. Ten representative soil samples were collected from the north edge of the pad down the 

hillside to the retention pond. Beta-gamma activity in soil ranged from 2.7 x 10 7 to 1.5 x 10 2 

μCi/g. According to Rockwell, water samples from the retention pond revealed nothing in excess 

of normal groundwater activity.

Decontamination efforts included chipping the pad surface and removing soil on the hillside. 

Sixteen barrels of concrete and 46 barrels of contaminated soil were collected for disposal. 

Follow-up soil samples found beta-gamma contamination ranging from 5.3 x 10-6 to 4.8 to 10-5 

μCi/g. Smears taken of the concrete pad showed beta-gamma activity ranging from 4 to 20 dpm 

/ 100 cm2, with the exception of one sample that was 40 dpm/100 cm2. A survey of the pad 

found readings between 0.08 and .15 mrad/h which were deemed by Rockwell as low enough to 

resurface the pad (A004).1

The HSA provides a detailed chronology of previous radiological investigations and 

decontamination/cleanup of releases. DOE released this location for unrestricted use July 23, 

1985.

1 EPA/HGL, Final HSA 6, p. 124, citation 1: 

1 Borg, J. and E.J. Marcotte, Atomics International Internal Letter, re: Steam Clean Pad Incident, June 2, 

1960. 

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Section 5.4.3 

Building Number: 4733 

Building Alias: 733/T733 

Building Name/Function: SRE Sodium Cleaning Pad 

Notes: Tritium 

Radionuclides of Concern: U-238, U-234, U-235, U-236, Pu- 239, Pu-240, Pu-241, Pu-242, 

Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-59, Co- 60, Kr-85, Sr-89, Sr-90, Sb-125, 

I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, Nb-95, Ru-103, Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151 

5.4.3.1 Description of Operations & Processes: 

Building 4733 was constructed in about 1957 and used as a sodium cleaning pad. 1 It was 

demolished in the 1970’s. 

The preliminary MARSSIM classification for the Site 4733 area is Class 1 because of its former 

use within the SRE complex. 

5.4.3.2 Building 4733 Radiological Incident Reports 

There has been one incident associated with Building 4733 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing. Summaries of all available incident reports are 

provided in Attachment A. 

5.4.3.3 Building 4733 Radiological Incident Report Summary - Data Provided by Boeing 

*Isotopes are written as they are presented in the incident database. The HSA research team believes that “MFP” is 

an acronym for Mixed Fission Product. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0282 5/12/1961 SRE North Pad MFP Steam cleaning operation contaminated 

a concrete pad, equipment and shoes.

• On May 12, 1961 parts were being steam cleaned near an open SRE high bay door. A smear 

survey of the area outside the open door indicated contamination levels as high as 1,200 dpm 

beta-gamma. The area was roped off and a meter survey of personnel was made. 

Contamination was found on the bottom of two employees’ shoes and the shoes had to be 

discarded. Additionally, 60 packages that were wrapped for storage outside the high bay doors 

were contaminated. A smear survey of the control room floor indicated 50 dpm beta-gamma. 

1 EPA/HGL, Final HSA 6, p. 129, citation 1: 

1 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. 

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The control room and areas outside the high bay door were decontaminated (A0282).1

The HSA provides a detailed chronology of previous radiological investigations and 

decontamination/cleanup of releases. DOE released this location for unrestricted use in 1985.

1 EPA/HGL, Final HSA 6, p. 129, citation 2: 

2 Galperin, A. Atomics International Internal Letter, re: Radiological Safety Incident Report, SRE, 5-12-1961, 

May 19, 1961. 

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Section 5.4.4 

Building Number:

Building Alias: 

4773 

773/T773 

Building Name/Function: SRE Sodium Pond / Waste Disposal Site 

Notes: Tritium

Area II Drainage 

Possible Radionuclides of Concern: U-238, U-234, U-235, U-236, Pu- 239, Pu-240, Pu-241, 

Pu-242, Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-59, Co- 60, Kr-85, Sr-89, Sr-90, 

Sb-125, I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, Nb-95, Ru-103, Ru-106, Xn-133, 

Xe-135, Pm-147, Sm-151 

5.4.4.1 Description of Operations & Processes: 

Site Area 4773 was constructed in about 1956 as a retention pond and dam for the SRE. 1 It 

received wastewater from the SRE complex and drained directly to Area II ponds that provided 

water for the Site-Wide Water Reclaim System, used by rocket-engine test stand personnel to 

hose down work areas and during rocket engine testing. The pond, the septic sewer system, 

and the industrial waste system were all disposal locations for waste water, and radioactive 

contamination was found in the pond.2 

It was an 800,000-gallon capacity retention pond with a compacted native earth dam. The dam 

was damaged by storm flow in 1958 and the repairs included the installation of a 1.5-foot 

diameter valved outlet pipe. A year later, as a result of complaints from downstream property 

owners, additional repairs included the installation of a 6-foot diameter overflow pipe and a 

pumped sump located at the confluence of the two main drain channels upstream of the pond. 

The sump collected all stormwater from the SRE. The pump, acting on an automatic level switch 

in the sump, pumped the water at 350 gpm through a 4-inch diameter overland pipe to a 

channel connecting to the Area II ponds serving rocketry personnel. 

When the sump filled with silt, it was abandoned in place and the pump suction inlet was 

changed to a raft floating on the SRE retention pond. In 1977, the pond was at about 1/8 

capacity (100,000 gallons); it was weed infested and algae laden. An unknown quantity of silt 

from the SRE had been deposited on the bottom. The last known cleanup was in 1958 as part 

1 EPA/HGL, Final HSA 6, p. 138, citation 1: 

1 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. 

2 EPA/HGL, Final HSA 6, p. 141, citation 1: 

1 Letter from B. D. Sujata, The Boeing Company, to J. Evans, County of Ventura, re: Information Regarding 

Permit – Septic Tank and Leach Field, October 23, 2001. 

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of the dam repair project. The 800,000-gallon capacity pond was below the 16-million gallon 

requirement for regulation and licensing by the California Department of Dam Safety.1 

The preliminary MARSSIM classification for the Site 4773 area is Class 1 because of its former 

use as the retention pond and drainage control dam for the SRE reactor. 

5.4.4.2 Building 4773 Radiological Incident Reports 

There were no official incident reports compiled for this site. However, a chronology of 

radiological incidents cited in Rockwell International reports is as follows: 

• During operations, the liquid waste storage tanks overflowed near Building 4653, the 

Intermediate Radioactive Waste Vault. The event resulted in contamination of soil in the run-off 

area. The spill was cleaned up shortly afterward.2 

• In 1964, the draining of test water from the new liquid radioactive waste tanks T2 and T3 sent 

radioactive solutions to the SRE pond and subsequently to the Area II ponds. Reportedly, the 

total release did not exceed 60 μCi. The concentration of radioactivity in the SRE retention 

pond was reportedly less than 2 pCi/cm3, and in the Area II ponds was 0.1 pCi/cm3 . C0-60 

was found to contribute about 5% of the radioactivity. 3

The HSA provides a detailed chronology of previous radiological investigations and 

decontamination/cleanup of releases. In brief, in 1979 Rockwell drained the SRE pond and 

allowed it to dry out. According to Rockwell, all areas of the pond bottom that exceeded 100 

pCi/g gross detectable beta activity were removed and disposed of as radioactive waste. AFter 

this decontamination process, Rockwell collected soil samples from the pond bottom and the 

pond was subsequently returned to service. Soil sampling resulted indicated that all samples 

were below 100 pCi/g, with a mean value of 29 pCi/g. All survey results were below Rockwell’s 

applicable limits for release for unrestricted use.4 In March, 1983 Rockwell recommended the 

1 EPA/HGL, Final HSA 6, p. 138, citation 2: 

2 Stelle, A. M., SRE Activity Requirement No. 27, D & D of Building 143 Retention Pond and Sanitary Sewer, 

Rockwell International Report No. N704ACR990024, September 14, 1981, pp. 6-7. 

2 EPA/HGL, Final HSA 6, p. 139, citation 3: 

3 Stelle, A. M., SRE Activity Requirement No. 27, D & D of Building 143 Retention Pond and Sanitary Sewer, 

Rockwell International Report No. N704ACR990024, September 14, 1981, pp. 6-7. 

3 EPA/HGL, Final HSA 6, p. 139, citation 1: 

1 Groundwater Resources Consultants, Inc., Assessment of Pond Sediments in R2, SRE and Perimeter 

Ponds at the Rockwell International Corporation Rocketdyne Division, Santa Susana Field Laboratory, 

Ventura County, California, Report No. 8640M-101, July 26, 1990, pp. 4-5. 

4 EPA/HGL, Final HSA 6, p. 139, citation 4: 

4 

Wallace, J. H., Radiological Survey results – Release to Unrestricted use, SRE Region VII, Rockwell 

International Report No. N704TI990033, May 13, 1983, pp. 7-9. 

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water being held in the SRE pond be pumped to the Rockwell Site-Wide Water Reclaim 

System, and the bottom gate valve be opened to allow for natural run-off. 1

1 EPA/HGL, Final HSA 6, p. 140, citation 2: 

2 Rockwell International internal letter from P. L. Kleinsmith to K. Johns, re: Deactivation of SRE Retention 

Pond, March 28, 1983. 

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5.5 SNAP REACTORS (DOE-NASA) 

The 2006 Site Profile provides an incomplete list of buildings, processes, and incidents 

associated with the SNAP (Systems for Nuclear Auxiliary Power) Complex. Additional SNAP

facilities and a description of operations and incidents are provided in Section 6.2. 

SNAP was a DOE-NASA project.1 Research and development of SNAP reactors was intended 

to perfect nuclear space propulsion. SNAP involved North American Aviation Atomics 

International and Rocketdyne Space & Information Division personnel, some of whom were 

associated with Area I, II or III “Time Clock Locations.” In addition, some DOE Facilities that 

operated to support SNAP were located outside Area IV, and will be addressed in a subsequent 

report. 

The SNAP Program operated from 1956 to 1971. SNAP reactors were uranium-zirconium 

hydride reactors that used fully enriched 93% uranium dispersed in fuel rods containing 

zirconium hydride. SNAP reactors that were developed and tested at SSFL: SNAP 2 SER -

Experimental Reactor, SNAP 8 S8ER - Experimental Reactor, SNAP 2 S2DR - Developmental 

Reactor, SNAP 8 S8DR - Developmental Reactor, SNAP 10-A Flight Reactors - FS1, FS2, FS3. 

5.5.1 Rockwell International Adopts Its Own “ACCEPTABLE LIMITS” for SNAP 

It should be noted that Rockwell International adopted its own “acceptable limits” for the 

radiological investigation of all buildings and facilities associated with the SNAP program, based 

on enriched uranium that was used for SNAP. According to Rockwell International, the ambient 

gamma exposure inside a 

SNAP Criticality Facility was 

actually lower than outdoor 

natural background; and 

e l e v a t e d r a t e s w e r e 

persistently attributed to 

“primordial isotopes in building 

material,” thus requiring no 

further investigation according 

to Rockwell International. 

1 Atomics International News Release, “NASA Contract to Study SNAP Reactors,” May 20, 1963. 

HDMSP001802805.pdf 

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5.5.2 Perspective: Below, an Atomics International press release announces the SNAP AECNASA contract.1 

Atomics International Inter-Office Letters 


To: All Supervision 


From: Public Relations 


Subject: News Release Date: May 20, 1963 


A news announcement on a NASA contract to study applications of 
 


SNAP reactor systems for use on the moon will be released to the 
 


press tomorrow. Essence of the story follows. 


Compact nuclear power sources for supplying electricity for men 
 


and equipment on the moon will be studied for the National 


Aeronautics and Space Administration (NASA) by Atomics 
 


International, a division of North American Aviation, Inc., 
 


according to a contract announced today. 


The contract, which is directed by NASA’s Lewis Research Center, 
 


Cleveland, Ohio, for the Office of Advanced Research and 
 


Technology, calls for a nine-month study to determine design 


modifications required to adapt SNAP reactor systems for possible 


lunar missions. SNAP, for Systems for Nuclear Auxiliary Power, is 


the Atomic Energy Commission program to develop compact, light­
 

weight, nuclear electric power systems for space applications. 


Some uses for SNAP reactor systems on the moon might include: 
 


power for communication stations that would operate unattended 
 


for long periods; power for a manned experimental laboratory or 
 


astronomical observatory, and a power source for charging the 
 


batteries of roving electrical vehicles. 


Richard E. Jespersen 


Public Relations 


1 Atomics International News Release, “NASA Contract to Study SNAP Reactors,” May 20, 1963. 

HDMSP001802805.pdf 

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5.5.3 

Building Number:

Building Alias: 

4010 

010 / T010 

Building Name: SNAP Experimental Reactor (SER)

SNAP 2 Experimental Reactor (S2ER)

SNAP 8 Experimental Reactor (S8ER) 

Building Function: Power & Endurance Tests for SNAP Reactors 

Associated Buildings: 4807 / 4808 - Electrical Equipment Pads

4809 - Air Blast Heat Exchanger Pad 

Notes: SNAP: DOE-NASA

Known source of Tritium contamination to groundwater well. 

Radionuclides of Concern: Sb-125, Am-241, Cs-134, Cs-137, Co-60, Eu-152, Eu-154, 

Pu-238, Pu-239, Pu-240, Pu-241, Sr-90, H-3, U-234, U-235, U-238. Be-10 is a potential

radionuclide of concern at Building 4010 because beryllium reflectors were used in the SNAP

reactor. 

5.5.3.1 Description of Operations & Processes 

SNAP reactors were uranium-zirconium hydride reactors that used fully enriched uranium 

dispersed in fuel rods containing zirconium hydride. 

Building 4010 was constructed in 1959 as the SNAP Experimental Reactor (SER). It was used 1

for power demonstration and endurance tests of the SER and the SNAP 2 Experimental 

Reactor (S2DR). The SER/S2ER operated in Building 4010 from September 1959 to December 

1960 at a power level of 50 kWth. After completion of the SER/S2ER operations, the reactor and 

associated test equipment were removed from the building and sent offsite for disposal. 

In 1961, improvements and modifications were made to Building 4010 to allow for testing of the 

higher power level SNAP 8 Experimental Reactor (S8ER), a prototype for the SNAP 8 reactors. 

The S8ER operated in Building 4010 from May 1963 to April 1965 at a power level of 600 kWth. 

At the conclusion of the S8ER experiment in 1965, the reactor, primary and secondary sodium 

loops, and control panels were removed from Building 4010. The primary vault, which still 

contained surface contamination, and the reactor containment vessel, which still contained 

induced radioactive materials, were covered with shielding blocks to restrict access and shield 

the remaining radiation. 

1 EPA/HGL, Final HSA 5-B, p. 13, citations 1 & 2:

1 Stelle, A.M., S8ER Facilities Decommissioning Final Report, ESG-DOE-13237, Rockwell International, 

Atomics International Division, Energy Systems Group, February 28, 1979, p. 17.

2 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. 

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An Atomics International internal letter (September 27, 1965) 1 authorizes Building 4010 

personnel to receive 600 MREM of exposure during the removal of the S8ER core vessel and 

disassembly.

Plans for facility reuse did not materialize and Building 4010 remained unused for nine years 

until being declared surplus in 1974. D&D occurred through the winter of 1977 and 1978.2 

In the late 1980’s, a large structure consisting of four cooling towers (Sodium Components Test 

Installation, or “SCTI” Power Pak Cooling Tower Building 4710) was constructed in the 

approximate location of former Building 4010. It was operated until the mid-1990’s. The cooling 

towers were removed in 2003.3 

The HSA provides a detailed chronology and description of Building 4010’s features, equipment 

and schematics. In addition, EPA included numerous employee interview excerpts related to 

operations at Building 4010. In addition, the HSA provides a chronology of radiological 

investigations concerning Building 4010, and details about its D&D. 

1978 Rockwell International radiological investigations of Building 4010 indicated that 

contamination levels were below cleanup guide limits. However, in 1979, Argonne National 

Laboratory (ANL) conducted a radiological survey for post remedial assessment of Building 

4010 to ensure that it met unrestricted release criteria. A walkover survey indicated some 

elevated readings on the asphalt, ranging from 15 to 30 μR/hr with a natural background 

ranging from 9 to 15 μR/hr. Further investigation suggested these readings were a result of 

radioactive materials stored on the hill east of the Radioactive Materials Handling Facility 

(RMHF). The Building 4010 area was eventually released for unrestricted use by DOE in 1982. 

Radiological Use Authorization No. 18 pertained to radiography in Building 4010. Radiological 

Use Authorization No. 111 pertained to the D&D of Building 4010. The HSA research team did 

not find any other Use Authorizations pertaining to this location. 

The preliminary MARSSIM classification for Building 4010 is Class 1 because of its former use. 

5.5.3.2 Building 4010 SER / S2ER / S8ER Radiological Incident Reports 

There have been several incidents associated with Building 4010 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

1 Atomics International, Internal Letter to M.L. Dina from R.M. Hill, Re: Approval for More Than 600 MREM Exposure 

During S8ER Core Vessel Removal & Disassembly,” September 27, 1965. File: HDMSp01640465.pdf 

2 EPA/HGL, Final HSA 5-B, p. 15, citation 4:

4 Stelle, A.M., S8ER Facilities Decommissioning Final Report, ESG-DOE-13237, Rockwell International, 

Atomics International Division, Energy Systems Group, February 28, 1979, pgs. 15, 27. 

3 EPA/HGL, Final HSA 5-B, p. 15, citations 5 & 6:

5 Montgomery Watson Harza, DOE Leach Fields (Area IV AOC) RCRA Facility Investigation Report, Santa 

Susana Field Laboratory, Ventura County, California, Draft, October 2003, pgs. 2-2–2-3.

6 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962-

November 1992. 

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presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

5.5.3.4 Building 4010 Radiological Incident Report Summary - Data Provided by Boeing 

*Isotopes are written as they are presented in the Incident Database. The EPA research team believed that “MFP” is 

an acronym for mixed fission products and “ACP” is an acronym for activation products. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0598 4/30/1961 SER BLDG 10 Refers to samples from April incident at 

SER.

A0277 1/1/1964 SNAP-8 CORE MFP Fission product release from cladding 

failure of SNAP 8 Reactor Fuel

A0349 10/19/1965 High Bay ACP Saw cutting of control rods resulted in 

contamination of building.

• On April 30, 1961 an unknown incident occurred in Building 4010. Incident Report A0598 

referenced the April incident when it indicated that the processing of samples from the incident 

would be delayed. The report states that three of the five samples resulting from the recent 

SER incident will be delayed due to a backlog of analyses for the gamma spectrometer. The 

three samples for which analysis was delayed were the “coolie hat smear; the smear of the 

upper grid plate, and the metallic fragments scraped from the core.” Analysis of two samples 

following the incident was completed. The first sample was composed of dust removed from 

the constant air monitor filter. The air monitor had been operating inside the reactor at the time 

of the incident. The second sample consisted of smears taken of the interior sides of the 

reactor below the core. The following radioisotopes were found in both samples: Cs-137, 

Sb-124, Co-60, Sc-46, Mn-54, Sr-90, and Y-90. The air filter dust sample had a total activity of 

2.20 x 10 1 uC/g and the smears below the core had a total activity of 1.23 x 10 1 uC (no units 

of weight were included, since this was considered a surface sample). (A0598).1

• On January 1, 1964 mixed fission product was released to the cover gas and sodium 

potassium (NaK) coolant as a result of cladding failure of SNAP 8 reactor fuel. A “substantial 

release of fission products due to cladding failure occur[red] in about 80% of the fuel rods 

during the reactor’s extended endurance run.” The uranium-zirconium fuel swelled, stretching 2

the cladding beyond its ductility limit. This caused cracks in the cladding that allowed fission 

products to diffuse into the NaK. There was no melting of the fuel or cladding and all fission 

products were, reportedly, completely contained within the reactor system. However, the 

Atomics International task force that was created to analyze the behavior of SNAP 8 fuel noted 

1 EPA/HGL, Final HSA 5-B, p. 18, citation 1:

1 Internal Correspondence from Copeland, A.A. to Carpenter G.D., Atomics International, Re: Status Report 

on Analysis of Samples Resulting from the Incident at SER in April, 1961, June 27, 1961. 

2 EPA/HGL, Final HSA 5-B, p. 18, citation 3:

3 Zwetzig, G.B., Survey of Fission- and Corrosion-Product Activity in Sodium- or NaK-Cooled Reactors, AIAEC- MEMO-12790, Atomics International, a Division of North American Rockwell Corporation, February 28, 

1969, p. 10. 

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1 that cracked cladding contained broken fuel in several of the elements. (A0277). 2

• A 2008 RFI report lists a July 5, 1961 and June 11, 1964 incident related to Building 4010. The 

details of these incidents were unclear to the EPA HSA research team. The documents 

referenced for the information in the RFI were reviewed by the research team, and did not 

suggest that radiological incidents had occurred on these two dates. A review of Boeing’s 

Radiological Incident Database does not include incident reports for these two dates. It is 

unclear if the statements in the RFI were accurate. EPA’s research team specified that 

additional information regarding possible incidents will be updated, if more is obtained.3

• In addition to the incident reports, two leaks in cooling lines are known to have occurred. One 

leak occurred in the shield cooling water lines under the vault floor and another leak was 

identified in the reactor containment vessel cooling lines in the earth near the concrete shield. 

Both leaks were repaired, and no extensive decontamination was found to be required or 

performed. However, these leaks may have promoted the release of tritium. Building 4010 has 

been identified as the source of tritium in a neighboring groundwater well.4

• On May 18, 1965 a NaK fire occurred in the reactor pit in Building 4010. The fire was 

extinguished by placing the cover on the reactor pit and flooding it with argon. No personnal 

injury or property damage was reported to be sustained.5

1 EPA/HGL, Final HSA 5-B, p. 18, citations 4 & 5:

4 Oldenkamp, R.D. and Mills, J.C., Nuclear Operations at Rockwell’s Santa Susana Field Laboratory – A 

Factual Perspective, N001ER000017, Rockwell International, December 20, 1989, p. 86. 

5 S8ER Fuel Failures, Unknown Author, October 5 1999, HDMSp01708835. 

2 EPA/HGL, Final HSA 5-B, p. 18, citation 2:

2 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. L-2. 

3 EPA/HGL, Final HSA 5-B, p. 18, citation 6:

6 Group 5 – Central Portion of Areas III and IV RCRA Facility Investigation Report, Santa Susana Field 

Laboratory, Ventura County, California, Volume VIII – RFI Site Reports, Appendix P, United States 

Department of Energy Leach Fields 2, CH2M Hill, Draft in Progress November 2008, p. Table P.2-1. 

4 EPA/HGL, Final HSA 5-B, p. 19, citation 3:

3 The Boeing Company, Rocketdyne Environmental Affairs, Building 4010 – SNAP-8 Experimental Reactor, 

February 10, 2000, pgs. 1-2. 

5 EPA/HGL, Final HSA 5-B, p. 19, citation 7:

7 Internal Correspondence from Wilmes, R. to Lang, J., Atomics International, a Division of North American 

Aviation, Inc., Reference: Monthly Progress Report for Industrial Hygiene and Safety, Santa Susana, Period 

Ended May 29, 1965, dated June 2, 1965, p. 3. 

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Section 5.5.4 

Building Number: 4024 

Building Alias: 024 / T024 

Building Name / Function: SNAP 2 Development Reactor (S2DR)

SNAP 10-A Flight Simulation Reactor (S10FS1)

SNAP 10-A Flight Simulation Reactor (S10FS3)

SNAP 10-A Flight Simulation Reactor (S10FS5) Storage

SNAP Transient Reactor (SNAPTRANS-1)

Liquid Metal Fast Breeder Reactor (LMFBR) Support

Clinch River Mock-up Reactor Operations

Storage, D&D of Hot Lab (Bldg. 4020) Equipment 

Notes: DOE/NASA

Tritium 

Radionuclides of Concern: The primary radiological constituents of concern have been 

identified as Co-60 and Eu-152 as a result of activation in the concrete and piping of the 

subterranean test vaults. Secondary radiological constituents of concern at Building 4024 

include Cs-137 and Sr-90 (fission products); H-3, Eu-154, Fe-55, Ni-59, Ni-63, Mn-54, 

potassium-40 (K-40), and sodium-22 (Na-22) (neutron activation products); Th-232, U-234, 

U-235, and U-238 (nuclear fuel material); and Am-241, Pu-238, Pu-239, Pu-240, Pu-241, and 

Pu-242 (transuranic elements).1 

5.5.4.1 Description of Operations & Processes 

Building 4024 was constructed in 1959 and used for testing SNAP reactors in a simulated space 

environment. The sub-grade test cell complex comprising Building 4024 consisted of three 

parallel cells (two power-test cells and a center transfer cell), a partial rear corridor that 

interconnected the cells, and the operating gallery. A detailed description of Building 4024’s 

features and measurements is provided in the EPA-HSA. 

The SNAP 2 Developmental Reactor (S2DR) prototype was tested at a nominal power level of 

65 kWt without any power conversion system equipment. It was tested in Vault No. 1 of Building 

4024. The SNAP 10-A Flight System 3 (S10FS3) reactor was tested in a different vault. The 

SNAP Transient (SNAPTRAN-1) reactor criticality tests occurred at Building 4024, as well as 

tasks related to the Liquid Metal Fast Breeder Reactor (LMFBR) and Clinch River Mock-Up 

reactor. During later years of site operations, Building 4024 was used for the storage and 

decontamination tasks associated with Building 4020 Hot Laboratory. 

1 EPA/HGL, Final HSA 5-A, p. 42, citations 167-169:

167 The Boeing Company, “Building 4024 Decontamination and Decommissioning Engineering Evaluation/

Cost Analysis,” May 1, 2007.

168 Tuttle, R.J., Listing of Locations in SSFL Area IV Associated With Radioactive Materials,” September 

1989.

169 The Boeing Company, Engineering Evaluation & Cost Analysis (EE/CA) Community Meeting Microsoft 

PowerPoint® Presentation, February 21, 2007. 

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Building 4024 was modified in 1962. An additional control room was added, along with a new 

room (“Vault Number 2”), which contained vacuum pumps and other heavy equipment 

associated with the SNAP 10-A Space Nuclear Reactor (S10FS-1 / S10FS-3). The vacuum 

chamber allowed for a simulation of the operational environment in outer space.1 

Power Test Vault Number 1 housed a SNAP-2 Demonstration (a.k.a. “Developmental”) Reactor 

(S2DR) that operated for 5,000 hours at 30 and 50 thermal kilowatts (kWt) between April 1961 - 

December 1962. The S2DR generated 13 MWd of power and contained 390 x 103 Ci of 

radioactivity at the end of operation. The Hot Lab (Building 4020) was used to examine the fuel 

and components from the reactor following the operation. The EPA research team could not find 

additional historical information regarding the operation of the S2DR in Building 4024. 

According to a 1962 document, Power Test Vault Number 2 was scheduled to provide shielding 

and containment for the SNAP 10-A Flight System (S10A-FS-1) in a vacuum chamber located 

within the test vault. The vacuum chamber was designed to simulate the space environment, 

and the reactor was put through various tests, including shock and vibration testing, that 

simulated space-launch. The reactor system test was scheduled to begin in late 1962 and would 

be completed and removed some time in 1963. The EPA research team did not find 

documentation indicating that the S10A-FS-1 was tested in Building 4024, but the EPA research 

team did locate documentation that the S10A-FS-3 was tested there.2 

CORE Advocacy located an Atomics International document3 that cautioned against exceeding 

specifications limits due to the rate of temperature rise. The document cautioned that the results 

of the intended testing were not exactly known and emphasized that the rate of the heat rise

would be the governing factor of a successful test, rather than the ultimate temperature 

(provided that the specification limits were not exceeded). Part of the testing involved exposing 

the reactor to thermal shock and extreme temperature fluctuations, but Atomics International 

stressed that the effect on the device could not be accurately predicted; according to Atomics 

International, the thermal shock was greatest during the first few seconds of energizing at room 

1 EPA/HGL, Final HSA 5-A, p. 25, citation 84:

84 Atomics International, Document NAA-SR-7300 Special, “Evaluation of the Atomics International Nuclear 

Development Field Laboratory as a Location for Reactor Facilities,” May 25, 1962. 

2 EPA/HGL, Final HSA 5-A, p. 26, citations 89-93: 

89 Rockwell International, Document N001ER000017, Rev. C.,Nuclear Operations at Rockwell’s Santa 

Susana Field Laboratory – A Factual Perspective, May 30, 1991.

90 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005

91 Clark, R., and Potter, G., SEC Petition Evaluation Report Summary: SEC-00093, Santa Susana Field 

Laboratory- Area IV, April 28, 2009.

92 Atomics International Document, N704FDP990006 Rev. A., “Building T024 (SETF) Facilities Dismantling 

Plan,” July 31, 1981.

93 Atomics International, Document NAA-SR-7300 Special, “Evaluation of the Atomics International Nuclear 

Development Field Laboratory as a Location for Reactor Facilities,” May 25, 1962. 

3 J. Susnir / I. Rowe, Atomics International Interoffice Letter, “Duty Cycle SNAP 10-A Actuators, Room Ambient 

Conditions,” May 20, 1963. HDMSP001906220.pdf 

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temperature conditions when full voltage to the reactor was applied. The heat (watts) generated 

was almost three times the amount that it was designed for (at 800 ̊ F). 

Atomics International warned that continuous duty for approximately one minute, just once, may 

or may not produce deleterious effects but the repetition of this would probably have a harmful 

effect, and the degree was uncertain. The author of the letter (I. Rowe, Supervisor of the 

Electro-Magnetic Devices Component Development Group in the Reactor Development

Department), stated: 

“I hesitate to give consent to any unregulated operations for an extended time and

recommend constant current control for these operations. However, if this is impossible, 

then I would reluctantly and with great reservation consent to not more than any one

minute of unregulated energizing and more important, limit this to three times during the

life of the actuator. Even so, the effects on the unit are indeterminate. A failure later on 

due to stresses caused is entirely feasible.” 

The prototype SNAP 10-A Flight System (S10FS-3) reactor operated for 10,000 hours at about 

40 kWt between January 1965 and March 1966 and set the record for the longest continuous 

operation of a nuclear reactor system with 10,000 hours of uninterrupted operation. 

On April 27, 1966 a preliminary draft radiation engineering analysis report by the Radiation 

Safety Unit of Atomics International presented information on the removal of the S10FS-3 

reactor core, coolant pump and associated piping from the building. The materials were to be 

transported to the Hot Laboratory, Building 4020. At the time of removal, the reactor was to have 

been shut down for a period of 45 days. The EPA HSA provides a detailed account of the 

sequential removal of the items but the EPA research team was unable to locate documentation 

indicating the route that may have been followed during the transfer to Building 4020, which 

may have put workers at other locations at risk of exposure.1 

On July 14, 1966, an internal letter reported that the building 4024 cells were being

decontaminated following the conclusions of the SNAP10FS-3 ground tests. At the time of the 

letter there was no immediate work planned for the cells, and the letter therefore requested the 

maximum permissible contamination levels the cells could be left in, on standby status.2 Such 

maximum permissible contamination levels should be taken into consideration when estimating 

worker dose at Building 4024, because the building remained functional and in use between 

1966 and 1976. 

A November 1976 document outlined the working procedures to be used to remove the vacuum 

vessel from the west cell of Building 4024. The document, in advance of a general radiological 

1 EPA/HGL, Final HSA 5-A, p. 27, citations 94-95: 

94 Heine, W.F., Atomics International, Preliminary Draft Radiation Engineering Analysis: Radiation Safety 

Analysis, April 27, 1966. HDMSP001852876.

95 Lords, R.E., Westinghouse Idaho Nuclear Company, Inc., SNAP and AI Fuel Summary Report, Report

WINCO- 1222, August 1994. 

2 EPA/HGL, Final HSA 5-A, p. 27, citation 96: 

. 96 Wilson R.F., Atomics International Letter Re: Building 024 – FS-3 Vault and Transfer Cell Contamination 

Levels, July 14, 1966. 

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survey of the cells, indicated that the vacuum vessel was slightly activated. While the document 

outlined the dismantling procedure, it did not indicate the date of final removal.1 

SNAP Critical Assembly 4B operated in the east cell for a short time at low power. SNAP 

Transient Test (SNAPTRAN-1) support reactor, also critical, operated in the east cell for a short 

time. Typically SNAPTRAN-1 was operated at low power, except for some pulsed operation. 

This reactor last operated in 1971, and it was the last reactor to be tested in Building 4024.2 

According to an undated Rocketdyne Environmental Affairs document, following termination of 

each project, all equipment and fuel was removed from the facility. The document does not 

indicate where all the equipment and fuel were taken upon removal; however, an August 1994 

SNAP and Atomics International Fuel Summary Report states that within a few months of 

shutdown, reactors were disassembled and select elements were examined in extensive postirradiation studies at Building 4020 (Hot Laboratory). Following examination, the fuels were sent 

to the Idaho Chemical Processing Plant (ICPP). As of 1994, the fuels remained in underwater 

storage at the ICPP. EPA’s HSA research team could not locate additional historical information 

for the termination of each project.3 

On January 30, 1973 permission was granted in an internal letter for the storage of Building 

4019’s SNAP 10A FS-5 system in Building 4024.4 A document drafted December 13, 1972 

outlined the procedures for the transfer of the reactor from Building 4019 to Building 4024. 

According to the document, the SNAP 10A FS-5 was a complete power system that was loaded 

with fuel and NaK, and acceptance tested. The fuel was identified as 93% enriched U-235 with 

a total mass of 5 kilograms of material, 4.75 kilograms compromising U-235. 

The system had been stored in Building 4019, and was to be removed and transferred to 

Building 4024 for permanent storage in 1973, via the south end of Building 4019 to the east rollup door of the high bay of Building 4024.5 The permission letter reported that a nuclear safety 

analysis (NSA) should be prepared for handling the elements outside the reactor and should be 

1 EPA/HGL, Final HSA 5-A, p. 28, citation 97:

97 Rockwell International Document, N704DWP990024, “Dismantling and Removal of 10FS3 (SNAP)

Vacuum Vessel from Building T024, Detailed Working Procedure,” November 16, 1976. 

2 EPA/HGL, Final HSA 5-A, p. 28, citation 98:

98 Atomics International Document, N704FDP990006 Rev. A., “Building T024 (SETF) Facilities Dismantling 

Plan,” July 31, 1981. 

3 EPA/HGL, Final HSA 5-A, p. 28, citations 99 & 100:

99 Rocketdyne Environmental Affairs, Building 4024 – SNAP Environmental Test Facility, Undated.

100 Lords, R.E., Westinghouse Idaho Nuclear Company, Inc., SNAP and AI Fuel Summary Report, Report

WINCO- 1222, August 1994. 

4 EPA/HGL, Final HSA 5-A, p. 28, citation 101:

101 Ketzlach, N., Internal Letter Re: New MBA for SNAP 10FS-5 Storage, January 30, 1973. 

HDMSP001840067. 

5 EPA/HGL, Final HSA 5-A, p. 28, citation 102:

102 Heneveld, W.H., NSA-652-160-001 Transfer of SNAP 10A FS-5 from Bldg 019 to Bldg 024, December 

13, 1972, HDMSP001856104. 

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reviewed by the Fuels Committee. However, information regarding the final transfer of the 

system to, and the removal of the system from, Building 4024 could not be located.1 

According to a December 1976 radiological survey plan for Building 4024, the Building 4024 test 

vaults became radioactively contaminated during the dismantling of the reactors upon

completion of operations. The plan stated that the test vaults were lined with aluminum and 

were decontaminated by site personnel to safe limits of less than 50 dpm. 

It is interesting to note that the DOE-ETEC site closure web site2 provides a photograph and 

detailed explanation of the use of aluminum vault lining for Building 4024. According to DOE, 

even with the use of the aluminum vault lining, severe activation of the concrete walls occurred 

and was detectable 15” into the concrete during later radiological surveys. However, the web 

site does not reference any aluminum lining for the test vault located at Building 4059, although 

aluminum vault lining was specifically recommended due to the unshielded, extreme testing that 

SNAP reactor systems required. A serious incident occurred during the S8DR tests at Building 

4059, and there is no indication that aluminum lining was present to contain mixed fission 

releases. DOE’s web site neither references the incident, nor the lack of aluminum shielding at 

Building 4059. More details about the S8DR Building 4059 incident are provided in the Building 

4059 section, below. 

Radiation surveys performed after decontamination of the test vaults indicated activation of the 

vault walls and equipment within the vaults, including TV camera traverse rails, vault light 

fixtures, and the environmental test vacuum vessel for S10FS3. In addition, beryllium metal 

contamination was also detected to be present during the SNAPTRAN reactor operations. 

According to the plan, the beryllium contamination resulted from the use of bare (not anodized) 

beryllium metal plates as neutron reflectors for the SNAPTRAN. The plan stated the beryllium 

2

contamination was cleaned up to less than 0.1 μg/100 cm following SNAPTRAN operations.3 

According to the 1976 plan, no significant radioactive contamination was ever detected in the 

radioactive liquid waste retention tanks; radioactive gaseous waste retention tanks contained 

atmosphere from the test vaults. Rockwell International stated in the plan that analysis of the 

gas within the tanks identified only argon-41, which has a decay half-life of 1.83 hours. Rockwell 

International indicated that no other long-lived isotopes were reported to have been detected 

during reactor operations. 4 However, it bears noting that several of the assertions made by 

Rockwell International contradict the standardized regulations of the time. (For instance, 

according to a report authored by Rockwell International, a reading of 7,500 dpm / 100 cpm2 

1 EPA/HGL, Final HSA 5-A, p. 28, citation 103:

103 Ketzlach, N., Internal Letter Re: New MBA for SNAP 10FS-5 Storage, January 30, 1973. HDMSP001840067.
 

2 http://www.etec.energy.gov/Operations/Major_Operations/SNAP.html 

3 EPA/HGL, Final HSA 5-A, p. 28, citation 104:

104 Rockwell International Document, N704TP99009, “Radiological Survey Plan, Support of D&D Program

Operations – T-024 (SNAP 2 and 10), December 9, 1976. 

4 EPA/HGL, Final HSA 5-A, p. 29, citation 105:

105 Rockwell International Document, N704TP99009, “Radiological Survey Plan, Support of D&D Program

Operations – T-024 (SNAP 2 and 10), December 9, 1976. 

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was considered “acceptable” for surface contamination of beta and gamma emitters. This report 

is discussed in detail, below). 

Following the termination of the SNAP program, Building 4024 was used intermittently on 

different Liquid Metal Fast Breeder Reactor (LMFBR) tasks. In 1987, the most recent program to 

utilize the building was the testing of a mockup of the Clinch River fuel handling systems. SSFL 1

worker records indicate that Rocketdyne employees previously thought to be confined within 

SSFL Areas I, II and III routinely performed Clinch River-related job tasks at Building 4024.2 

5.5.4.2 Radiation Surveys & Surveillance - Building 4024 

Progress reports for maintenance and surveillance of facilities in 1987 show that monthly 

maintenance and surveillance activities included radiation surveys and groundwater sampling. 

In April 1987, it was reported in the progress report that no significant removable activity was 

found during the monthly radiation survey, and that sampling of water showed to radioactivity 

above background. However, in January 1987 samples of water from the pipe chase room at 

Building 4024 showed low levels of Co-60. It is important to note the monthly report did not 

provide information to indicate what other radionuclides were included in the sampling program, 

how frequently samples were taken, or what the reportable levels were.3 In 1989, DOE’s Office 

of Environmental Audit indicated insufficient monitoring and sampling practices at SSFL.4 

Weekly surveillance and maintenance activities were conducted in 1988 and 1989. According 5

to a 1989 factual perspective of the SSFL facilities, Building 4024 contained approximately 15 

mCi of confined activation radioactivity in concrete. Because of the design of the facility, the cost 

of D&D of the facility was determined to be expensive. According to the report, “the confined 

radioactivity is decaying and will meet release criteria when the SSFL is released for 

unrestricted use.” Accordingly, the report stated Rockwell would continue surveillance of the 

facility in the interim. 

According to Boeing, between 1997-1999 Building 4024 was used as a staging and 

decontamination area for the Hot Lab (Building 4020) concrete blocks. Weekly monitoring 

measured airborne releases to the environment. Monitoring included, but was not limited to, the 

following radionuclides that were determined to possibly be present in the process of 

decontaminating the blocks: Sr-90, Cs-137, Th-230, U-234, U-235, U- 238, Pu-238, Pu-239, 

Pu-240, and Am-241. 

1 EPA/HGL, Final HSA 5-A, p. 29, citation 108: 

108 ETEC, Site Consolidation Assessment, April 16, 1987. p. 41. 

2 Worker Records provided upon request, redacted unless authorized to review. 

3 EPA/HGL, Final HSA 5-A, p. 30, citations 110 & 111:

110 Wieseneck, H.C., ETEC Monthly Progress Report-April 1987, May 20, 1987.

111 Wieseneck, H.C., ETEC Monthly Progress Report-January 1987, February 20, 1987. 

4 U.S. DOE Office of Environmental Audit, “Environmental Survey Preliminary Report, U.S. DOE Activities at SSFL,” 

1989. 

5 EPA/HGL, Final HSA 5-A, p. 30, citation 112:

112 Gaylord, G.G., SFMP Weekly Reports, 1988 through 1989. 

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As of 2011, D&D of Building 4024 was incomplete. A chronology of radiological investigations at 

Building 4024 follows: 

• A radiation survey was carried out in test cell 2 of Building 4024 on April 1, 1966, to establish 

the dose rates surrounding the S10FS-3 reactor core. The purpose of the survey was to provide 

measured dose rate data for comparison with calculated dose rate data prior to the removal of 

the test cell shield blocks. At the time of the survey, the reactor had been shut down for a period 

of 15 days. The data were obtained by taking measurements in 1-foot increments by lowering 

the detection instrument into the #3 reactor instrument thimble. The radiation dose rates various 

from 24.0 R/hr 8 feet lowered from the high bay floor, to 87.0 R/hr at the midline of the reactor 

core 15 feet lowered from the high bay floor, to 36.0 R/hr 21 feet lowered from the high bay 

floor, at the floor level of reactor test cell 2.1 

• A remote radiation survey of the transverse corridor of test cell 2 was performed on April 30, 

1966. The purpose of the survey was to evaluate the radiation levels at the entrance to the 

transverse corridor with door B-104 open. The levels at three additional points were also 

evaluated. At the four points where measurements were taken, the radiation intensity varied 

from .055 R/hr to 22.5 R/hr. The survey concluded that operations personnel could enter the 

transfer lock with door B-104 open, if necessary.2 

• On May 3, 1966, an Atomics International internal letter provided the results of a second 

radiation survey of SNAP 10FS-3, located in test cell 2 of building 4024. The survey, 

performed on April 30, 1966, measured radiation of the reactor after the reactor had been shut 

down for a period of 45 days. The data were obtained by lowering a Victoreen Radacon Model 

510 detector into the #3 reactor instrument thimble, approximately 8 feet from the axis of the 

reactor core. The results measured that, if the reactor core is assumed to be the point source, 

the maximum dose rate 1 foot from the core was approximated to be 3.07 x 103 R/hr. If the

source geometry was assumed to be a 1 foot line source when the detector is located at the 

midline of the core, the dose rate was approximated to be 2.6 x 103 R/hr at 1 foot from the 

core.3 The letter did not provide information to indicate the purpose of the survey; however, it 

should be noted that the tests on the reactor ceased in March 1966 and decontamination of 

the cells had begun by July 14, 1966.4 It should also be reiterated that maximum permissible 

1 EPA/HGL, Final HSA 5-A, p. 33, citation 128: 

.

128 Owen, R.K. Atomics International Letter Re: Radiation Survey of SNAP 10FS-3 Reactor in Test 

Cell 2, Building 024, April 12, 1966. 

2 EPA/HGL, Final HSA 5-A, p. 33, citation 129: 

.	 129 Owen, R.K., Atomics International Letter Re: Remote Radiation Survey of Transverse Corridor of Test 

Cell 2, Building 024, May 3, 1966. 

3 EPA/HGL, Final HSA 5-A, p. 33, citation 130: 

.	 130 Owen, R.K., Atomics International Letter, Re: Radiation Survey #2, SNAP 10FS-3 Reactor in Test Cell 2, 

Building 024, May 4, 1966. 

4 EPA/HGL, Final HSA 5-A, p. 33, citation 131: 

.	 131 Wilson R.F., Atomics International Letter Re: Building 024 – FS-3 Vault and Transfer Cell Contamination 

Levels, July 14, 1966. 

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contamination levels were sought, in order to leave the cells in “standby” status.1

• A February 11, 1977 internal letter from R.K. Owen to W.F. Heine of Rockwell International 

provided preliminary survey results of Building 4024, noting that the “complete radiological 

survey of building 024 appears to require more extensive effort than originally expected.” 

According to the letter, test vault 1 (S2DR and SNAPTRAN-1), test vault 2 (SNAP10FS-3), the 

vacuum equipment room, and radioactive liquid waste system were marked with a 1-meter 

square grid for a detailed smear and instrument survey. The letter did not indicate the type of 

instrument used. In test vault 1, the floor showed 55 dpm/100 cm2 in one location, while the 

remainder of the survey grids showed less than 20 dpm. The letter does not indicate whether 

the contamination was fixed or removable. A light fixture that was surveyed on the south wall 

measured 4 mR/hr. According to the letter, “one storage hold lid was lifted. The liner appears 

to be floating, suggesting that the lower part of the hole is filled with water, but this could not 

be checked at the time.” 

The EPA HSA research team did not find any additional information on the “storage hold lid” or 

its location. As a result, EPA could not (at the time the HSA was written) determine if it was in 

contact with groundwater.2 In addition, the floor of test vault 2 was randomly smear-checked 

following removal of the vacuum vessel. The letter does not provide information to indicate the 

number of smears performed; but the smears found no contamination about 20 dpm. 

Radiation levels in the vacuum equipment room were reported to be at the 1977 background 

levels. The letter did not provide 1977 background levels. The radioactive liquid waste system 

measurements comprised the “low level waste tank water,” the “suspect tank water,” and the 

“sump water.” The letter did not provide information to indicate the location of these liquid 

wastes. The results as reported in the letter are presented as follows:3 

•	 Low level waste tank water
 

-9 o alpha – 1.2 x 10 μCi/mL 

o beta – 1.7 x 10 8 

μCi/mL 

•	 Suspect tank water
 

-9 o alpha – 2.1 x 10 μCi/mL 

o beta – 1.4 x 10-8 

μCi/mL 

1 EPA/HGL, Final HSA 5-A, p. 27, citation 96: 

.	 96 Wilson R.F., Atomics International Letter Re: Building 024 – FS-3 Vault and Transfer Cell Contamination 

Levels, July 14, 1966. 

2 EPA/HGL, Final HSA 5-A, p. 33, citation 132:

132 Owen, R.K., Rockwell International, Re: Preliminary Survey Results – Building 024, February 11, 1977. 

3 EPA/HGL, Final HSA 5-A, p. 33, citations 133 & 134:

133 Owen, R.K., Rockwell International, Re: Preliminary Survey Results – Building 024, February 11, 1977.

134 The alpha and beta measurements are those reported in the letter; however, their values are very low. It 

is important to note that the micro were hand written on the document, which may have been a result of the

original document having been drafted on a type writer, or may have been added in error. 

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• Sump water

o alpha – 8.8 x 10-10 

μCi/mL o beta – 2.5 x 10 -8 

μCi/mL 

• According to an undated report titled, “Long-Range Plan for Decommissioning Surplus 

Facilities at the SSFL,” a partial decontamination project of Building 4024 was performed in 

1978. The source document does not indicate how the items were decontaminated or where 

the decontamination took place. The acceptable limits for residual radioactivity for 

surface contamination were reported by Rockwell International to be 7,500 dpm/100 

cm2 maximum beta-gamma emitters and 300 dpm/100 cm2 maximum alpha emitters. 

Accessible surfaces within the cell and support areas were cleaned to acceptable removable 

contamination limits of 100 dpm/100 cm2 beta-gamma and 20 dpm/100 cm2 alpha.1

• A November 1978 radiological survey report presents the findings of a September 1978 

survey to ensure that the facility met unrestricted release criteria. The survey included 800 

smears from the walls, floors, and remaining piping and equipment. No beta contamination in 

excess of 50 dpm/100 cm2 was reported; no alpha activity was reported for any areas of 

Building 4024 including the test vaults, transfer cell, operating gallery, high bay, and 

equipment rooms and support area. The acceptable limits for residual radioactivity for 

surface contamination were reported by Rockwell International to be 7,500 dpm/100 

cm2 maximum beta-gamma emitters and 300 dpm/100 cm2 maximum alpha emitters.

According to the report, smears were counted for alpha and beta activity on a Nuclear 

Measurements Corporation automatic counting system with an average background of 25 

cpm for beta and a counting efficiency factor of 2.35 dpm/cpm for beta. The report stated that 

alpha contamination was not suspected; however, had there been alpha contamination it 

would have been detected by the Nuclear Measurements Corporation system, and during the 

course of the survey of removable contamination no alpha activity was detected.

Rockwell International also performed a survey of surface radiation. This survey was 

conducted throughout the building and surrounding area using a beta-gamma ion chamber. 

Beta-gamma surface dose rate limits were 0.1 mrad/hr (average) and 0.5 mrad/hr (maximum). 

The maximum beta-gamma surface contamination detected outside of the power vaults was 

0.07 mrad/hr with an average background of 0.05 mrad/hr. Inside the power vaults, betagamma surface contamination was found to range from 0.5 mrad/hr to 2.5 mrad/hr. Inside the 

corridor to the power vaults, beta-gamma surface contamination ranged from 0.02 mrad/hr to 

1.8 mrad/hr. According to the report, the center floor drain in the lower level operating gallery 

was found to be contaminated during decommissioning and disposal operations. Following an 

acid cleaning, water and smear samples indicated 0.08 mrad/hr using the ion chamber and 

200 cpm using a Technical Associates Model Pug-1. The report summarized that 

contamination was “probably located in the threads at the bottom of the vertical pipe.”

Soil samples were also collected in the yard during and following the removal of the liquid and 

gas holdup tanks. According to the report, all samples were less than 30 pico curies per gram 

(pCi/g). The soil samples were counted on a Nuclear Chicago automatic counting system with 

1 EPA/HGL, Final HSA 5-A, p. 35, citation 136:

Rockwell International Document N001T1000200, Long-Range Plans for Decommissioning Surplus

Facilities at the Santa Susana Field Laboratories, Date Unknown. 

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a KCI standard, with an average background of 30 cpm and a counting efficiency factor of 

3.29 dpm/cpm. The report approximated the natural activity of uncontaminated soil in this area 

to be 20 to 30 pCi/g; however, the report did not identify or provide a reference to how these 

natural activity values were determined.

Concrete cores drilled in the power vault walls and corridors were found to have a maximum 

specific activity of 818 pCi/g. The average specific activity was 103 pCi/g. Rockwell 

International analyzed fourteen cores at 1-inch increments to a depth of between 10 and 11 

inches, for a total of 144 samples. The average specific activity was 103 pCi/g.

Rockwell International collected water samples from the drain pipe in the operating gallery, hot 

waste storage vaults, cooling system water waste holdup tanks, groundwater during the 

removal of the waste tanks, and the vacuum cleaning line in Power Vault 2. The report 

indicated all the samples were below the limit for Sr-90. However, some water was found to be 

contaminated above the limit during decommissioning and disposal work, and it was 

transferred to the RMHF for disposal. The report did not indicate the volume of water that was 

found to be contaminated above the limit during decommissioning and disposal.1 

• On March 26, 1981, additional concrete sampling in the power vaults began to determine the 

amount of concrete needed to be removed to meet 1981 unrestricted release criteria. 

Potential radiological hazards were identified as being limited to the high bay area (including 

cell complex), electrical/mechanical support, and yard areas. Two general areas of concern in 

the high bay were the cells and the S10FS-3 reactor support equipment room. Additionally, the 

electrical/mechanical support area contained other equipment and systems that were 

potentially contaminated.

As a result of exposure to neutrons escaping from the two operating reactors, the walls, 

ceiling, floor and remote handling equipment of the test cells were activated. The survey 

indicated that 12 to 22 inches of concrete would need to be removed for surface radiation to 

meet the acceptable dose rate of 0.1 mR/hr. Only two radionuclides, Co-60 and Eu-152, were 

identified as the principal contaminants found to contribute significantly to radiation greater 

than background. The report identified the following activation products as contributing to 

contamination at Building 4024: Fe-55, Mn-54, and Eu-154.2 

• According to a December 1992 safety review report regarding tritium production and release 

at SSFL, production of tritium by the operation of the S2DR, and later the S10FS3, occurred 

1 EPA/HGL, Final HSA 5-A, p. 36, citations 137 & 138: 

.	 137 Rockwell International Document N001T1000200, Long-Range Plans for Decommissioning Surplus

Facilities at the Santa Susana Field Laboratories, Date Unknown.
 138 Rockwell International Document, N704TI990044, “Radiological Survey Results – Release to

Unrestricted Use, Building 024, SSFL,” November 28, 1978. 

2 EPA/HGL, Final HSA 5-A, p. 36, citations 139 & 140: 

.	 139 Atomics International Document, N704FDP990006 Rev. A., “Building T024 (SETF) Facilities Dismantling 

Plan,” July 31, 1981. (This document was referenced in the May 2005 HSA but has not yet been obtained.)

140 Rockwell International Document N001T1000200, Long-Range Plans for Decommissioning Surplus

Facilities at the Santa Susana Field Laboratories, Date Unknown. 

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by the following ways 1:

o Neutron absorption by lithium-6 present in the concrete biological shield – The concrete 

used for the inner 2 feet of the shield differed from the concrete usually used at SSFL in that 

limestone aggregate was substituted for the normally used granitic gravel. The natural lithium 

content of this concrete was measured in 1992 to be 6.2 ppm. Using this concentration and an 

average flux of 1.2 x 108 n/cm2/sec at 37 kWt for the first 112 cm of concrete (and considering 

that this flux was uniform over all the walls, floor, and ceiling in the vault) the activity of tritium 

produced in the concrete of Vault 1 by operation of S2DR is estimated to be 0.16 Ci at 

shutdown in December 1962. In a similar manner, adjusting for power level and operating 

time, the tritium activity produced by operation of S10FS3 in Vault 2 is estimated to be 0.20 Ci 

at March 1966.

o Neutron absorption by lithium-6 present in soil surrounding the biological shield – The thick 

concrete biological shield with the added boron content absorbed nearly all the neutron that 

escaped from the reactor (less than 1 neutron per million produced by the reactor reached the 

soil), and so the production of tritium in the soil was negligible.

o Ternary fission – This was calculated to produce 0.21 Ci in the fuel of the S2DR by 

December 1962, and 0.26 Ci in S10FS3 by March 1966.

o Neutron absorption by lithium-6 present as an impurity in NaK coolant – Tritium produced in 

this manner was negligible for both reactors.

o Neutron absorption by lithium-6 in the lithium hydride shield – This activity was estimated 

from the calculation for S8DR at Building 4059, adjusting for power and time, for the S10FS3 

operation. The natural isotopic fraction of 7.42 atom percent was used to estimate a 

production of 2,200 Ci in the S10FS3 shield.

• In September 1995, ORISE conducted an independent verification survey at Building 4024. 

Surface scans were performed over 50 to 100% of accessible floors and lower walls (up to 2 

meters) for alpha, beta and gamma activity. In the fan room, elevated direct beta radiation was 

identified. While, in all other areas, alpha, beta and gamma radiation was within the 1995 

range of ambient site background. Acceptable contamination limits and gamma exposure 

rates for releasing a facility for unrestricted use are prescribed in the U.S. Department of 

Energy (DOE), the U.S. Nuclear Regulatory Commission (NRC), and the State of California 

guidelines. The lowest, most conservative limits were chosen from these guidelines and 

incorporated into the final survey criteria for Building 4024. The surface contamination limits 

for alpha and beta were excerpted from DOE Order 5400.5 and NRC Regulatory Guide 1.86 

(see Table below). The ambient gamma exposure rate limits at 1 meter were excerpted from 

an NRC Dismantling Order because at 5 microroentgens per hour (μR/hr) it was more 

1 EPA/HGL, Final HSA 5-A, p. 36, citation 141: 

.	 141 Tuttle, R.J., Rockwell International Report RI/RD92-186, Tritium Production and Release to Groundwater 

at SSFL, December 1, 1992. 

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conservative than the DOE value of 20 μR/hr, and more consistent with as low as reasonably 

achievable principles.1 

5.5.4.3

Surface Contamination Guidelines from DOE Order 5400.5 (1990) and NRC Regulatory Guide 1.86 (1974)

Allowable Total Residual Surface Contamination (dpm/100 cm2) 

Radionuclides Average Maximum Removable

Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa-231, 

Ac-227, I-125, and I-129

100 300 20

Th-Natural, Th-232, Sr-90, Ra-223, Ra-224, U-232, I-126, 

I-131, and I-133

1,000 3,000 200

U-Natural, U-235, U-238, and associated decay products 5000α 15,000α 1,000α

Beta-gamma emitters (radionuclides with decay modes 

other than alpha emission or spontaneous fission) except 

Sr-90 and others noted above

5,000β-γ 15,000β-γ 1,000β-γ

External Gamma Radiation: The average level of gamma radiation inside a building or habitable structure on a site that has 

no radiological restriction on its use shall not exceed the background level by more than 20 μR/hr. Source: U.S. Atomic 

Energy Commission (now NRC) Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors, June 1974, p. 

1.86-5 U.S. Department of Energy Order 5400.5, Radiation Protection of the Public and the Environment, February 8, 1990, p. IV-6. 

Surface activity measurements were conducted at 76 floor and wall locations. Excluding the 

power vaults, surface activity levels were less than 55 dpm/100 cm2 for alpha and ranged from 

less than 1,400 to 33,000 dpm/100 cm2 for beta. Removable alpha measured less than 12 dpm/ 

100 cm2, and removable beta measured less than 16 dpm/100 cm2. The maximum beta-gamma 

total surface activity guidelines (15,000 dpm/100 cm2) as exceeded in the hot gas compression 

room of Building 4024. Exposure rate measurements were made at four locations in Building 

4024, but none were made in the power vaults. Excluding the power vaults, exposure rates 

ranged from 11 to 13 micro roentgen per hour (μR/hr). Background was 8 μR/hr. Interior 

exposure rates satisfy DOE and NRC exposure rate guidelines. ORISE determined that existing 

documentation for Building 4024 was inadequate to support the determination that DOE 

guidelines for unrestricted release were met. Deficiencies noted included inadequate final status 

survey methods, no discussion of specific contaminants, inconsistent specification of all 

applicable guidelines and presentation of data that may be compared to the guidelines, absence 

of quantitative laboratory data, and absence of adequate figures documenting remediated areas 

and measurement and sampling locations.2 

The EPA-HSA contains more details about subsequent, chronological radiological investigations 

of Building 4024 and recommendations for the final D&D activities. 

1 EPA/HGL, Final HSA 5-A, p. 36, citation 142:

142 ORISE Report, 96/C-5, “Verification Survey of Buildings T019 and T024, Santa Susana Field Laboratory, 

Rockwell International, Ventura County, California,” February 1996. 

2 EPA/HGL, Final HSA 5-A, p. 38, citation 143: 

143 ORISE Report, 96/C-5, “Verification Survey of Buildings T019 and T024, Santa Susana Field Laboratory, 

Rockwell International, Ventura County, California,” February 1996. 

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5.5.4.4 Building 4024 Radiological Use Authorizations 

A review of a database documenting the use authorizations related to SSFL, the EPA research 

team identified two use authorizations relating to Building 4020 (Hot Lab). Authorization No. 46 

relates to the use of Cs-137 and Pu-Be sources in Building 4024, and Authorization No. 112 

relates to the “surveillance of Building 4024.” 

Authorization No. 46 was issued to Atomics International August 5, 1971 and expired on 

September 5, 1971. As referenced above, the authorization was for the use of a Cs-137 source 

and a Pu-Be neutron source in the high bay section of Building 4024. The authorized materials 

included 4.5 Ci of sealed source Cs-137, 5 Ci or sealed Pu-Be source identified as Pu-239, and 

an additional 10 mCi of sealed source Cs-137. According to the user application dated August 1

3, 1971, the purpose of the test was to observe the effect of gamma rays and neutrons on an 

infrared sensor. The application noted that there would be no contaminated materials since the 

sources were encapsulated. 

Authorization No. 112 relates to the surveillance of Building 4024 under the Surplus Facilities 

Management Program. The first available use authorization was issued on January 16, 1978 

and expired January 16, 1979. The authorization operation was listed as “decontamination and 

disposition of Building 024.” the authorized materials were listed as being, “activation product 

radioactivity” of an unknown quantity in the building’s structure and concrete.2 The authorization 

was renewed yearly; and in March 1992, the authorization was modified to include the following 

operations; surveillance and maintenance activities for contaminated areas, future 

decontamination and decommissioning activities (however not scheduled for the duration of the 

1992 annual authorization), and storage of empty radioactive materials containers. By 1994, 

Authorization No. 112F included required environmental monitoring at Building 4024. This 

monitoring included quarterly radiation surveys and quarterly contamination surveys.3 

Subsequent use authorizations after Boeing took over site operations (1996) added the storage 

and survey of contaminated waste materials from Building 4020 Hot Lab decontamination and 

decommissioning operations. The requirement was that a running inventory of Building 4020 

materials stored in Building 4024 would be maintained, but no additional controls for these 

operations were identified in the authorizations. In 1997, added operations included the 

decontamination of Building 4020 materials. 

1 EPA/HGL, Final HSA 5-A, p. 40, citation 153:

153 Baumesh, L., and Heine, W., Authorization No. 46., August 5, 1971. 

2 EPA/HGL, Final HSA 5-A, p. 40, citation 155:

155 Tuttle, R., Authorization No. 112, January 16, 1978. 

3 EPA/HGL, Final HSA 5-A, p. 41, citations 156-161:

156 Barnes, J.G., and Rutherford, P.D., Authorization No. 112D, March 18, 1992.

157 Barnes, J.G., and Rutherford, P.D., Authorization No. 112F, April 20, 1994.

158 Barnes, J.G., and Rutherford, P.D., Authorization No. 112H, January 30, 1996.

159 Barnes, J.G., and Rutherford, P.D., Authorization No. 112I, January 14, 1997.

160 Barnes, J.G., and Rutherford, P.D., Authorization No. 112I, January 14, 1997.

161 Barnes, J.G., and Rutherford, P.D., Authorization No. 112L, March 15, 1999. 

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While active decontamination of Building 4020 materials occurred in Building 4024, personnel 

were to perform weekly routine surveys of radiation and contamination levels. During periods 

when decontamination activities did not occur, personnel were instructed to perform monthly 

surveys. By 1999, use authorizations expanded to include the storage of radioactive sources 

pending transfer to a third party recycling or disposal vendor, and the authorization continued to 

list the storage and survey of contaminated waste material from Building 4020 as an “operation.” 

However, the decontamination of building blocks was expanded to include “L-85 lay down area,” 

but did not provide additional detail to describe what was meant by, “L-85 lay down area.” 

5.5.4.5 Building 4024 Open Storage Areas 

Aerial photographs over the years show open storage areas between 1967-1972, container 

leakage and open storage areas in 1980, 1983, and potentially 1985. The preliminary 

MARSSIM Classification for the building 4024 area is Class 1, due to its site operations, results 

of previous radiological investigations, and current building status. EPA’s research team 

emphasized that there were radiological incidents evidence of radiological releases; significant 

information is lacking regarding excavation activities. Additionally, EPA emphasized that past 

characterization studies for Building 4024 were focused on delineating the extent of 

contamination to standards that were applicable at the time, and EPA made additional 

recommendations for appropriate characterization of the Building 4024 area. 

5.5.4.6 Building 4024 Radiological Incident Reports 

There have been several incidents associated with Building 4024 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

5.5.4.7 Building 4024 Radiological Incident Report Summary: Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0535 3/6/1962 SETF High Bay Active 

Corrosion 

Products

Welder burned with NaK in High Bay

A0634 2/19/1970 Building 4024 

General Mills

None Identified Maintenance Workers Unknowingly 

Worked on Contaminated General Mills

A0686 4/28/1998 Building 4024 Yard None Identified Non-Contaminated Skin Rash

• On March 6, 1962 a welder began grinding on the upper weld of the “PCS” in the high bay. 

The welder stopped, removed the tape over the lower weld and NaK ran out onto his hands 

and left leg. The welder “flipped” the NaK off his hands and his coveralls started to burn. The 

welder removed the coveralls and entered the shower. The Fire Department cleaned up the 

approximately 1/4 cup of NaK. According to the incident report, health and safety personnel 

found no detectable beta and gamma contamination in the high bay or on the welder 

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(A0535).1

• On February 19, 1970, maintenance workers unknowingly worked on contaminated general 

mills (A0634). 2

• On April 28, 1998, an employee working at Building 4024 developed a skin rash on left thigh 

and both ankles. The employee believed the rash to be “beta burns” caused by beta radiation. 

A thorough investigation of the worker and the area found no radiation or contamination. The 

employee was diagnosed with poison oak at the West Hills Medical Center (A0686). 3

1 EPA/HGL, Final HSA 5-A, p.32, citation 122:

122 Sessions, S.D., “Radiological Safety Incident Report A0533, SETF High Bay, Building 24,” March 21, 

1962. 

2 EPA/HGL, Final HSA 5-A, p.32, citation 123:

123 The research team has not yet received the incident report for this incident. The incident was 

documented in Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV 

Santa Susana Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, 

p. V-3. 

3 EPA/HGL, Final HSA 5-A, p.32, citation 124:

124 Liddy, Patricia, “Incident Report File A0686, Worker Concern Over Skin Rash,” May 1, 1998. 

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Section 5.5.5 

Building Number:

Building Alias: 

4028 

028 / T028 

Building Name: Shield Test & Irradiation Reactor (STR / STIR)

Liquid Metal Fast Breeder Reactor (LMFBR) 

Building Function: Experimental Shield Test Facility 

Notes: DOE-NASA

Tritium

N16 (Neptunium) found in reactor coolant 

Radionuclides of Concern: The primary contaminant within Building 4028 is natural and 

depleted uranium (U-234, U-235, and U-238). A 1 Ci plutonium beryllium (Pu-Be) source was 1

used with initial criticality and low power experiments and N-16 was found in the reactor coolant. 

Radioactive polonium-210 (Po-210) would also be present in the coolant if a break should occur 

in the aluminum plate covering the bismuth window separating the thermal column from the 

reactor core.2 Other radionuclides of concern include isotopes of plutonium (Pu-238, Pu-239, 

Pu-240, Pu-241), americium-241 (Am-241), Cs-137, cobalt-60 (Co-60), europium isotopes 

(Eu-152, Eu-154), tritium (H-3), and Sr-90.3 

5.5.5.1 Description of Operations & Processes 

Building 4028 was a DOE-owned facility constructed in 1960 as the experimental facility for the 

Shield Test Reactor (STR), also referred to as the Systems for Nuclear Auxiliary Power (SNAP) 

Shield Test Experiment (STE) and the Shield Test Facility (STF).4 It was designed to perform 

tests on space reactor shields in order to obtain physical parameters necessary to prove the 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 223, citation 6:

6 

Vitkus, T.J., Verification Survey of the Old Conservation Yard, Building T064 Side Yard, and Building T028, Santa 

Susana Field Laboratory, Rockwell International, Ventura County, California, Oak Ridge Institute for Science and

Education, October 1993, p. 6 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 223, citation 7:

7 

Randen, et al., Hazards Summary Report The Shield Test and Irradiation Reactor Modifications for One Megawatt 

Operation, NAA-SR-9129, Atomics International, December 15, 1963, pgs. 29, 70, 93. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 223, citation 8:

8 

Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana Field 

Laboratory, Ventura County, California, Volume 1 – Methodology, May 2005, p. 2-9. 

4 EPA/HGL, Final HSA 7-3-NBZ, p. 195, citations 1-3: 

1 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962-

November 1992.

2 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. K-1.

3 Department of Energy, Real Property and Site Development Planning FY 1988-FY 1992, January 1988. 

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safety of the reactor system. The original reactor was the Shield Test Reactor (STR); research 

and experimentation began in 1961 and ended in 1964. 

Experiments consisted of exposing slabs of shielding materials to be used in the SNAP 2, 8, 

and 10-A programs to the fission spectrum from the fission plate, and measuring their 

attenuation characteristics for various energies of neutrons. The parameters studied included 

the energy and angular neutron distraction as a function of shield thickness. The decision was 

made to make modifications to increase the power from 50 kilowatts to 1 megawatt. 

The STR was modified to become the Shield Test and Irradiation Reactor (STIR); research and 

experimentation began in 1964 and continued to 1972. The STIR was used for reactor physics 

experiments, studies in radiation damage to electronic and instrumentation components, and 

neutron radiography. The STIR program was used to study neutron radiography of electronic 

explosive devices for the Saturn & Apollo space programs, including Apollo 7 - Apollo 17. 

A North American Rockwell letter (1970) describes a Reactors Committee Chairman’s visit to 

the STIR facility to view a “dry-run” in preparation for neutron radiography, which was 

anticipated to introduce “many curies of radioactivity into the facility, with attendant possibilities 

of direct radiation exposure and fission product release.” The Chairman cautioned that review of 

criticality considerations had not yet been requested by the Operating Staff and, based on his 

reading of the rough-draft outlining the procedure, he could not recommend the approval of the 

experiment. “The missing ingredient is a safety analysis.” (See Building 4059 below for a 1

detailed description of the major incident that led to this proposed “dry run”). 

In March, 1976, STIR was decommissioned and removed from Building 4028. From 1977-1981, 

Building 4028 was used to conduct research on the behavior of molten uranium, which caused 

the building to again become contaminated. 

Operations were terminated in 1984, and the building remained inactive until 1988 when clean 

out and decontamination began. Decommissioning and Demolition (D&D) began in 1988. 

The filters were to be routinely monitored and would also be changed if the dose rate at the filter 

surface approached 100 mrem/hr. According to the 1961 manual, under normal operation the 

gaseous or particulate airborne radioactive effluent released at the stack would not exceed the 

maximum permissible levels indicated by the Chicago Operations Office Atomic Energy 

Commission Manual at that time. The 1961 manual does not state what the maximum 

permissible level was at the time.2 

There are several conflicts in facility records and manuals regarding Building 4028. While 1961 

Atomics International manuals note no major radioactive waste system was associated with the 

STE, provisions were made for handling low-level liquid wastes. A 1968 STIR manual notes that 

the sink and shower in the change-room were connected to the sanitary sewer system, thus 

prohibiting discharge of radioactive materials into either. An undated facility information 

1 North American Rockwell Internal Letter, R.O. Williams, R.E. Durand, Re: Neutron Radiography of S8DR Fuel

Elements in the STIR Facility,” May 4, 1970 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 198, citation 4:

4 

Johnson, R.P. et al., SNAP Shield Test Experiment Preliminary Operations Manual, NAA-SR-5897, Atomics 

International, May 8, 1961, pgs. 6-9. 

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document notes that there was no radioactive liquid holdup tanks at Building 4028, but a 1965 

document on changes that needed to be made in the STIR operations manual states that the 

radioactive waste sink in the laboratory fume hood drained to a 20-gallon holdup tank.1 

A 1966 reactor safety survey report states that radioactive wastes generated at the STIR 

consisted of solid wastes only. According to the report, ice cream cartons containing low-level 

waste were picked up by the Radioactive Materials Handling Facility (RMHF) approximately 

once a week. Radioactive wastes of larger sizes and/or higher radiation levels were put in 55-

gallon drums, but the report notes that this material averages no more than one drum every 6 to 

9 months.2 In 1977, Building 4028 was used for uranium oxide experiments under the Liquid 

Metal Fast Breeder Reactor (LMFBR) Fuel Safety Building, or Arc Melt Facility, to investigate 

the behavior of molten uranium oxide relative to simulated reactor accidents. These tests 

continued until 1981. The former shield test vault, known as the basement or Room B101 during 

this time, housed experimental equipment such as the arc-melting vacuum furnace and 

associated appurtenances. 

5.5.5.2 Bldg. 4028 Radiological Use Authorizations 

During the course of building operations from 1960 through 1975, the use of radioactive 

materials in Building 4028 was exempt from California licensing requirements as a U.S. Energy 

Research and Development Administration (ERDA)-owned facility. After 1975, decontamination 

and disposition of the STIR, which involved low-level contamination, were authorized. In 1977, 

Use Authorization 108 was initially issued for the use and storage of UO2 for melting operations 

in Building 4028. In 1989, the retirement of Use Authorization 108 was requested as all uranium 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 199, citations 4-7: 

4 

Johnson, R.P. et al., SNAP Shield Test Experiment Preliminary Operations Manual, NAA-SR-5897, 

Atomics International, May 8, 1961, p. 41.

5 Golliher, K.G., Shield Text and Irradiation Reactor Operations Manual, NAA-SR-MEMO-12606, Atomics 

International, September 1, 1968, p. 6-12.

6 

Unknown Author, Facility Information, Building # 028, Unknown Date, BNA01247350.

7 

Unknown Author, Changes in STIR Operations Manual, October 15, 1965, BNA 02011149. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 199, citation 8:

8 Health and Safety Division, Chicago Operations Office, Reactor Safety Survey Report, Shield Test and Irradiation 

Reactor (STIR), Atomics International, February 16-17, 1966, p. 8. 

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was removed from Building 4028 and the facility had been demolished, with the exception of the 

vault area.1 

The preliminary MARSSIM classification for Building 4028 is Class 1, due to its former use as a 

nuclear reactor, radiological incidents, and previous investigations. 

5.5.5.3 Building 4028 Radiological Incident Reports 

There have been several incidents associated with Building 4028 that could have resulted in a 

release to the environment. The following table provides information presented in an incidents 

database provided by Boeing. Summaries of all available incident reports are provided in 

Attachment A. 

5.5.5.4 Building 4028 Radiological Incident Report Summary, Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

AO278 3/5/65 Irradiation Tub High Airborne Activity During Sample 

Removal After In-Core Irradiation

A0280 5/8/65 Irradiation Tub High Airborne Activity During Sample 

Removal After In-core Irradiation

A0279 6/17/65 Irradiation Tub Extremity Exposure From Handling 

Irradiated Bag Sealed With Green Tape

A0437 8/6/65 STIR Office U235 Unmarked Irradiated Fission Foil Moved 

in Private Car to Clean Office

A0055 6/2/75 RMDF & Adjacent Mixed Fission Nine Spills Have Been Identified @ 

RMDF Complex

A0065 1/10/78 Furnace Area U A Small Uranium Fire In An Arc Melting 

Furnace

A0077/

A0232

1/17/79 / 

1/30/79

Floc Tower Area / 

Pond

Mixed Fission R/A Water From Flocculation Tower 

Contaminated Drainage Ditch and Pond / 

Contaminated Crucible Stored Outside 

Exposed to Element[s]

1 EPA/HGL, Final HSA 7-3-NBZ, p. 223, citations 1-5:

1 

Radioactive Material Authorizations, Rockwell International, BNA02647573.

2 Internal Correspondence from Burgess, D.D. to Isotope Committee Chairman, Rockwell International, Re: 

Application for Renewal of Authorization for Use of Radioactive Materials No. 108C, June 8, 1981.

3 Nagel, W.E., Authorization for Use of Radioactive Materials or Radiation Producing Devices, Authorization 

Nos. 108D, 108E, 108F, 108H, 108K, Rockwell International, June 22, 1981 through June 22, 1988.

4 Internal Correspondence from Schmidt, F.G. to Moore, J.D., Rockwell International, Re: Retirement of User 

Authorization No. 108 – Uranium Melt Facility T028, September 20, 1989.

5 Kartman, A.S., Aerial Photographic Analysis of Santa Susana Field Laboratory – Area IV, Ventura County , 

California, Volume 1 & 2, U.S. Environmental Protection Agency, Office of Research and Development, 

Environmental Sciences Division, March 2010. 

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A0087 7/24/1981 Yard U Contaminated Crucible Stored Outside, 

Exposed to Elements.

A0096 1/20/82 Pond Stuck Float Switch Pumped Pond Dry 

Increasing Background to Cause Alarm

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

• On March 5, 1965 the constant air monitor located in the reactor room alarmed during the 

removal of an irradiated sample from the north irradiation facility tube referred to as a “glory 

hole.” The removal operation consisted of “hand hauling” a length of plastic coated electronic 

hoop-up wire, from which the sample was suspended, to a position where the sampled would 

be allowed to “cool.” Following the alarm, all personnel in the area were evacuated to the 

reactor control room and office. High volume air samples and smear surveys were taken from 

the reactor room and the reactor control room. The operations personnel were requested to 

seal the “glory hole” opening with plastic to prevent further airborne contamination to leak into 

the area. Based on available data contained within the incident report, the “glory hole” was 

sealed approximately an hour and a half after the alarm sounded.

The incident report summarized that surveys of personnel involved in the incident and 

indicated no contamination in excess of the background of the instrument, or 30 

disintegrations per minute (dpm) beta and gamma. There were four 10-minute high volume air 

samples taken in the reactor room, including a sample taken at the “glory hole” (Sample No. 

3). The measurements ranged from 1.0x10-7 microcuries per cubic centimeter (μC/cc) at the 

“glory hole” to 5.8 x 10-10 μC/cc, which was approximately 30 minutes after the “glory hole” 

had been sealed with plastic. A high volume air sample taken in the reactor control room 

approximately 30 minutes after the incident indicated a gross beta-gamma activity of 1.7x10-9 

μC/cc upon immediate analysis.

The results of the smear surveys obtained from the reactor room and other areas traversed by 

personnel did not indicate any contamination in excess of 30 dpm gross beta-gamma, with the 

exception of the reactor room, which ranged from 40 to 402 dpm gross beta-gamma activity.

The incident report summarized that “as indicated by the high volume air sample obtained in 

the reactor control room, the building environmental air flow patterns permit the penetration of 

radioactive airborne contamination into areas from which it should be prohibited.” It was 

recommended that a precise air flow pattern of Building 4028 be examined to correct any 

deficiencies within the building (A0278).1

• On April 8, 1965 a special bioassay request was made for an employee as a result of the 

“emission of radioactive particulates from the irradiation thimble (“glory hole”) during removal 

of sample from thimble.” The incident report did not indicate how the emission of radioactive 

particulates occurred, but indicated that the “radioelements” involved in the bioassay included 

possible “hydrocarbons plus other activation products.” According to the document, nasal 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 209, citation 1: 

1 Owen, R.K., Rough Draft Internal Letter Re: High Level Radioactive Airborne Concentrations Incident at 

Building-028 (STIR), March 5, 1965, Undated. 

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swipes of the employee indicated a gross beta-gamma activity of 1 x 103 dpm. An air sample 

was reported to have measured 5.5x10 9 μC/cc. No additional information regarding this 

incident or the employee was included in the referenced document (A0280).1

• On August 6, 1965 an unmarked fission foil containing less than .05 grams of U-235 was left 

on a desk in the office area of Building 4028 after being hand-carried from Building 4009 to 

Building 4028 in his car. The outside of the container was contaminated to a maximum of 

12,000 dpm per 100 square centimeters (dpm/100 cm2). The U-235 foil had ben previously 

irradiated at Building 4028 in April 1965. The gasket under the top did not provide a seal for 

the can because the threads of some of the screws holding the top on were stripped. The can 

was placed in safe storage and the desk was decontaminated to less than 30 dpm/100 cm2 

beta gamma. The employee who transferred the can in his car was surveyed, and his car was 

surveyed, and no significant contamination was found (A0437).2

• An October 17, 1975 internal letter discusses nine significant radioactive spill areas and 

numerous “spot areas” at the RMHF complex. The activity associated with the spills ranged 

from 0.5 to 3,500 millirads per hour (mrad/hr). The spills were assumed to have occurred 

between 1972-1975 and to have followed the course of water. The letter noted that the rainy 

season was approaching and a spill area on the bank above Building 4028 would “assuredly 

spread and create considerably more radioactive waste.” The figure [indicated in the report] 

appears to show this area to have had an activity of 210 mrad/hr adjacent to the stairway 

northwest of Building 4028 and 35 mrad/hr adjacent to the stairway at the northeast corner of 

Building 4028. Two more suspect areas, presumably around Building 4028, were to be 

surveyed as they became accessible; however, the available documents do not provide 

information on these areas (A0055).3

• A November 18, 1975 Rockwell letter follows up on the October 17, 1975 letter discussed 

above and describes the current status of the RMHF complex. It describes a contaminated 

area extending along the fence west of the RMHF complex between Building 4028 and the 

RMHF perimeter fence. This area of contamination appears to be the same area described 

above in the October 17, 1975 letter. Contamination of dirt and asphalt was thought to be the 

result of contaminated barrels stored at the west end of the RMHF complex. The barrels may 

or may not have contained contaminated material. They may have become contaminated due 

to drainage from an upstream source. Presumably, one or more of the contaminated barrels 

filled with rain water and tipped over, or by some other means, possibly corrosion, spread 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 209, citation 2: 

2 Owen, R.K., Atomics International Internal Correspondence, Subject: Special Bioassay Request, 

April 9, 1965. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 210, citation 3: 

3 Tschaeche, A.N., Atomics International Internal Correspondence, Re: Incident Report, Santa Susana Bldg.

028, August 6, 1965, August 30, 1965. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 210, citation 4: 

4 

Badger, F.H., Rockwell International Internal Letter, Re: Radioactive Spills in RMDF, October 17, 1975. 

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1 

Page 149 of 418 

contamination down the hill to the south for approximately 100 feet by several feet wide. This 

report, in its entirety, precluded the RMHF Section of the Proposed 2016 SSFL Site 

Description.

• On January 10, 1978 there was a small uranium fire at Building 4028. According to the 

incident report, the health physicist entered the test area to take a low volume air sample and 

take several spot smears on the floor and high ledges to determine if airborne activity had 

been present. The incident report states that, “when both the air sample and smear were 

attempted to be counted, several problems were encountered with the counting system in the 

area.” The report did not elaborate on the problems encountered, but indicated that alpha 

contamination of 4 to 7 counts per minute (cpm) appeared to be located on the “planchette 

slide holder.” The EPA research team has been unable to determine exactly what this 

“planchette slide holder” is, or where it was located.

Subsequent investigation found contamination of approximately 60 dpm/100 cm2 alpha on the 

lip edge of the arc-melt furnace. This area was decontaminated by personnel assigned to the 

building. The incident report stated that the constant air-monitor in the area was not operating 

at the time of the incident, but did not provide detailed information on the actual incident, 

including the duration and extent of the fire (A0065).2

A 1978 letter report provides some additional information on the fire. According to the report, 

before the pyrophoric nature of UO2 was known, “a small amount of UO2 fell out on the plastic 

covering on the floor while opening the door.” Presumably, this means the UO2 fell out of the 

furnace door and onto the floor of the former shield test vault room. The door was closed and 

the fire was extinguished by argon flow from a hose. No spread of contamination resulted. To 

prevent reoccurrence, metal trays were installed under the doors and the trays were covered 

with calcium carbonate to catch the UO2 and slow down the oxidation process.3

• On January 30, 1979 a water sample from the Building 4028 retention pond located to the 

west of the building showed an unusual increase in the concentration of radioactivity within the 

pond. The Building 4028 retention pond is also known as the RMHF 4614 Holdup Pond. 

Accordingly, this incident is addressed by EPA in detail, in the EPA Final HSA, 7-3 NBZ 

provided in its entirety with this document.

• A July 24, 1981 handwritten report indicates that radioactive material was being stored on the 

pad northwest of Building 4028. According to the report, a copper crucible for melting depleted 

uranium was stored in a wooden box, but the box lid had been removed. A survey of the upper 

sides of the crucible indicated contamination of less than 50,000 dpm. The report indicated 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 211, citation 1: 

1 Internal Correspondence from Harris, J.M. to McCurnin, W.R., Rockwell International, Re: Plan of Action 

for RMDF, November 18, 1975. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 211, citation 2: 

2 Owens, D.E., Rockwell International Internal Letter, Subject: Emergency Response-Uranium Fire at T028, 

March 28, 1978. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 211, citation 3: 

3 Gunderjahn, C.A. and Campbell, D.C., Operation of the Arc Furnace in Building 028, Santa 

Susana, Rockwell International, September 7, 1978. 

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that the box, and a similar adjoining box, was badly weathered with the plastic wrap on the 

crucible “all but gone.”

The report stated that drainage from the storage area and building 4028 was directly to Simi 

Valley. The document indicated there to be violations of existing regulations and policies and 

the author requested immediate attention. The document did not provide additional 

information on the duration of these storage activities, and additional information regarding the 

fate of the crucible and wooden box could not be located by EPA in the documents reviewed 

during the HSA (A0087). 1

• On January 20, 1982 the water monitor in the Building 4028 Pond alarmed during heavy rain. 

The Building 4028 retention pond was also known as the RMHF 4614 Holdup Pond. 

Accordingly, this incident is also discussed at length in the EPA-HSA, 7-3 NBZ, which is 

provided in its entirety with this document. 

The EPA HSA provides a detailed chronology of all radiological investigations and 

decontamination details associated with Building 4028. EPA specified that it did not agree with 

the background or maximum acceptable contamination limits that were provided in Rockwell 

International’s Final Decontamination and Radiological Survey (1991), which also appears to 

conflict with the ORISE Verification Survey conducted in 1992. 2 However, according to the 

ORISE report, Building 4028 met the release requirements for unrestricted use. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 212, citation 1: 

1 Author Illegible, Handwritten Correspondence, Subject: Storage Area Northwest of T028, July 24, 1981. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 218. 

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Section 5.5.6 

Building Number:

Building Alias: 

4059 

059 / T059 

Building Name: SNAP 2 Developmental Reactor (S8DR)

Large Leak Test Rig (LLTR)

Ground Prototype Test Facility 

Building Function: Unshielded SNAP 8 DR Prototype Endurance Testing 

Notes: MAJOR INCIDENT

DOE-NASA 

Verified Tritium Groundwater Contamination 

Radionuclides of Concern: The S8DR contained 211 fuel/moderator elements of zirconiumuranium hydride ((Zr-U)Hx) with beryllium reflector agents. Some parts of Building 4059 became 

activated by neutrons produced by the reactors. Tritium was produced and was detected in 

groundwater. Potential radioactive contaminants include U-238, U-234, U-235, Pu-238, Pu-239, 

Pu-240, Pu-241, Am-241, Cs-137, Sr-90, H-3, Fe-55, Co-58, Co-60, Ni-63, Ba- 133, Eu-152, 

Eu-154, Eu-155, Pm-147, and Ta-182. 

5.5.6.1 Description of Operations & Processes 

The Systems for Nuclear Auxiliary Power (SNAP) 8 Developmental Reactor (S8DR) was 

intended to power space missions. Building 4059 was constructed in 1961 and modified in 1964 

in order to create a simulated space environment for ground-testing the S8DR prototype. The 

S8DR was unshielded, and due to its design objectives, it would require extensive testing and 

experimentation to prove its longevity and resilience in orbital spaceflight. One of the 

requirements was a 12,000-hour test run. 

According to the 2006 SSFL Site Profile, “No incidents of any consequence” occurred involving 

the S8DR, other than cracked cladding that proved to be a minor issue, and all fission products 

were contained in the reactor system. However, the S8DR was an unshielded, experimental 

reactor that was operated for a year, despite signs of failing fuel.1 EPA describes the event at 

Building 4059 as one of the worst nuclear incidents at SSFL, second or third in comparison only 

to the SRE event in 1959. This incident is described in detail below. 

Building 4059 contained a below-ground test vault, a pipe chase room, vacuum equipment 

room, cement vault walls (which contained activation products), and a contaminated liquid 

holdup system. There is no indication that Building 4059 used the recommended aluminum 

vault lining that was implemented in the construction of Building 4024’s test vaults. The pipe 

chase room (adjacent to the vault) was sand-filled and contained the vessel shield cooling pipes 

and a line from the vacuum chamber to the vacuum equipment room. In 1964, a French drain 

was added around three external sides below the foundations, to collect and remove any 

1 Atomics International, “Findings of the SNAP 8 Developmental Reactor (S8DR) Post-Test Examination,” June 30, 

1971 by A.H. Lillie and V.L. Rooney, Jr. 

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infiltrating groundwater.1 Gas produced from the reactor was pumped to gas holdup tanks and 

after a delay for the needed decay of radioactive argon and zenon, discharged to the 

atmosphere through the facility stack. The HSA provides a detailed description of Building 

4059’s features, equipment, and schematics. 

5.5.6.2 Building 4059 S8DR Serious Incident 

In August 1964, Bechtel was contracted to provide the AEC with a conceptual report, detailing 

the types of modifications Building 4059 would need to be capable of handling 10,000+ hour 

endurance tests of an unshielded, experimental reactor in a simulated space environment. The 

S8DR would need to endure extreme operating conditions associated with orbital spaceflight, 

and to provide a 600 kwt power output, 1300 ̊F NaK coolant outlet temperature, and 12,000 

hours of continuous operation with high reliability. Bechtel’s report provided specific 

recommendations that would be required to modify Building 4059, and to improve both the 

capability and quality of reactor testing, while ensuring worker safety.2 

Bechtel acknowledged the risks of such a degree of experimentation with an unshielded reactor. 

The containment of mixed fission products in the event of an accident was a primary concern. 

Bechtel’s answer was to ensure that the facility itself was properly shielded, and recommended 

the use of a special aluminum alloy to line the walls of the test vault. Bechtel reasoned that the 

use of special aluminum alloy lining on the walls of the test vault would limit the activation of 

cement vault walls; ensure containment of mixed fission products in the event of an accidental 

release; facilitate ease of decontamination between tests; and permit personnel to safely reenter 

the vault 45 days after a 10,000+ hour test run. Bechtel calculated that a worker’s expected 

exposure under such tightly controlled conditions and flawless operations would be 

approximately 2.5 mrem/h, 45 days after the reactor was shut down and removed from the test 

vault. 

“The facility will be shielded to reduce the total neutron and gamma ray dose rate to a

level that will permit occupancy of areas adjacent to those containing radiation sources.

Occupancy will be based on shifts of 8 hours per day, 5 days per week. Special 

materials will be used in the test vault to reduce activation of the test vault walls, thus

permitting safe re-entry (2.5 mrem/h) into the test vault 45 days after shutdown and

removal of the reactor.”3 

The EPA HSA research team provided a detailed description of Building 4059’s construction 

materials. There were no references to the presence or use of aluminum alloy lining the walls of 

the test vault. EPA indicated that the cement vault walls were highly activated with mixed fission 

products. Review of construction details and documentation related to Building 4059 suggest 

1 EPA/HGL, Final HSA 5C p. 15, citation 1: 

1 Vitkus, T. J. and Morton, J. R., Radiological Survey of the Building 059 Reactor Vault, Santa Susana Field 

Laboratory, Rockwell International, Ventura County, California, Final Report, Oak Ridge Institute for Science 

and Education ORISE 95/G-18, June 1995, pp. 1-2, 8-9. 

2 Bechtel Corporation for the U.S. AEC, “Title 1 Report: Preliminary Services for SNAP 8 Flight Prototype Test Facility, 

Building 056 [sic], Santa Susana, California,” August 1964. File: 4609929.pdf 

3 Ibid. 122119551 

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that Bechtel’s recommended modifications, which included the addition of aluminum vault lining 

to improve safe reactor operations, where never implemented. 

The DOE-ETEC site closure web site1 describes the use of aluminum vault lining for Building 

4024, and a photograph is provided to illustrate the vault walls lined with aluminum alloy. 

However, no such information is provided for Building 4059 and there is no photograph of the 

vault. At this point in time, CORE Advocacy has found no references to the documented use of 

aluminum alloy lining of the Building 4059 reactor vault, to ensure safer operations of the 

unshielded, experimental S8DR during its 12,000 hour endurance test, the majority of which 

was conducted with clear indications of fuel failure. 

In addition, while the Bechtel modification guide specifies a waiting period of 45-days before 

employees should enter the vault following a reactor test, incident reports and other documents 

associated with Building 4059 and S8DR operations indicate that personnel were allowed to 

enter the test vault immediately following reactor tests. There is no indication that workers were 

required to wait for a specified time period, before entering the vault. 

A 12,000-hour endurance test of the S8DR began in January, 1969. According to Atomics 

International’s “post-mortem” examination of S8DR fuel, soon after the test began there were 

clear indications of fuel element cladding ruptures. 2 However, despite the indications of 

problems, the reactor continued to operate and by July, the report indicates that it had become 

apparent that a number of the fuel elements had ruptured inside the reactor. The coolant 

temperature was reduced by 100 ̊F and the test continued. The endurance test was terminated 

after 7,000 hours in December, 1969. 

It is currently unclear what date the S8DR was removed from Building 4059. However, 

according to a May 4, 1970 Atomics International internal memo, the S8DR had been removed 3

from Building 4059 and set-up to function at the STIR Facility (Building 4028) by April 30, 1970. 

It was prepared for a “dry run” in preparation for neutron radiography of 211 S8DR fuel 

elements. The Chairman of the Reactor Committee, R.E. Durand, specified that the “dry run” of 

the S8DR “would introduce many curies of radioactivity into the facility, with attendant 

possibilities of direct radiation exposure and fission product release.” In addition, the process of 

radiography posed its own risk, with employees present during the operation. Mr. Durand 

refused to approve the dry-run of the S8DR because of the clear risks involved, and because 

the process had not been adequately evaluated for safety. 

It is currently unclear what date the S8DR was removed from the STIR Facility (Building 4028) 

and transported to the Hot Laboratory (Building 4020), where the cause of the fuel element 

cladding ruptures would be determined. However, once at the Hot Lab, Atomics International 

determined that 72 fuel elements in the core had cracked cladding tubes typical of stress 

caused by temperatures in excess of the reactor’s design capability. In addition, it was 

1 http://www.etec.energy.gov/Operations/Major Operations/SNAP.html 

2 Atomics International, “Findings of the SNAP 8 Developmental Reactor Post-Test Examination,” by A.H. Lillie and 

V.L. Rooney, Jr. June, 1971 for the AEC, Contract AT(04-3)-701 

3 Atomics International, Internal Letter, “Neutron Radiography of S8DR Fuel Elements in STIR Facility,” R.E. Durand 

to R.O. Williams, May 4, 1970. File: HDMSP01715643.pdf 

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determined that the damage had been incurred “relatively early in life,” from element-to-element 

loadings that resulted from bowing and clustering of the fuel elements.1 

It is unlikely that mixed fission products would have been contained in the S8DR during fuel 

failure, given the lack of shielding of the reactor, and the lack of aluminum alloy shielding at the 

facility (which had been recommended by Bechtel). S8DR’s operation for 7,000 with failing fuel 

and inadequate reactor or building shielding should be considered when evaluating 

environmental data for SSFL workers, and when assuming probable worker exposure as a 

result of Building 4059 operations. 

5.5.6.3 Building 4059 Tritium Contamination 

In 1983, during a routine facility inspection, Rockwell discovered groundwater leaking into the 

south test cell of the reactor vault and becoming contaminated. Radioactivity measurements in 

the water were less than maximum concentration permissible (MCP) limits in effect at that time. 

Countermeasures were established to pump out the water and prevent leaching of radioactivity.2 

Rockwell’s 1987 inspection found more water on the north test cell floor of the reactor vault 

floor. Two radionuclides (Eu-152 and Na-22) were found to be above their MCP limits. This 

discovery prompted Rockwell to begin a decontamination and decommissioning program to 

remove the remaining radioactivity.3 

In July 1989, EPA sampling identified tritium in water from the French drain around the west end 

of the basement of Building 4059. Tritium was found at a level far above what could be 

described as a background, but orders of magnitude below the maximum contaminant level at 

the time. It was recommended that further testing be conducted to determine the origin and 

spread of tritium at the site.4 

EPA indicated that there were two findings of interest on the samples taken at Building 4059’s 

French drain. First, that Co-60 or other gamma emitters were not detected, and second, that 

tritium was. According to EPA, tritium was found in a level suggested it came from the facility; 

Rocketdyne previously did not test water for tritium activity. According to EPA, SSFL

personnel were apparently unaware of the presence of tritium at Building 4059. 

1 Atomics International, “Findings of the SNAP 8 Developmental Reactor Post-Test Examination,” by A.H. Lillie and 

V.L. Rooney, Jr. June, 1971 for the AEC, Contract AT(04-3)-701 

2 EPA/HGL, Final HSA 5C p. 22, citation 4: 

4 

Liddy, P., Building 4059 Final Status Survey Report (Phase 1), Boeing Report No. RS-00008, September

11, 1999, p. 4. 

3 EPA/HGL, Final HSA 5C p. 22, citation 4: 

4 

Liddy, P., Building 4059 Final Status Survey Report (Phase 1), Boeing Report No. RS-00008, September

11, 1999, p. 4. 

4 EPA/HGL, Final HSA 5C p. 22, citation 2: 

2 Dempsey, Gregg, Report on Environmental Samples Collected at the Rocketdyne Santa Susana

Field Laboratory, Environmental Protection Agency, Office of Radiation Programs, Las Vegas Facility, July 

1989, pgs. 5-6. 

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5.5.6.4 Building 4059 Additional Information 

Building 4059 housed the SNAP 8 prototype reactor, S8DR, Large Leak Test Rig and Ground 

Prototype Test Facility. By January 1969, the unshielded SNAP 8 prototype reactor was 

operating at 600 kW in the north cell. It continued operating until December 1969 at 600 kW with 

a short period at 1MW. In 1970, the reactor and associated equipment were removed and the 

facility became inactive. In 1973, the Liquid Metal Engineering Center (LMEC) occupied the 

building, removed the shielding and remote handling equipment, and erected the Large Leak 

Test Rig (LLTR) in the test vault and high bay areas.1 In 2004, the building shell and concrete 

cells and foundations were removed after which the large excavation area was backfilled with 

soils that had been removed from the excavation. 

Building 4059 was inactive between 1970 and 1973. In 1973, Rockwell dismantled and removed 

all control consoles and reactor instrumentation, and all contaminated items were removed from 

the test vault. What remained in place were extensive below-grade, radiologically activated 

structures, including steel containment vessels and vacuum lines, concrete biological shielding, 

and sand fill.2 

In 1978, Rockwell removed the radioactive liquid and gas holdup tanks. The reactor core and 

Sodium-Potassium (NaK) systems were removed.3 Residual contamination remained in place. 

Rockwell conducted a survey for the building, soil, and water. The reactor vault was excluded 

from the survey. Beta gamma contamination was measured above the 1978 acceptable limit of 

1,000 dpm/100 cm2. Decontamination was conducted.4

The HSA provides a detailed chronology of radiological investigations and decontamination/

cleanup of release(s) associated with Building 4059. The preliminary MARSSIM Classification 

for Building 4059 is Class 1 due to its use as a reactor building. 

5.5.6.5 Building 4059 Radiological Incident Reports 

There has been one major incident associated with Building 4059, and several incidents that 

could have resulted in a release to the environment and worker exposure. The following table 

1 EPA/HGL, Final HSA 5C p. 15, citation 4: 

4 SNAP 8 Ground Prototype Test Facility – Building 059, fragment, circa 1980, p.1. 

2 EPA/HGL, Final HSA 5C p. 21, citation 6: 

6 Graves, A. W., Decontamination and Disposition Facilities Program Plan, Atomics International Report No.

PP- 704-990-002, January 23, 1975, p. 11. 

3 EPA/HGL, Final HSA 5C p. 21, citation 7: 

7 Sturtevant, W.C. et al., Building 059 Remediation Program, Phase II Reactor Test Cell Remediation, 

EID-04422, June 25, 1999, pp. 24-25. 

4 EPA/HGL, Final HSA 5C p. 22, citation 1:

1 Rockwell International, Radiological Survey Results – Interim Status, Building 059, Santa Susana Field

Laboratory, N704TI990043, November 28, 1978. 

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provides information presented in an incidents database provided by Boeing. Summaries of all 

available incident reports are provided in Attachment A. 

5.5.6.6 Building 4059 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

None 10/29/1969 S8DR Reactor H2 Fuel elements leaking hydrogen and 

mixed fission products during operation.

A0633

Incident 

Report 

Missing

02/12/1970 T059 Test Vault Absolute filter on vacuum cleaner 

ruptured contaminating area

A0576 02/29/2970 Pipe Chase B59 Cutting a NaK line in Pipe Chase Room 

caused NaK fire.

None 11/04/1981 Building 59 Fire broke out, burning arms and hand of 

one worker.

A0183 01/27/1988 Vacuum Duct Room Co-60 Photographer sent into controlled area 

without proper training or protection.

A0187 08/12/1988 Pipe Chase Room Co-60, Eu-152, 

Eu-154

Torch cutting in pipe chase resulted in 

contamination of operators.

A0189 08/19/1988 Pipe Chase Room Co/Eu Breathing airline became disconnected in 

airborne area.

A0194 01/31/1989 Pipe Chase Room Co/Eu Employee entered radiation area without 

film badge on chest, with dosimeter.

A0197 04/06/1989 Filter Bank Co-60 Improper procedure on filter change 

caused high airborne contamination.

A0204 04/12/1990 Test Vault Area Employee entered controlled area w/o 

prescribed monitoring/protective 

equipment.

A0206 05/08/1990 Vacuum Equipment 

Room

Co-60 Exit survey revealed contamination on 

pant leg.

A0212 01/17/1991 Electrical Room K/Co clean area floor sweeping found to 

contain radioactive material

A0214 02/25/1991 Test Cell Co-60 Torching operation generated high 

airborne activity in adjoining high bay.

A0216 09/20/1991 Test Cell Employee entered controlled area 

without a film badge.

A0219 09/02/1992 Test Cell Torching slag violated protecting clothing 

contaminating shoe

A0307 12/03/1992 Test Cell 1 Employee’s dosimeter went off-scale 

during D&D operations

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*Isotopes are written as they are presented in the Incident Database. The HSA research team believes that “MFP” is 

an acronym for “mixed fission products.” 

A0579 07/29/1993 Test Cell 1 Exit survey indicated contamination on 

Tee-Shirt

A0692 12/22/1998 T059 HiBay Contamination of Fork Truck / Gloves 

was due to naturally occurring radon 

daughters.

A0691 01/05/1999 SSFL Naturally occurring radon daughters 

plated on hard hat and gloves.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

• MAJOR INCIDENT: In 1969 Atomics International found fuel elements leaking hydrogen and 

fission products. A group of experts was gathered to analyze the leak and find its cause. At the 

1 time, the zirconium hydride reactor was being considered for numerous space missions. This 

was the second or third most serious accident at SSFL. The reactor operated for a year with 

failing fuel. Detailed description above.

• On February 13, 1970 an absolute filter on a vacuum cleaner ruptured contaminating the 

Building 4059 test vault. The incident report has not been located (A0633). 2

• On February 19, 1970 a sodium/potassium (NaK) fire broke out in the Pipe Chase Room while 

an employee was cutting a pipe during demolition activities. Approximately one teacup of NaK 

spilled into a drip pan on the floor and began to smolder. The fire was extinguished by 

shoveling calcium carbonate onto the pan. When a dense cloud of white smoke filled the vault, 

personnel evacuated, bolting the access door in place. A subsequent inspection of the exhaust 

duct filter revealed no increase in airborne radioactivity (A0576).3

• On November 4, 1981 a fire broke out at Building 4059. The cause and region of the fire are 

unclear from the report. A worker received burns to his right hand and arms and is taken to the 

“nurse.” The local fire service extinguished the blaze using three hoses. 4

• On January 27, 1988 a photographer was sent to the Vacuum Duct Room without proper 

safety precautions. He entered a “High Radiation Area” without protective clothing, a film 

badge, or a pocket dosimeter. Upon exiting the room, one pant leg was found to be 

contaminated. Radiation levels in the area were in excess of 100 mR/h). Loose contamination 

1 EPA/HGL, Final HSA 5C p. 18, citation 1: 

1 U.S. Atomic Energy Commission, Letter from M. Klein, AEC, to J. J. Flaherty, Atomics International, dated 

October 29, 1969. 

2 EPA/HGL, Final HSA 5C p. 18, citation 2: 

2 The Boeing Company Radiation Incident Database, 2010. 

3 EPA/HGL, Final HSA 5C p. 18, citation 3: 

3 Bresson, J. F., Rockwell Internal Letter to W. F. Heine, re: Fire in Building T059 Vault, February 26, 1970. 

4 EPA/HGL, Final HSA 5C p. 18, citation 4: 

4 Incident Report, dated November 4, 1981 (BNA05615738-744). 

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consisted primarily of Co-60 in sand. The contamination was removed successfuly at the 

RMHF (A0183).1

• On August 12, 1988 two employees were contaminated as a result of cutting in the Pipe 

Chase Room during demolition activities. Contamination was measured at 17,500 dpm of 

beta-gamma contamination per 100 square centimeters (17,000 dpm / cm2). These employees 

were subsequently decontaminated (A0187).2

• On August 19, 1988 an employee working in the Pipe Chase Room in an air-line respirator felt 

his filter cartridge pod disconnect from his face mask. He picked it up and had his co-worker 

try to reconnect it. His coworker thought the coupling was broken so they evacuated the room 

within two minutes. Nasal and facial smears and the inside of the face mask indicated no 

detectable activity. Analysis of the work area air sample indicated 1.5 x 10-11 μCi/mL, below 

the maximum permissible concentration of 9 x 10 9 μCi/mL. A warning letter was issued 

alerting respirator users to the risk of uncoupling (A0189).3

• On January 31, 1989 a radiation worker entered a radiation area for work without a personally 

issued film badge although he wore the required pocket dosimeter, two TLD LiF chips finger 

rings (one on each hand), a visitor film badge on his foot, and a TLD-CAF. A review of the 

pocket dosimeter and the TLD indicated an accumulated exposure of 74 mrem. Greater 

caution to observe safe work practices was recommended (A0194).4

• On April 6, 1989 an employee dropped an open box of used filters (approximately 10 feet) 

while changing the exhaust system filters during torching operations on the upper level of 

Building 4059. A cloud of dust arose from the filters when they landed on the floor and a “lot of 

dust” settled on the floor. Higher counts were recorded by the radiation detector. The 

employee’s hands (4,350 dpm) and shoes (5,800 dpm) were subsequently decontaminated. 

no other activity was detected on his body or clothing. The filters were to be disposed of as 

radioactive waste and the area was to be vacuumed (A0197).5 

1 EPA/HGL, Final HSA 5C p. 19, citation 1: 

1 Tuttle, R. J., Rockwell Internal Letter to J. A. Chapman, re: Radiological Safety Incident Report, T059 

Vacuum Duct Room, January 27, 1988. 

2 EPA/HGL, Final HSA 5C p. 19, citation 2: 

2 Wallace, J. H., Rockwell Internal Letter to Radiation and Nuclear Safety Group, re: Radiological Safety

Incident Report, T059 Pipe Chase Room, August 12, 1988. 

3 EPA/HGL, Final HSA 5C p. 19, citation 3: 

3 Wallace, J. H., Rockwell Internal Letter to Radiation and Nuclear Safety Group, re: Radiological Safety

Incident Report, T059, August 19, 1988. 

4 EPA/HGL, Final HSA 5C p. 19, citation 4: 

4 Saba, V. B., Rockwell Internal Letter to Radiation and Nuclear Safety Group, re: Radiological Safety

Incident Report, SSFL Building 059, Reactor Vault, January 31, 1989. 

5 EPA/HGL, Final HSA 5C p. 19, citation 5: 

5 Begley, F. E, Rockwell Internal Letter to Radiation and Nuclear Safety Group, re: Radiological Safety

Incident Report, Lower Level of T059, April 6, 1989. Ref. 122119552 

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• On April 12, 1990 an employee entered the test vault area on the lower level of Building 4059 

without approval, without plastic shoe covers, and without a personally issued film badge or 

pocket dosimeter. When discovered and reprimanded, the employee left the area 

approximately three minutes after entering the vault (A0204).1

• On May 8, 1990 at the end of his shift, an employee surveyed himself and found a hot spot 

on his pant leg. Fixed contamination was localized to the 1-inch by 1-inch area on the pant 

leg reading 300 cpm. The employee had been decontaminating the vacuum equipment room 

shield blocks wearing a red-line lab coat. After several unsuccessful decontamination 

attempts, the pants were disposed of as radioactive waste. It was recommended that future 

work with shield blocks would require red-line coveralls (A0206).2

• On January 17, 1991 floor sweepings in the electrical room were found to contain Co-60 

(0.85 pCi/g). The sweepings were disposed of as radioactive waste (A0212). 3

• On February 25, 1991 open-tent torching operations generated high airborne Co-60 

radioactivity in the adjoining high bay during demolition activities. Nasal smears revealed 

that three of seven workers had collected detectable activity. All seven employees submitted 

1-day and 5-day samples for bioassay. Operations ceased until the Co-60 radioactivity in the 

air dropped to a safe level (A0214).4

• On September 20, 1991 an employee noticed that his film badge was missing while working 

on a plastic containment shroud above the test cell tent in the reactor vault. He left the area 

and notified the facility health physicist. A search for the film badge was conducted, but the 

film badge was not found. A new film badge was issued to the employee (A0216). 5

• On September 2, 1991 an employee performing torching operations had hot slag penetrate 

two pairs of plastic booties and lodge in the bottom of his shoe. The shoe was 

decontaminated by napping the leather and using duct tape to remove the slag. After 

1 EPA/HGL, Final HSA 5C p. 20, citation 1: 

1 Saba, V. B., Rockwell Internal Letter to Radiation and Nuclear Safety Group, re: Radiological Safety

Report, SSFL Building 059, April 12, 1990. 

2 EPA/HGL, Final HSA 5C p. 20, citation 2: 

2 Wallace, J., Rockwell Internal Letter to Radiation and Nuclear Safety Energy Systems Group, re: 

Radiological Safety Incident Report, T059 Vacuum Equipment, May 8, 1990 

3 EPA/HGL, Final HSA 5C p. 20, citation 3: 

3 McGinnis, E. R., Rockwell Internal Letter to Radiation Protection and Health Physics Services, re:

Radiological Safety Report, T059 Electrical Room, January 17, 1991. 

4 EPA/HGL, Final HSA 5C p. 20, citation 4: 

4 McGinnis, E. R., Rockwell Internal Letter to Radiation Protection and Health Physics Services, re:

Radiological Safety Report, T059 Restricted Area, February 25, 1991. 

5 EPA/HGL, Final HSA 5C p. 20, citation 5: 

5 McGinnis, E. R., Rockwell Internal Letter to Radiation Protection and Health Physics Service, re:

Radiological Safety Report, T059 Vault, September 20, 1991. 

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decontamination no beta-gamma activity was detected on the shoe (A0219).1

• On December 3, 1992 an employee’s dosimeter showed an off-scale reading after exiting 

Test Cell 1 during decontamination and decommissioning activities. The employee’s film 

badge was sent to the vendor to be read, and his dosimeter was sent to the instrument 

laboratory for an “as found” test. The film badge was processed the next day and read 170 

mR for the quarter, within acceptable limits. Rockwell concluded that the increased dose rate 

could have been due to exposure of activated steel behind the portion of wall being removed 

(A0307).2

• On July 29, 1993 an employee upon exiting a pit was found to have 150 cpm beta gamma 

contamination on the rear shoulder of his personal tee shirt. He removed the tee shirt and no 

contamination was detected on his skin. The employee was wearing the tee shirt because of 

a previous contamination event in Building 4020 (Hot Laboratory) (A0579).3

• On December 22, 1998 a fork lift truck was surveyed for radioactive contamination in the 

Building 4059 high bay. Naturally occurring radon daughters were found to have been plated 

on a hard hat and gloves. Rockwell concluded that this was a natural phenomenon 

(A0692).4

• On January 5, 1999 an employee was found to have a hard hat and leather gloves 

contaminated with an alpha emitter. Based on isotopic analysis, the material was determined 

to be naturally occurring radon daughters. Rockwell concluded that this was a natural 

phenomenon (A0691).5

1 EPA/HGL, Final HSA 5C p. 20, citation 6: 

6 McGinnis, E. R., Rockwell Internal Letter to P. D. Rutherford, re: Radiological Safety Report, T059 Test Cell 

1, September 2, 1992. 

2 EPA/HGL, Final HSA 5C p. 21, citation 1: 

1 McGinnis, E. R., Rockwell Internal Letter to P. D. Rutherford, re: Radiological Safety Report, T059 Test Cell 

1, December 3, 1992. 

3 EPA/HGL, Final HSA 5C p. 21, citation 2: 

2 Wallace, J. H., Rockwell Internal Letter to P. D. Rutherford, re: Radiological Safety report, T059 RPT Pit, 

July 29, 1993. 

4 EPA/HGL, Final HSA 5C p. 21, citation 3: 

3 Redacted, Rockwell Incident Report, re: Contaminated Fork Truck/Gloves, December 22, 1998. 

5 EPA/HGL, Final HSA 5C p. 21, citation 4: 

4 

Redacted, Rockwell Incident Report, re: Contaminated Hard Hat and Gloves, January 5, 1999. 

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6.0 CRITICAL TEST FACILITIES 

Several program used critical test facilities (i.e. low-power reactors) in Area IV. Use of these lowpower reactors began in 1954 and continued until 1974. Some of the Critical Test Facilities at 

SSFL were used for the production of electricity, and others were used to develop energy 

resources for the United States Space Program under DOE-NASA contracts (like the SNAP). 

6.1 SNAP DEVELOPMENT TEST FACILITIES 

Test facilities in Area IV for the SNAP Program included Buildings 4373, 4012, 4019 and 4024. 

CORE Advocacy identified several locations associated with SNAP operations and support that 

should have been included in the 2006 Site Description, but were not. Some were nonradiological facilities that are documented as having engaged in radiological processes. They 

are described in detail, in Section 6.2. 

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Section 6.1.1 

Building Number:

Building Alias: 

4373 

373 / T373 

Building Name: SNAP Critical Facility

Building Function: Built to store and test highly explosive solid rocket fuel

First SNAP Critical Facility 

Notes: DOE-NASA Program

Tritium 

Prior to 1957, this location was used for rocketry purposes and operated by Rocketdyne, 

verifying (again) worker rotation and cross-facility operations. 

Radionuclides of Concern: Uranium, plutonium, and their decay and daughter products: 

U-235, U-238, and Th-232 decay chains include Th-234, Th-228, Ac-228, Ra-226, Pb-214, 

Bi-214, Pb-212, Bi212, and Tl-208. In addition, H-3, Sr-90, Cs-137, Na-24, Co-57, Co-60, 

Eu-152, and Am-243 would have been formed. Also gadolinium (Gd-64) and samarium 

(Sm-153). 

6.1.1.1 Description of Operations & Processes 

The 2006 Site Description provides confusing and inaccurate information about Building 4373, 

indicating that it ceased operations in 1963 and was “replaced by Building 4012.” However, 

numerous criticality tests are documented at Building 4373 after 1965; it was referred to as a 

“Critical Cell and Development Test Building” on industrial maps as late as 1975; as of 1987 it 

was reported to be unoccupied and abandoned. As of 1989, it still had not been cleared of old 1

equipment, and was determined to be a “mess.”2 In 1995, Rockwell International devised a 

procedure to demolish the rubidium loop that remained in the building, which had been installed 

in the 1960’s. In addition, during the pre-demolition survey, three radioactive pre-filters from the 

ventilation system were found on the roof of Building 4373, thought to have been in continuous 

use from 1957 to 1962; they were not replaced or removed since building operations ceased 

and had apparently remained on the roof in excess of 30 years. Radioactivity of each pre-filter 

indicated11,250 dpm/100 cm2 of beta activity. Gamma spectroscopy determined that the activity 

was predominantly Cs-137. As a result, the incident report summarized that the high levels of 

Cs-137 were the result of building operations, as well as from atmospheric weapons testing 

(Incident Report A0664).3 

1 EPA/HGL, Final HSA 5D p. 109, citations 6&7: 

6 Schrag, F.C., Internal Letter Re: Building 373, SSFL, June 19, 1987.

7 Schrag, F.C., Job Improvement Request, May 22, 1987. 

2 EPA/HGL, Final HSA 5D p. 109, citation 9: 

9 Robinson, K.S., Rockwell International Internal Letter Re: Alkali Metal Waste in Area IV, January 22, 1989. 

3 EPA/HGL, Final HSA 5D p. 111, citation 1: 

1 

Harcombe, Richard, Internal Letter Re: Radiological Incident Report A0664, August 25, 1995. 

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According to the HSA, Building 4373 was constructed in 1956. It was originally intended for use 

as a testing and handling facility for high explosive solid rocket fuels. Use of this location for 

rocketry-related purposes further supports the lack of separation between DOE nuclear 

operations and NASA rocketry endeavors, at SSFL and illustrates that North American Aviation 

Rocketdyne employees likely associated with the Space and Information operations in Areas I, II 

and III also had access to certain Area IV facilities associated with rocketry research, and space 

nuclear propulsion. This information supports the inclusion of Area I-III workers to EEOICPA. 

A 1957 and 1958 progress report indicated that the building was first modified to provide field 

test facilities for the performance of the “ANN” program.1 One test call was modified for critical 

assembly research supporting SNAP in 1957. Test areas were separated from the 

administrative and general operating areas by a concrete shield wall. Part of the building was 

used for miscellaneous non-nuclear tests that appear to have changed throughout the operation 

of the building.2 

The test cells were originally designed as explosion blow-out rooms and were provided with 12-

inch thick concrete inside walls. An additional 2 feet of magnetite concrete was added to the 

reactor test cell wall adjacent to the administrative area, and 2 feet of ordinary concrete on the 

wall separating the reactor test cell and adjacent test cell. The roof consisted of metal joists and 

cemesto panels with built-up roofing material; it did not include any shielding. According to a 

February 1962 summary hazards report and operations manual for the SNAP 8 critical 

experiments to be conducted in the building, the location of the critical machine within the test 

cell had been changed at some time not documented in the report. According to the report, the 

structures of the previous “critical machines” had been removed from the cell. The document 

does not provide information to indicate how the previous critical machines had been disposed. 

A figure of the layout of the test cell shows the critical machine to be located at the center of the 

test cell.3 

Based on information included in the 1988 survey report, Building 4373 was reportedly designed 

to contain radioactivity in the form of fissile material and activation foils; accordingly, no liquid 

radioactive waste was produced and no liquid radioactive waste holdup tanks were included at 

building 4373.4 

1 EPA/HGL, Final HSA 5D p. 103, citation 3: 

3 Atomics International, Facilities and Data Department, Progress Report Fiscal Year 195 and Forecast 

Fiscal Year 1958, January 1958. 

2 EPA/HGL, Final HSA 5D p. 103, citation 6: 

6 Ashley, R.L., Atomics International Document No. NAA-SR-7300, Evaluation of the Atomics International 

Nuclear Development Field Laboratory as a Location for Reactor Facilities, May 25, 1962. 

3 EPA/HGL, Final HSA 5D p. 104, citation 1: 

1 Thiele, A.W., Atomics International Document NAA-SR-MEMO-7029, SNAP 8 Critical Experiment 

Summary Hazards Report and Operations Manual, February 1, 1962. 

4 EPA/HGL, Final HSA 5D p. 104, citations 2 & 3: 

2 ETEC Document, GEN-ZR-0012, Radiological Survey of Buildings T373 and T375, August 8, 1988.

3 Author Unknown, Facility Information Bldg 373, Date Unknown. 

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The administrative and general operating area, located on the opposite side of the shield wall 

from the test area, included a 9’x11’ change room that led to the fuel storage room and main test 

cell, a 20’x20’ L-shaped control room, a laboratory area, instrument shop, equipment room, and 

office area. The building was surrounded by a chain-link security fence and included a 

secondary fence that prevented the use of the areas west and south of the main test cell. 

According to the report, the walls at this location “may not provide adequate shielding in the 

event of an excursion.”1 

According to the January 1962 summary hazards report and operations manual, the fuel 

elements used in the building were stored in the fuel storage room in the shipping bird cages 

provided by “Fuel Fabrications.” While only one bird cage was permitted in the test cell during 

fuel loading operations, no cages were reportedly permitted in the test cell curing critical 

experiments. 

A Log Book of operations at Building 4012 describes several incidents over many months, 

involving contaminated birdcages containing broken pieces of SNAP 8 fuel and highly irradiated 

metal shavings, which were apparently circulated around the SNAP complex repeatedly. 

Several incident reports throughout the SNAP Complex detail the receipt of “contaminated 

birdcages” in various stages of severe contamination.2 

Between 1957 and May 1962, several tests on SNAP critical assemblies were completed, as 

indicated in the summary below.3 

! SCA-1. Initiated October 1957. This assembly consisted of a pseudo sphere of 

zirconium hydride-enriched uranium dioxide blocks. Basic reactor parameters of the 

SNAP 2 reactor concept were determined. 

! S2ER Critical Experiment. Initiated June 1959. The S2ER components were 

assembled and preliminary tests conducted at zero power. 

! SCA-2. Initiated about September 1960. A clean, cylindrical geometry core was 

studied, using the core and reflector components from the conduction-cooled SNAP10 

reactor. 

! SCA-3. Initiated October 1961. An assembly of about a 1-cubic foot core volume was 

built to study the characteristics of the SNAP4 reactor. Plate-type fuel elements and a 

water coolant were used. 

1 EPA/HGL, Final HSA 5D p. 104, citation 4: 

4 Ashley, R.L., Atomics International Document No. NAA-SR-7300, Evaluation of the Atomics International 

Nuclear Development Field Laboratory as a Location for Reactor Facilities, May 25, 1962. 

2 Atomics International Health Physics Log Book, Building 4012, 10/12/62-7/27/66 File: HDMSP001852993.pdf 

3 EPA/HGL, Final HSA 5D p. 105, citation 5: 

5 Ashley, R.L., Atomics International Document No. NAA-SR-7300, Evaluation of the Atomics International 

Nuclear Development Field Laboratory as a Location for Reactor Facilities, May 25, 1962. 

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Future experiments that were planned, but have not yet been confirmed, included the following:1 

! SCA-4C. This assembly was to use prototype SNAP2/10A core and reflector 

components, as a final design check on the flight reactor system. The test was 

tentatively scheduled to begin in mid-1962. 

! S8ER Critical Experiment. The S8ER components were to be used to determine the 

nuclear parameters of the SNAP 8 reactor design. The test was tentatively scheduled to 

begin in mid- 1962. 

! SCA-4A. This assembly was to provide a flexible reflector assembly to permit 

optimum design of the reflector controls for the SNAP2/10A reactor. This test was 

tentatively scheduled to begin late in 1962. 

The evaluation report indicated that following these tests, the facility was to be deactivated and 

that the critical experiment program would be continued in Building 4012. Prior to being 

deactivated, a meeting held in March 1960 discussed the expanded use of Building 4373. 

According to a letter summarizing the meeting, it was recommended that the “water test room” 

adjacent to the critical assembly in Building 4373 be used for intermittent liquid metal tests. The 

letter stated that the measured dose in the room is 1 mrem/hr during operation of the critical 

assembly. The letter did not indicate the nature of the liquid metals tests to be conducted.2 

According to some documents, SNAP critical tests in Building 4373 ceased in 1962 or 1963. 

However, an undated facility information sheet for Building 4373 indicated that a log book of the 

facility operations was kept and maintained in the control room. Documents received to date 

have included the Building 4373 log book, dated January 11, 1962 through May 28, 1965, that 

appear to indicate operations continued after 1963. A summary of select entries of the log book 

is presented below. However, the log book indicates routine surveys of beryllium and birdcages 

containing 93% enriched uranium.3 

An April 8, 1963 internal letter provided a summary of a March 20, 1963 meeting that provided a 

review of Building 4373 operations. According to the letter, Building 4373 was being used to test 

the SCA-4C in early 1963. The meeting summary indicated that the experimental program in 

Building 4373 was extensive despite previous correspondence indicating that Building 4373 was 

not recommended for a long-range program of critical experiments, due to “certain undesirable 

aspects of the ventilation system and the building design.” However, it was reported the 

maximum credible accident for the facility assumed no containment and that changes to the 

ventilation system would require extensive modification to the building structure. It was 

recommended that Buildings 4012 and 4024 be used as much as possible for critical tests at the 

SSFL. The letter also indicated that the meeting recommended that “serious consideration 

1 EPA/HGL, Final HSA 5D p. 105, citation 5: 

5 Ashley, R.L., Atomics International Document No. NAA-SR-7300, Evaluation of the Atomics International 

Nuclear Development Field Laboratory as a Location for Reactor Facilities, May 25, 1962. 

2 EPA/HGL, Final HSA 5D p. 106, citation 1: 

1 Jarrett, A.A., Letter Re: Expanded Use of Building 373, March 30, 1960. 

3 EPA/HGL, Final HSA 5D p. 106, citation 3: 

3 Log Book, Building 373, January 11, 1962 to May 28, 1965. 

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should be given to using a weak source during fuel loading, in order to provide possible 

additional safety to personnel.” The meeting minutes indicated that no source was used during 

fuel loading in Building 4373 prior to this time.1 The HSA research team reviewed the entire 

collection of available documents provided by Boeing, DOE and other sources and could not 

find information to indicate if a radioactive source was used in Building 4373 after March 1963. 

According to the log book for Building 4373, the SCA-4C was being dismantled in May 1963 for 

storage at Building 4005.2 

In addition, Atomics International held a semiannual review of Building 4373 on September 16, 

1964. According to the minutes of that meeting, Building 4373 continued to conduct experiments 

under the SNAP Critical Assembly Program. From March to September 1964, testing was 

completed in support of the SNAP-10B program. The meeting summary indicated the testing 

involved “the investigation of the worths of reflector under different arrangements, core critical 

loading measurements and materials reactivity effect studies.” The experiments were classified 

in more detail, as follows:3 

“Some experiments were studies involving materials such as samarium, gadolinium, and 

various fuel element cladding materials. In these experiments, the reactivity effects of 

the materials placed in the core were measured. Other experiments included studies of

different types of control elements, and included the sliding drum, the tapered drum, and 

split drum type.” 

The meeting minutes also indicated that Building 4373 would be used in the next six months in 

support of the SNAP-10B and SNAP 8 reactor concepts, and experiments would be conducted 

to obtain data on neutron lifetime, the worth of individual control drums and other reactor 

constants. The minutes indicated that an addendum to the Summary Hazards Report would be 

required for experiments related to the SNAP-8 program, and the building would require 

modification. Documents were not located to indicate whether the building modifications 

required to accommodate the SNAP-8 program were made to Building 4373 following this 

meeting.4 

The meeting summary indicated that no unusual incidents occurred during the “past survey 

period.” It was reported, however, that a “slight case of U-235 foil contamination did occur.” 

1 EPA/HGL, Final HSA 5D p. 106, citation 4: 

4 Remley, M.E., Internal Letter Re: Review of Bldgs 012 and 373 (NDFL) Recommendations and Minutes of 

Meeting of March 20, 1963, April 8, 1963. 

2 EPA/HGL, Final HSA 5D p. 106, citation 5: 

5 Log Book, Building 373, January 11, 1962 to May 28, 1965. 

3 EPA/HGL, Final HSA 5D p. 106, citation 6: 

6 Remley, M.E., Internal Letter Re: Semiannual Review of Building 373 – Recommendations and Minutes of 

Meeting of September 16, 1964, September 24, 1964. 

4 EPA/HGL, Final HSA 5D p. 107, citation 1: 

1 Remley, M.E., Internal Letter Re: Semiannual Review of Building 373 – Recommendations and Minutes of 

Meeting of September 16, 1964, September 24, 1964. 

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According to the minutes, “contamination levels were of minor concern and were easily cleaned 

up.”1 

According to a June 24, 1965 internal letter regarding a March 29, 1965 meeting and a review of 

Buildings 4012 and 4373 for the use of poison-loaded Mylar films in critical experiments, 

Building 4373 had been approved for Mylar film experiments in SCA-41 and SCA-4B. The letter 

indicated the purpose of the experiments was to use Mylar films to introduce neutron absorbing 

material in experimental studies of proposed SNAP 2 and SNAP 10 reactor cores. The objective 

was to develop a reactor core that would be subcritical when immersed in water, but would have 

the desired excess reactivity when filled with NaK and surrounded by the reflector and control 

drum assembly. The experiments were designed to have the fuel elements and core vessel 

have layers of gadolinium of several thicknesses. The gadolinium was to be attached with Mylar 

film. The experiments in Building 4373 were to be with dry cores.2 

On June 8, 1965 Atomics International submitted comments on a draft reactor safety survey 

report dated April 22, 1965. The report was titled, “SNAP Critical Assembly-4A, Building 373.” 

Atomics International provided the following additional information regarding the Mylar 

experiments being conducted in Building 4373: 

“At the time of the survey an experiment was in progress to provide information that

might result in the design of SNAP reactors that would be subcritical when accidentally 

immersed in water. This reduction in the reactivity of the water flooded and reflected 

systems as compared to the beryllium reflected system would be achieved through the

use of an absorber with a strong low energy resonance, such as gadolinium, which

would produce a shift in the effective energy of the neutrons causing fission. In this 

experiment the poison was in the form of gadolinium oxide applied to a Mylar film, which

was applied to either the fuel elements or to the core vessel as required by the

experiments program.” 

The letter also stated that Atomics International was in the process of developing and 

implementing a procedure for the in-place testing of “absolute” filters, including those in Building 

4373, which was to be done in the near future. Available documents reviewed by the research 3

team do not provide the results of those tests. 

According to the log book, it also appears that Building 4373 served as a monitoring and tagging 

station for “birdcages” containing various fuels. According to a number of log book entries, 

birdcages were monitored and tagged at Building 4373 and were then shipped to Building 4064. 

A log book entry dated July 11, 1963 indicated that personnel “tagged and shipped 8 birdcages 

1 EPA/HGL, Final HSA 5D p. 107, citation 2: 

1 Remley, M.E., Internal Letter Re: Semiannual Review of Building 373 – Recommendations and Minutes of 

Meeting of September 16, 1964, September 24, 1964. 

2 EPA/HGL, Final HSA 5D p. 107, citation 3: 

3 Remley, M.E., Internal Letter Re: Review of Buildings 012 and 373 and the Use of Poison Loaded Mylar 

Films in the SCA-4A and -4B Experiments, Minutes of Meeting of March 29, 1965, June 24, 1965. 

3 EPA/HGL, Final HSA 5D p. 108, citation 1: 

1 Remley, M.E., Letter Re: Comments on Draft Report of Reactor Safety Survey Report, SNAP Critical 

Assembly- 4A, Building 373, June 8, 1965. (Export Controlled Document) 

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containing SNAP 10FS-1 backup fuel” to Building 4064. The same day, it was reported that the 

SCA-4 fuel would be returned to Building 4373. On July 17, 1963, personnel monitored and 

tagged 8 birdcages containing SCA-4 fuel to be shipped to Building 4064, and personnel 

received 4 birdcages containing S10FS-1 fuel. The entry that day indicated that, “a total of 44 

S10FS-1 fuel is now in” Building 4373. There are a number of additional references to the 

tagging and monitoring of birdcages at Building 4373; however, additional information regarding 

the nature and extent of these operations could not be located.1 

At the conclusion of the SNAP critical tests, the facility was modified to include a non-nuclear 

liquid metal NaK test loop to support the SNAP Experimental Reactor. At this time, industrial 

planning maps refer to the building as a “critical cell and development test building,” and it is 

referred to as such until 1975, when industrial maps refer to Building 4373 as a “test building.” 

According to the undated facility information sheet for Building 4373, NaK experiments were 

conducted in the North Laboratory Room 102 and in the North Test Room 113. The small test 

loop was used to pre-test a NaK/NaK heat exchanger, EM pumps, electrical heaters, valves, 

piping and instrumentation. Other non-nuclear test loop programs carried out in the Building 

4373 were RuK test loops, boiling mercury test loops and boiling potassium loops. The facility 

has since been used intermittently for storage of non-radiological materials.2 

According to a 1988 survey report, Building 4373 had no significant radioactivity releases during 

its operation. Building 4373 was reassigned for non-nuclear use following a radiological survey 

of the building to ensure the area was “clean.” Documents available to date did not include the 

referenced radiological survey. The report stated that “most” radioactive materials handled in 

Building 4373 were totally encapsulated or “in a condition which would not release radioactivity 

to the environment.” Radioactive materials handled at the building included fully encapsulated 

highly enriched uranium. Activation foils were used for flux mapping in Building 4373.3 

A July 1966 review of Building 4373 indicated it had been shut down “for some time.” The 

review stated that the SCA-4A “machine” remained in Building 4373 and was being considered 

for transfer to Building 4012 for further operations. Based on a 1969 safety analysis report, the 

1 EPA/HGL, Final HSA 5D p. 108, citation 2: 

2 

Log Book, Building 373, January 11, 1962 to May 28, 1965. 

2 EPA/HGL, Final HSA 5D p. 108, citations 3-8: 

3 ETEC, Document No. GEN-ZR-0012, Radiological Survey of Buildings T373 and T375, August 8, 1988.

4 Rockwell International, A4CM-AN-003, Area IV Characterization Survey Plan, March 30, 1994.

5 Rockwell International, Document No. N001ER000017, Nuclear Operations at Rockwell’s Santa Susana 

Field Laboratory – A Factual Perspective, December 20, 1989.

6 Author Unknown, Facility Information Bldg 373, Date Unknown.

7 Atomics International, Santa Susana Map , July 14, 1964.

8 Atomics International, Nuclear Development Field Laboratories, Santa Susana, March 1975. 

3 EPA/HGL, Final HSA 5D p. 108, citation 9: 

9 ETEC, Document No. GEN-ZR-0012, Radiological Survey of Buildings T373 and T375, August 8, 1988. 

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machine remained in Building 4373 in 1969. Atomics International received permission from the 

AEC to operate the SCA-4A in Building 4012 on February 12, 1970.1 

In addition, the July 1966 annual review indicated that following the removal of fuel from the 

facility, a terminal radioactivity and beryllium survey was performed by Health and Safety. The 

results of the survey were not presented in the annual review report.2 However, a 1984 letter 

regarding environmental assessments indicated that Building 4373 was contaminated with 

radioactive materials and that decontamination of the facility had been performed. In 1987, the 

building was reported to be unoccupied, abandoned, and inactive. It was reported that it had not 

been used in over 20 years. It was requested that the building be cleaned out, the trash 

disposed of, and a post added to designate the building as “inactive.” 

In 1988, Building 4373 was reported as being dilapidated inside, with many interior walls 

removed. Conduit and miscellaneous debris was scattered about and hanging from the ceiling. 

The 1988 survey report described the building as a “mess.” In 1989, a search for alkali wastes 3

was conducted by Rockwell International in Area IV. Sodium, potassium, or lithium were 

identified as being located in an old loop within Building 4373, indicating that the building had 

not been cleared of old equipment. As of 1994, the building remained abandoned; however it is 4

unclear whether its condition had been improved or changed since the 1988 survey report.5 

In 1995, Rockwell International developed a procedure to demolish the rubidium loop still 

located in Building 4373. According to the procedure report, the original loop was built in the 

early 1960’s to investigate the heat transfer and fluid flow characteristics of boiling rubidium at 

temperatures of 1,400 to 1800 ̊ F. 

6.1.1.2 Building 4373 Radiological Incident Reports 

There were no incident reports during the operational period of Building 4373, contained in the 

Incident Report Database. However, documents obtained by EPA from Boeing indicate that on 

September 23, 1963 an incident occurred involving a high airborne activity reading on the air 

monitor. During this incident, alpha activity was observed to rise from 50 to 200 cpm, and beta 

activity rose from 1,000 to 10,000 cpm. 

1 EPA/HGL, Final HSA 5D p. 109, citation 4: 

4 Remley, M.E., Atomics International Letter Re: Annual Review of Buildings 012 and 373 – Minutes of 

Meeting of May 3, 1966, July 7, 1966. 

2 EPA/HGL, Final HSA 5D p. 109, citation 4: 

4 Remley, M.E., Atomics International Letter Re: Annual Review of Buildings 012 and 373 – Minutes of 

Meeting of May 3, 1966, July 7, 1966. 

3 EPA/HGL, Final HSA 5D p. 109, citation 8: 

8 ETEC, Document No. GEN-ZR-0012, Radiological Survey of Buildings T373 and T375, August 8, 1988. 

4 EPA/HGL, Final HSA 5D p. 109, citation 9: 

9 Robinson, K.S., Rockwell International Internal Letter Re: Alkali Metal Waste in Area IV, January 22, 1989. 

5 EPA/HGL, Final HSA 5D p. 109, citation 10: 

10 Rockwell International, A4CM-AN-003, Area IV Characterization Survey Plan, March 30, 1994. 

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According to a February 1964 recounting of this incident, the cause of the temporary high 

reading was unknown but was believed to have been from “naturally occurring radioactive 

elements.” This, according to the document, was “based on the fact that the incident is 

qualitatively reproducible by leaving the cell door open with the fan off, during certain weather 

conditions.” The health physicist assigned to Building 4373 at that time discounted the 

possibility of a fuel leak or fuel contamination because smear surveys did not show any 

contamination within the building. It was noted that the incident could have been related to an 

earthquake that had occurred early during the same day.1 The HSA research team was unable 

to locate additional information about this event. 

A September 1964 semiannual review of Building 4373 operations indicated that a “slight case 

of U-235 foil contamination” occurred, but that “contamination levels were of minor concern and 

were easily cleaned up.” Additional documentation regarding this contaminating incident could 

not be located in available historical documents.2 

During the pre-demolition survey of Building 4373, on August 22, 1995 a Health Physicist (HP) 

found three radioactive prefilters in the ventilation system on the roof of Building 4373. 

According to the incident report, the filters were thought to have been in continuous use from 

1957 to 1962, and were not replaced or removed since building operations ceased. The filters 

were removed and taken to Building 4020 for storage in a radioactive materials storage area. 

Following the removal of the prefilters, which had apparently remained on the roof of Building 

4373 in excess of 30 years, the ventilation system was surveyed and released, although the 

incident report did not provide the results of the survey. The survey results of each pre-filter 

indicated 11,250 dpm/100 cm2 of beta activity. Gamma spectroscopy determined that the activity 

was predominantly Cs-137. As a result, the incident report summarized that the high levels of 

Cs-137 were the result of building operations, as well as from atmospheric weapons testing 

(Incident Report A0664).3 

The HSA provides a detailed account of the 1998 Rocketdyne Radiological Survey that involved 

Buildings 4373, 4374, and 4375 as well as subsequent radiological investigations and 

radiological use authorizations. 

The preliminary MARSSIM Classification for Building 4373 is Class 1 because of the building’s 

former operations as a critical facility in support of the SNAP program. 

1 EPA/HGL, Final HSA 5D p. 110, citation 6: 

6 Remley, M.E., Atomics International Letter Re: Semiannual Survey of the SNAP Critical Facility and 

Minutes of Meeting of February 17, 1964, February 27, 1964. 

2 EPA/HGL, Final HSA 5D p. 110, citation 7: 

7 Remley, M.E., Atomics International Letter Re: Semiannual Review of Building 373 – Recommendations 

and Minutes of Meeting of September 16, 1964, September 24, 1964. 

3 EPA/HGL, Final HSA 5D p. 111, citation 1: 

1 

Harcombe, Richard, Internal Letter Re: Radiological Incident Report A0664, August 25, 1995. 

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6.1.1.3 Building 4373 Select Log Book Entries: January 11, 1962 - May 28, 1965 

Date Log Entry

10/19/1962 A smear survey was made of the birdcage that was decontaminated. The contamination level was 

< 1dpm alpha, and <30 dpm beta-gamma.

10/19/1962 A smear survey was made of 6 birdcages, reactor room, storage room, and change room. The 

results follow. All birdcage 3 dpm/100 cm2 alpha to 60 dpm/100 cm

< 30 dpm/100 cm

10/31/1962 Called operations and told them the results of a smear survey taken on the birdcages and in the 

reactor vault. Everything was clean except for the tritium source. A smear was taken on the outside 

of the glass tube. 10,000 dpm tritium.

11/5/1962 Made a smear survey for tritium, beryllium, and beta-gamma. Found four positive tritium smears in 

the reactor vault, three of these were on the neutron accelerator. 11,700 dpm end plate / 18,050 

dpm first glass tube / 600 dpm on the aluminum pipe / 500 dpm on the mockup plate - all were 

tritium. Found on positive beta-gamma near the outside of the RX vault door, 89 dpm beta-gamma.

11/30/1962 Monitored and tagged 8 birdcages containing 93% Enr. U. fuel elements. Smears showed <1 dpm 

alpha and <30 dpm beta-gamma. All 8 birdcages were shipped to Bldg. #064 [4064].

12/3/1962 1100 - Monitored and tagged 6 birdcages containing 93% enriched U fuel elements. Birdcages 

read 0.8 mR/hr @ the surface.

12/3/1962 Smeared out 5 birdcages for shipment to Building [4064]. Smeared critical assembly and 

components. There were less than 30 dpm. The control drums and drives are going to Building 

[4010] and the core container is going to Building [4012].

1/22/1963 A complete smear survey of the building was taken. [Beryllium] smears of the SNAP 4 reactor, and 

tritium smears of the neutron-tritium target accelerator were taken. The results are listed below.*

Tritium Smears in cpm.**

1-A = 143, 2-A=103, 3-A=185, 4-A=Bkg, 5=A=Bkg, 6-A=Bkg

3/6/1963 A smear survey of Building [4373] revealed <30 dpm/100 cm

encountered from Bldg [4028] when these smears were being counted (1.83-30 dpm). Operations 

were notified and received their copy of the smear survey. Approx. 5 smears were counted on the 

manual counter to check the possibility that contamination exists. All smears indicated <30 dpm/

100 cm2 beta-gamma.

3/7/1963 A smear survey was conducted on two birdcages containing fuel cladding and fuel. The smear 

revealed the following:

I. The birdcage containing one fuel rod (no cladding)

Bottom of can, sides, and top: 125 dpm/100 cm

Fuel rod - 140 dpm/100 cm

II. The birdcage containing fuel cladding

4 fuel casings (cladding): 28 dpm/100 cm

5/1/1963 A smear survey of two birdcages containing 93% enriched fuel elements to be shipping to Building 

4064 revealed 1.3x10

revealed 80 dpm alpha inside and 9 dpm alpha outside. The operations crew was requested to 

decontaminate the cages on the outside prior to shipping.

6/24/1963 A smear survey was conducted by H&S on 10 ea. birdcages at Building 4373. The birdcages 

contained 10FS1 fuel elements. The birdcages were shipped to 4373 from HQ. Results 

included:***

Cage #1: Small metal chips were found on the outside of the birdcage in the bottom corner - 3,400 

dpm alpha and 5,000 dpm beta-gamma (approx.)

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11/11/1963 Took Be smear of reactor, inserts, and parts bins. Sent them to HQ, made a radiation survey of a 

10 min critical of a loading. The top of the reactor read 100 mR/hr, the center rod 120 mR/hr, the 

peripheral rods read 50 mR/hr apiece, the highest birdcage (with a load of 5 elements) read 2.5 

mR/hr at the surface. Smear results: no βγ activity. However, alpha contamination was found inside 

the B of E containers and 18 dpm alpha was found in the corner of the lattice on a birdcage. No 

other alpha smear was over 8 dpm. Bkg. was 0.1 cpm.

11/14/1963 Received verbal results of the Be smear survey conducted 11 Nov. 1 parts bin, 2 reflectors and the 

inserts were over allowable limits.

2/4/1964 A smear survey at the request of [redacted] of the reactor room following the removal of foils from 

the reactor yielded the following:

Reactor table - 4.4x10

Reactor room floor - 60 dpm/100 cm

Change room floor - 15 dpm/100 cm

Control room floor - 100 dpm/100 cm

[Redacted] was notified and a follow-up survey initiated. Immediate decontamination was 

requested. The request was complied with.

2/25/1964 A smear survey of Building 4373 recorded the following:

Reactor room and SS storage vault floor - ~10dpm/100 cm

Change room floor - 12 dpm/100 cm

Control room floor - ~4 dpm/100 cm

Hall floor - ~2 dpm/100 cm

3/11/1964 A smear survey of the reactor room, SS vaults, change room, and control room revealed 

contamination to be within permissible limits with the exception of one spot in the control room 

which was 2 dpm/100 cm

4/14/1964 Smear survey of reactor room, SS vault, change room, control room, and hall failed to reveal any 

significant α contamination.

7/31/1964 Smear survey of reactor room indicated 2x10

and adjacent area was <5 dpm/100 cm

Date Log Entry

* The log book did not present any results of the beryllium smears.

** The locations (i.e. 1-A, 2-A) corresponded to a position on the neutron-tritium target accelerator.

*** The research team did not provide the result of each of the birdcages measured and reported.

Bkg=background
 

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Section 6.1.2 

Building Number:

Building Alias: 

4012 

012 / T012 

Building Name: SNAP Generalized Critical Bldg. 

Building Function: 1962-1968 SNAP Critical Testing

Neutron Pulse Generator 

1970-1972 NASA Heavy Metal Reflected Fast Spectrum Reactor 

(HMRFSR)

1979 - ETEC Radiography & X-Ray 

Notes: DOE-NASA

Tritium Use & Contamination 

Beryllium (Be) Use & Contamination

NASA Nuclear Reactor: HMRFSR 

Radionuclides of Concern: americium (Am-241), beryllium (Be), cesium-137 (Cs-137),

plutonium isotopes (Pu-238, Pu-239, Pu-240, Pu- 241), strontium-90 (Sr-90), tritium, and natural 

and enriched uranium (U-234, U-235, U-238). 

6.1.2.1 Description of Operations & Processes 

Building 4012 was constructed in 1962 as the Systems for Nuclear Auxiliary Power (SNAP) 

Critical Test Facility. Building 4012 was located near the northern property line of Area IV, where 

the terrain near the boundary becomes very rugged and there is a 40-foot drop in elevation. The 

Building 4012 complex originally consisted of two sections connected by an enclosed walkway. 

The HSA provides a detailed description of the building’s features and schematics, however the 

Critical Cell (Room 110) was used to test SNAP critical assemblies and the Fuel Storage and 

Equipment Room (Room 109) was located adjacent to the critical cell. 

Operations began in 1962 with experiments using three SNAP critical assemblies (SCA-41, 

SCA-4B, SCA-5). Clad reactor fuel elements (U-ZrH) were stored in the fuel storage tubes 

located in Room 109. The SNAP critical experiments continued intermittently through 1968, 

when the fuel was shipped to Building 4064 (Source & Special Nuclear Material Storage Vault) 

and the facility was placed in “stand-by” mode. 

In 1970, critical experiments for the NASA-sponsored Heavy Metal Reflected Fast Spectrum 

Reactor (HMRFSR) began and were conducted until 1972. 

In 1979, a portion of the facility was modified for use by ETEC Quality Assurance in 

performance of X-ray machine and source radiography. The most major modification involved 

the removal of four fuel storage tubes and enclosure of the fuel storage room to serve as the 

radiographic darkroom. This use was terminated in 1992 and all radioactive sources were 

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transferred to the RMHF for storage. 1 In 1986, part of the Building 4012 complex was 

demolished to build Building 4228, the Sodium Component Test Installation (SCTI) Power Pak 

facility, on top of Building 4012. 

Contaminated items were generally placed in cardboard “ice cream” cartons, tagged for 

identification, and placed in collection bins outside the facility. 2 An estimated 1 gram of 

uranium-235 (U-235) was present for every cubic foot of trash. Periodically, the contents of this 

barrel were sent to the RMHF for burial or recovery.3 

The HSA provides a detailed description of radiological investigations and decontamination/

cleanup of releases at Building 4012. However, in 1973 fixed alpha contamination inside 

Building 4012 was painted over with eggshell-colored paint and stenciled, “Caution: Fixed Alpha 

Radioactive Material.” 

In 1985, a comprehensive radiological survey indicated the presence of alpha contamination in 

both Rooms 109 and 110. The Equipment Area in Room 109 exhibited alpha contamination at 

the entrance door (840-1,400 dpm/cm2), overhead light fixtures (2,800 dpm/cm2), air 

conditioning duct (840-2800 dpm/cm2), radioactive exhaust duct (4,200 dpm/cm2), and steel 

door frame between Rooms 109 and 110 (1,960 dpm/cm2). Spot checks of the concrete floor 

surface under the floor tile revealed contamination levels of 1,400-2800 dpm/cm2. 

Rockwell International adopted its own “acceptable limits” for the radiological investigation of 

buildings associated with SNAP, based on the use of enriched uranium for SNAP programs. 

According to Rockwell International, the ambient gamma exposure inside SNAP buildings was 

actually lower than outdoor natural background, and elevated rates inside the facilities were 

attributed to “primordial isotopes in the building material,” thus requiring little or no further 

investigation. Survey of the fuel storage area of Room 109 revealed contamination of the 

concrete floor (up to 6,500 dpm/cm2). Survey of the fuel storage tubes indicated contamination 

levels up to 6,000 dpm/cm2. [It should be noted that general SSFL Health Physics standards for 

safety placed a limit of 30 dpm/cm2 on surface contamination]. 

Rockwell International adopted its own “acceptable limits” for the radiological investigation of 

buildings associated with the SNAP program, based on enriched uranium used for SNAP

programs. According to Rockwell, the ambient gamma exposure inside SNAP Criticality 

buildings was actually lower than outdoor natural background, and elevated rates were 

attributed to “primordial isotopes in the building material,” thus requiring no further investigation. 

The MARSSIM preliminary classification for Building 4012 is Class 1 because of its former use. 

1 EPA/HGL, Final HSA 5B, p. 26. citation 7:

7 

Pascolla, A.L., Decontamination and Decommissioning of Building T012, ETEC No. 012-AR-0001, Boeing

North American Inc., Rocketdyne Division, May 8, 1997, pgs. 3-4, 10-11. 

2 EPA/HGL, Final HSA 5B, p. 30. citation 5:

5 Gavigan, F. X. et al., Reactor Safety Survey Report, SNAP Critical Assembly -5 (SCA-5), Building 012, 

February 23, 1965, pgs. 10-11. 

3 EPA/HGL, Final HSA 5B, p. 30. citation 6:

6 Internal Correspondence from Heneveld, W.H. to Schaubert, V.J., North American Rockwell, Reference: 

SS Material Control for Bldg 012, March 30, 1971, pgs. 3-4. 

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6.1.2.2 Building 4012 Radiological Incident Reports 

The EPA HSA research team did not locate any incident reports pertaining to Building 4012 in 

Boeing’s Incident Reports Database. However, CORE Advocacy located the SSFL Health 

Physics Department Log Book for Building 4012, which documented operations and incidents at 

Building 4012 between October 12, 1962 -July 27, 1966.1 

Incidents reflected in the Building 4012 Log Book are summarized below. The Log Book does 

not specifically reference isotopes involved in each incident. Clues of context and building 

operations suggest that uranium was the primary isotope of concern. 

The Log Book describes several incidents over the course of months that involved contaminated 

“birdcages” containing broken pieces of SNAP 8 fuel and highly irradiated metal shavings. 

CORE Advocacy notes that there are numerous references throughout incident reports 

associated with many facilities at SSFL, which document receipt of “contaminated birdcages” in 

similar condition, throughout this era of site operations. Details contained in the Building 4012 

Log Book and other incident reports associated with SSFL Area IV facilities suggest the 

contaminated birdcages were repeatedly circulated throughout the facility. By all indications, the 

problem of “contaminated birdcages” presented quite a phenomenon at SSFL over the course of 

several months. 

Tritium use and contamination resulting from the neutron pulse generator are referenced at 

length. Beryllium use and contamination are also noted. After 1963, the Log Book entries 

become sporadic, and there are only eight entries between July 1964 and August 1966. This 

could suggest a lapse in consistent record keeping. However, given the frequency of noted 

events and contamination of Building 4012, it is reasonable to assume that daily operations 

resulted in several incidents that were not documented, and/or where corresponding 

documentation has been lost or destroyed. 

Measurements in the table below are written as indicated in the Log Book. CORE Advocacy 

recommends a thorough review of the Log Book to ensure accurate interpretation, and addition 

of the Log Book to the Incidents Database. 

6.1.2.3 Building 4012 Health Physics Log Book Entries 

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

10/12/1962 - 

10/15/1962

4012 Uranium Contaminated “birdcages” of new fuel

10/19/1962 4012 Unknown Smear survey results taken from the floor and reactor vault door indicate 

contamination. 

10/24/1962 4012 MFP Map-5 fixed filter air monitor alarmed. Vacuum pump “very hot.” Air 

samples in the vault were 1.3x10

alpha. Alarm sounded 5 minutes after reactor shutdown.

1 Atomics International Health Physics Log Book, Building 4012, 10/12/62-7/27/66 File: HDMSP001852993.pdf 

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11/2/1962 4012 Air sample in reactor vault 2.8 x 10

11/12/1962 4012 Results of immersion water sample 2.08/10

11/12/1962 4012 Result of routine facility smear 341 dpm/100 cm

critical cell room.

12/18/1962 4012 Uranium 10 Birdcage smears revealed dose rate from each birdcage to be 2.5-3 

mR/hr. 

12/18/1962 4012 Fuel 

Storage 

Vault

Uranium Broken pieces of “hot” SNAP 8 fuel found loose/uncontained. Smallest 

piece counted 257 dpm alpha.

12/18/1962 4012 Fuel 

Storage 

Vault

Uranium More broken fuel found in birdcages (“slag”). Counts ranged between 

159 dpm alpha - 1394 dpm alpha. Top of the neutron source can 

measured 44 dpm beta-gamma “due to laying pieces of U on top.”

12/18/1962 4012 Fuel 

Storage 

Vault

Uranium Numerous smears taken of contaminated birdcages. All tested high. A 

planchet of residue found on birdcages measured 27,000 dpm / 100 cm

beta-gamma. Alpha activity was not counted “because of the possibility 

of contaminating the chamber.”

12/21/1962 4012 Fuel 

Storage 

Vault

Uranium 4 birdcages indicated surface measurements of 2.5 mR/hr beta-gamma.

1/7/1963 4012 

Equipment 

Room

Smear detected 50 dpm/100 cm

1/9/1963 4012 

Equipment 

Room

Smear detected 40 dpm/100 cm2

1/18/1963 4012 Fire in the air conditioning and furnace unit.

1/29/1963 4012 Smear of 75 dpm/100 cm

2/9/1963 4012 Smear 90 dpm/100cm

2/11/1963 4012 

Exhaust

Meter survey of exhaust filters revealed 

.08 mR/ht at surface. Smears were taken after filters were changed.

2/25/1963 4012 Pu-Be Pu-Be source used during mock-up loading of S8ER. Neutron survey 

revealed 6.5 mrem/hr 

2/26/1963 4010 to 

4012

Transfer of the mock-up S8ER reactor core from Bldg 4012 to 4010. 

“Motion pictures were taken to illustrate the method that will be used to 

transfer the core from Bldg. 4012 to 4010.”

3/8/1963 Tritium New tritium target installed in the neutron pulse generator. Tritium 

contamination noted on neutron generator. Employees observed working 

on neutron generator without gloves or lab coats. Employee smears 

revealed 2x10

immediate bioassay.

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

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3/11/1963 4012 Tritium Personnel notified to mop floors and clean all surfaces possibly 

contaminated with tritium on 3/8/1963.

3/15/1963 4012 

Radioactive 

Liquid 

Holdup Tank

Attempt to sample R/A liquid hold up tank revealed no provisions for 

sampling the system; no pump system for pumping R/A liquid out or to 

circulate the water. Tank was 3/4 full at the time of the discovery. The 

only drain in the tank went to sewage system via a valve and gravity 

drain. Recommendations made to address ASAP.

3/18/1963 4012 Tritium Smear of neutron generator revealed 120 dpm/100 cm

tritium survey could not be conducted at the time because the gas 

proportional counter was being repaired at the Instrumentation Bldg.

3/26/1963 4012 

Radioactive 

Liquid 

Holdup Tank

MFP Water sample obtained from R/A Liq. Holdup Tank. Metal pipe inserted to 

stir the water to obtain sample, which revealed 1.14 x 10-6 μc/cc. 

Dumped 500 gallons. No indication of where.

4/8/1963 4012 Fuel 

Storage 

Vault and 

Critical Cell

Uranium 15 birdcages found containing broken fuel fragments and shavings. 

Brought to 4024 (location of gas proportional counter). Revealed 5.5 x 

105 dpm alpha and 6.0 x 10

of birdcages revealed 200 dpm alpha on smear paper. A GM tube placed 

over black powder revealed .07 mR/hr beta-gamma at 2 inches. Fuel 

Vault and Critical Cell area red-tagged until complete smear survey could 

be conducted.

4/9/1963 4012 Fuel 

Storage 

Vault and 

Equipment 

Room

Unknown During birdcage smears smoke was observed and the air smelled like 

something was burning. Investigation by fire department declared 

building was safe to enter; probably a transformer that shorted out.

4/10/1963 - 

4/15/1963

4012 Fuel 

Storage 

Vault & 

Equipment 

Room

Uranium Investigation into contaminated birdcages (17) containing metal shavings 

and fuel fragments. Metal shavings and fuel fragments measured 

between 1.0x10

contamination and sampling are documented; continued problems with 

broken fuel and metal shavings found in the bottoms of birdcages 

delivered to SNAP.

4/22/1963 4012 Critical 

Cell

Air samples revealed 6x10

5/31/1963 4012 

Reactor 

Dunk Tank

Water sample revealed 7.9 uuc/liter beta-gamma.

7/31/1963 4012 5 Ice Cream Cartons of R/A compressible waste measured .05 mR/hr at 

the surface.

8/2/1963 4012 

Reactor 

Dunk Tank

Water sample was 6.9x10

to be disposed due to completion of the SCA-4 Water Emersion Tests.

11/14/63 - 

11/19/63

4012 Room 

110

Be Beryllium box containing approximately 65 lbs of Be pars ranged from 

0.1-1.0 ugm and 1.0-30 μgm.

12/10/1963 4012 11 ice cream cartons of R/A waste measured .15 mR/hr @ surface.

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

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1/23/1964 4012 R/A liquid holdup tank sample 2.2x10

released. 3.3x10

“released” refers to.

2/4/1964 4012 Smears of bldg. 4012 revealed 2x10

instrument repair room on the gamma spectrometer. Overage of 

4x102dpm/100 cm

change room. 

8/3/1964 4012 Fuel 

Storage 

Vault

93% EU .328 gr. of 93% enriched U spilled. Smears revealed no spread of 

contamination; no personnel contaminated. Maximum of 585 dpm/

100cm2 alpha.

2/2/1965 4012 Critical 

Cell

Reactor approx 50 watts max - 10 watts min. for 20 minutes. Sample 

volume 1.5x103 [illegible]. Activity was 8.6x10-10 μc/cc beta-gamma.

2/4/1965 4012 Fuel 

Storage 

Vault

R/A Ice Cream cartons of R/A compress. waste. all cartons indicated 51 

mR/hr beta-gamma.

3/30/1965 4012 Critical 

Cell and 

Fuel Vault

Routine smear survey of facility. Dose rate survey of reactor assembly 

indicated 500 mR/hr gamma at 1” inch. Alpha survey of Fuel Vault 

revealed fume hood with 2,000 dpm on ledge and 10,000 dpm inside.

7/22/1965 4012 Test 

Vault

Smear survey revealed 105 dpm/100 cm

area.

7/27/1966 4012 Smear survey results of 800 dpm beta-gamma including 588 dpm/100 

cm2 on hood apron. Max on floor was 229 dpm/100 cm

100 cm2 alpha, beta, gamma.

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

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Section 6.1.3 

Building Number:

Building Alias: 

4019 

019 / T019 

Building Name: SNAP Flight Systems Nuclear Qualification Tests (SNAP-10A)

Fissile Material Storage

ETEC Construction Staging & Computer Facility 

Building Function: Nuclear Qualification & Vacuum Tests / Criticality Acceptance 

Tests / Fissile Material Storage 

Notes: DOE-NASA

Possible Tritium 

Associated Buildings: 4719 Electrical Substation 

Radionuclides of Concern: Am-241, Cs-137, H-3 Pu-238, Pu-239, Pu-240, Pu-241, Sr-90, 

U-234, U-235, U-238, 

6.1.3.1 Description of Operations & Processes 

Building 4019 was constructed in 1962 as the Systems for Nuclear Auxiliary Power (SNAP) 

Flight Systems Nuclear Qualification Test Building, located near the northwest property line of 

Area IV. Building 4019 is located between SNAP Buildings 4013 and 4059. 

The HSA provides a detailed description of Building 4019’s features and schematics. It 

contained a high bay and low-bay office-control center, a below-grade test cell, and a vacuum 

test chamber. The low bay contained the control room, personnel change room, equipment 

room, and offices. It was originally built with cinderblock walls for nuclear fuel element storage. 

Building 4019 was built to conduct criticality acceptance tests of SNAP reactors before they 

were delivered to the AEC for launch as space-power systems. The criticality tests of the 

SNAP-10 Flight System 3 (S10FS3) reactor were conducted in Building 4019 in 1963 before the 

reactor was operated in the SNAP Environmental Test Facility (Building 4024).1 

In 1964 and 1965, three SNAP 10-A flight systems (FS-1, FS-4 and FS-5) were assembled and 

tested in Building 4019. Boeing indicated that all nuclear materials were removed from the 

building when the last SNAP reactor was removed in 1965, and the facility was used for “other 

1 EPA/HGL, Final HSA 5B, p. 40. citation 6:

6 The Boeing Company, Rocketdyne Environmental Affairs, Building 4019 – SNAP Flight System Critical 

Facility, January 8, 2003, p. 1. 

1 EPA/HGL, Final HSA 5B, p. 40. citation 8:

8 The Boeing Company, Rocketdyne Environmental Affairs, Building 4019 – SNAP Flight System Critical 

Facility, January 8, 2003, p. 1. 

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purposes” during the 1970’s and 1980’s.1 However, a 1968 letter indicates that fissile materials 

were being stored in Building 4019.2 

According to the U.S. Nuclear Regulatory Commission (NRC), fissile material is any “nuclide 

that is capable of undergoing fission after capturing low-energy thermal (slow) neutrons.” This 

current definition means the “three primary fissile materials are uranium-233, uranium-235, and 

plutonium-239. This definition excludes natural and depleted uranium that has not been 

irradiated, or has only been irradiated in thermal reactors.” 3 Upon termination of the SNAP 

program in 1970, all SNAP components were removed. Building 4019 was re-designated the 

ETEC Construction Staging and Computer Facility.4 

Rockwell International adopted its own “acceptable limits” for the radiological investigation of 

buildings associated with the SNAP program, based on enriched uranium used for SNAP

programs. According to Rockwell, the ambient gamma exposure inside SNAP Criticality 

buildings was actually lower than outdoor natural background, and elevated rates were 

attributed to “primordial isotopes in the building material,” thus requiring no further investigation. 

In 1995, ORISE performed verification survey activities to validate cleanup procedures and 

survey methods used by Rockwell-Rocketdyne. The verification survey found residual betagamma surface contamination in excess of U.S. DOE guidelines in a 1-square meter area of the 

Building 4019 high bay. The average beta-gamma total surface activity was 5,900 dpm/100 cm2,

exceeding the DOE limit of 5000 dpm/100 cm2. This result, combined with other deficiencies 

(including lack of access to the Building 4019 vault and inadequate documentation of previous 

surveys), led ORISE to recommend additional decontamination and surveys. 

The HSA provides a detailed chronology of radiological investigations and decontamination / 

cleanup for release surveys associated with Building 4019. 

The preliminary MARSSIM classification for Building 4019 is Class 1 due to former use. 

2 EPA/HGL, Final HSA 5B, p. 40. citation 9:

9 Internal correspondence from Bunch, D.F. to Alexander, R.E., Atomics International, a Division of North 

American Aviation, Inc., Reference: Emergency Preparedness Meeting with SAN Personnel, dated January

25, 1968. 

3 EPA/HGL, Final HSA 5B, p. 40. citation 1:

1 “Fissile Material,” U.S. Nuclear Regulatory Commission, http://www.nrc.gov/reading-rm/basic- ref/glossary/

fissile-material.html (August 2, 2010). 

4 EPA/HGL, Final HSA 5B, p. 40. citation 2:

2 

Liddy, P., Building 4019 Final Status Survey Report, The Boeing Company, June 10, 1999, p. 7. 

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6.1.3.2 Building 4019 Radiological Incident Reports 

There has been one incident associated with Building 4019 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing. Summaries of all available incident reports are 

provided in Attachment A. 

6.1.3.3 Building 4019 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0304 4/10/1976 LARGE TANK Ir-92 Stuck gammagraph source uncoupled by 

unbadged operator.

In addition to the incidents listed in Boeing’s Incident Report Database, the following sodium 

potassium (NaK) incidents were noted in a 1964 Log Book. 

• On June 23, 1964 a NaK leak detector alarm went off while testing SNAP 10AFS1. The 

following day, NaK was noticed inside the vacuum chamber, with the highest concentration of 

NaK around the instrument compartment. On June 25, 1964 the “vacuum chamber center 

section was raised above floor level ... to observe NaK from the spill.” The next day, the SNAP

10 FS1 unit was out of the vacuum center section, approximately 6 feet above the high bay 

floor. People were taking pictures, but only one was wearing the proper NaK protective 

clothing.1

• On June 29, 1964 a NaK fire was observed inside the A.P.V. of Building 4019. [APV may stand 

for Advanced Propulsion Valve or Vessel]. According to a log book, “the NaK fire looked quite 

large and close to out of control. 3 or 4 metal-X extinguishers were emptied before control was 

attained.” A large amount of NaK was pouring out of the instrument compartment and a 

fireman on duty assumed it was coming from the “Rx coolant system.” The expansion 

compensator, which was discovered to be the source of the NaK leak, was cut out of the Rx 

coolant system. The following day was spent cleaning up from the fire. 2

1 EPA/HGL, Final HSA 5B, p. 43. citation 2:

2 

Unknown Author, Log Book, Building 4019, 1/20/64 to 6/3/65, January 20, 1964,

HDMSP001853466- HDMSP001853468. 

2 EPA/HGL, Final HSA 5B, p. 43. citation 3:

2 

Unknown Author, Log Book, Building 4019, 1/20/64 to 6/3/65, January 20, 1964,

HDMSP001853466- HDMSP001853468. 

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Page 182 of 418 

Section 6.1.4 

Building Number: 4024 

Building Alias: 024 / T024 

Building Name / Function: SNAP 2 Development Reactor (S2DR)

SNAP 10-A Flight Simulation Reactor (S10FS1)

SNAP 10-A Flight Simulation Reactor (S10FS3)

SNAP 10-A Flight Simulation Reactor (S10FS5) Storage

SNAP Transient Reactor (SNAPTRANS-1)

Liquid Metal Fast Breeder Reactor (LMFBR) Support

Clinch River Mock-up Reactor Operations

Storage, D&D of Hot Lab (Bldg. 4020) Equipment 

Notes:	 DOE-NASA

Tritium 

Radionuclides of Concern: The primary radiological constituents of concern have been 

identified as Co-60 and Eu-152 as a result of activation in the concrete and piping of the 

subterranean test vaults. Secondary radiological constituents of concern at Building 4024 

include Cs-137 and Sr-90 (fission products); H-3, Eu-154, Fe-55, Ni-59, Ni-63, Mn-54, 

potassium-40 (K-40), and sodium-22 (Na-22) (neutron activation products); Th-232, U-234, 

U-235, and U-238 (nuclear fuel material); and Am-241, Pu-238, Pu-239, Pu-240, Pu-241, and 

Pu-242 (transuranic elements).1 

6.1.4.1 Description of Operations & Processes 

Building 4024 was constructed in 1959 and used for testing SNAP reactors in a simulated space 

environment. The sub-grade test cell complex comprising Building 4024 consisted of three 

parallel cells (two power-test cells and a center transfer cell), a partial rear corridor that 

interconnected the cells, and the operating gallery. A detailed description of Building 4024’s 

features and measurements is provided in the EPA-HSA. 

The SNAP 2 Developmental reactor prototype was tested at a nominal power level of 65 kWt 

without any power conversion system equipment. It was tested in Vault No. 1 of Building 4024. 

The SNAP 10-A Flight System 3 (S10FS3) reactor was tested in a different vault. The SNAP

Transient (SNAPTRAN-1) reactor criticality tests occurred at Building 4024, as well as tasks 

related to the Liquid Metal Fast Breeder Reactor (LMFBR) and Clinch River Mock-Up reactor. 

During later years of site operations, Building 4024 was used for the storage and 

decontamination tasks associated with Building 4020 Hot Laboratory. 

1 EPA/HGL, Final HSA 5-A, p. 42, citations 167-169:

167 The Boeing Company, “Building 4024 Decontamination and Decommissioning Engineering Evaluation/

Cost Analysis,” May 1, 2007.

168 Tuttle, R.J., Listing of Locations in SSFL Area IV Associated With Radioactive Materials,” September 

1989.

169 The Boeing Company, Engineering Evaluation & Cost Analysis (EE/CA) Community Meeting Microsoft 

PowerPoint® Presentation, February 21, 2007. 

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Building 4024 was modified in 1962. An additional control room was added, along with a new 

room (“Vault Number 2”), which contained vacuum pumps and other heavy equipment 

associated with the SNAP 10-A Space Nuclear Reactor (S10FS-1 / S10FS-3). The vacuum 

chamber allowed for a simulation of the operational environment in outer space.1 

Power Test Vault Number 1 housed a SNAP-2 Demonstration (a.k.a. “Developmental”) Reactor 

(S2DR) that operated for 5,000 hours at 30 and 50 thermal kilowatts (kWt) between April 1961 - 

December 1962. The S2DR generated 13 MWd of power and contained 390 x 103 Ci of 

radioactivity at the end of operation. The Hot Lab (Building 4020) was used to examine the fuel 

and components from the reactor following the operation. The EPA research team could not find 

additional historical information regarding the operation of the S2DR in Building 4024. 

According to a 1962 document, Power Test Vault Number 2 was scheduled to provide shielding 

and containment for the SNAP 10-A Flight System (S10A-FS-1) in a vacuum chamber located 

within the test vault. The vacuum chamber was designed to simulate the space environment, 

and the reactor was put through various paces including shock and vibration testing that 

simulated space-launch. The reactor system test was scheduled to begin in late 1962 and would 

be completed and removed some time in 1963. The EPA research team did not find 

documentation indicating that the S10A-FS-1 was tested in Building 4024, but the EPA research 

team did locate documentation that the S10A-FS-3 was tested there.2 

CORE Advocacy located an Atomics International document3 that cautioned against exceeding 

specifications limits due to the rate of temperature rise, describing that the results of the 

intended testing were not exactly known. The letter emphasized that the rate of the heat rise

would be the governing factor, rather than the ultimate temperature (provided that the 

specification limits were not exceeded). Part of the testing involved exposing the reactor to 

thermal shock and extreme temperature fluctuations, but Atomics International stressed that the 

effect on the device could not be accurately predicted; according to Atomics International, the 

thermal shock was greatest during the first few seconds of energizing at room temperature 

conditions when full voltage to the reactor was applied. The heat (watts) generated was almost 

three times the amount that it was designed for (at 800 degrees Farenheit). 

1 EPA/HGL, Final HSA 5-A, p. 25, citation 84:

84 Atomics International, Document NAA-SR-7300 Special, “Evaluation of the Atomics International Nuclear 

Development Field Laboratory as a Location for Reactor Facilities,” May 25, 1962. 

2 EPA/HGL, Final HSA 5-A, p. 26, citations 89-93: 

89 Rockwell International, Document N001ER000017, Rev. C.,Nuclear Operations at Rockwell’s Santa 

Susana Field Laboratory – A Factual Perspective, May 30, 1991.

90 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005

91 Clark, R., and Potter, G., SEC Petition Evaluation Report Summary: SEC-00093, Santa Susana Field 

Laboratory- Area IV, April 28, 2009.

92 Atomics International Document, N704FDP990006 Rev. A., “Building T024 (SETF) Facilities Dismantling 

Plan,” July 31, 1981.

93 Atomics International, Document NAA-SR-7300 Special, “Evaluation of the Atomics International Nuclear 

Development Field Laboratory as a Location for Reactor Facilities,” May 25, 1962. 

3 J. Susnir / I. Rowe, Atomics International Interoffice Letter, “Duty Cycle SNAP 10-A Actuators, Room Ambient 

Conditions,” May 20, 1963. HDMSP001906220.pdf 

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Atomics International warned that continuous duty for approximately one minute, just once, may 

or may not produce deleterious effects but the repetition of this would probably have a harmful 

effect, and the degree was uncertain. The author of the letter (I. Rowe, Supervisor of the 

Electro-Magnetic Devices Component Development Group in the Reactor Development

Department), stated, “I hesitate to give consent to any unregulated operations for an extended 

time and recommend constant current control for these operations. However, if this is 

impossible, then I would reluctantly and with great reservation consent to not more than any one 

minute of unregulated energizing and more important, limit this to three times during the life of 

the actuator. Even so, the effects on the unit are indeterminate. A failure later on due to stresses 

caused is entirely feasible.” 

The prototype SNAP 10-A Flight System (S10FS-3) reactor operated for 10,000 hours at about 

40 kWt between January 1965 and March 1966 and set the record for the longest continuous 

operation of a nuclear reactor system with 10,000 hours of uninterrupted operation. 

On April 27, 1966 a preliminary draft radiation engineering analysis report by the Radiation 

Safety Unit of Atomics International presented information on the removal of the S10FS-3 

reactor core, coolant pump and associated piping from the building. The materials were to be 

transported to the Hot Laboratory, Building 4020. At the time of removal, the reactor was to have 

been shut down for a period of 45 days. The EPA HSA provides a detailed account of the 

sequential removal of the items but the EPA research team was unable to locate documentation 

indicating the route that may have been followed during the transfer to Building 4020, which 

may have put workers at other locations at risk of exposure.1 

On July 14, 1966, an internal letter reported that the building 4024 cells were being

decontaminated following the conclusions of the SNAP10FS-3 ground tests. At the time of the 

letter there was no immediate work planned for the cells, and the letter therefore requested the 

maximum permissible contamination levels the cells could be left in, on standby status.2 Such 

maximum permissible contamination levels should be taken into consideration when estimating 

worker dose at Building 4024, because the building remained functional and in use between 

1966 and 1976. 

A November 1976 document outlined the working procedures to be used to remove the vacuum 

vessel from the west cell of Building 4024. The document, in advance of a general radiological 

1 EPA/HGL, Final HSA 5-A, p. 27, citations 94-95: 

94 Heine, W.F., Atomics International, Preliminary Draft Radiation Engineering Analysis: Radiation Safety 

Analysis, April 27, 1966. HDMSP001852876.

95 Lords, R.E., Westinghouse Idaho Nuclear Company, Inc., SNAP and AI Fuel Summary Report, Report

WINCO- 1222, August 1994. 

2 EPA/HGL, Final HSA 5-A, p. 27, citation 96: 

. 96 Wilson R.F., Atomics International Letter Re: Building 024 – FS-3 Vault and Transfer Cell Contamination 

Levels, July 14, 1966. 

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survey of the cells, indicated that the vacuum vessel was slightly activated. While the document 

outlined the dismantling procedure, it did not indicate the date of final removal.1 

SNAP Critical Assembly 4B operated in the east cell for a short time at low power. SNAP 

Transient Test (SNAPTRAN-1) support reactor, also critical, operated in the east cell for a short 

time. Typically SNAPTRAN-1 was operated at low power, except for some pulsed operation. 

This reactor last operated in 1971, and it was the last reactor to be tested in Building 4024.2 

According to an undated Rocketdyne Environmental Affairs document, following termination of 

each project, all equipment and fuel was removed from the facility. The document does not 

indicate where all the equipment and fuel were taken upon removal; however, an August 1994 

SNAP and Atomics International Fuel Summary Report states that within a few months of 

shutdown, reactors were disassembled (either on site or after having been transported whole to 

the Hot Laboratory in Building 4020), and select elements were examined in extensive postirradiation studies at Building 4020. Following examination, the fuels were sent to the Idaho 

Chemical Processing Plant (ICPP). As of 1994, the fuels remained in underwater storage at the 

ICPP. EPA’s HSA research team could not locate additional historical information for the 

termination of each project.3 

On January 30, 1973 permission was granted in an internal letter for the storage of Building 

4019’s SNAP 10A FS-5 system in Building 4024.4 A document drafted December 13, 1972 

outlined the procedures for the transfer of the reactor from Building 4019 to Building 4024. 

According to the document, the SNAP 10A FS-5 was a complete power system that was loaded 

with fuel and NaK, and acceptance tested. The fuel was identified as 93% enriched U-235 with 

a total mass of 5 kilograms of material, 4.75 kilograms compromising U-235. 

The system had been stored in Building 4019, and was to be removed and transferred to 

Building 4024 for permanent storage in 1973, via the south end of Building 4019 to the east rollup door of the high bay of Building 4024.5 The permission letter reported that a nuclear safety 

1 EPA/HGL, Final HSA 5-A, p. 28, citation 97:

97 Rockwell International Document, N704DWP990024, “Dismantling and Removal of 10FS3 (SNAP)

Vacuum Vessel from Building T024, Detailed Working Procedure,” November 16, 1976. 

2 EPA/HGL, Final HSA 5-A, p. 28, citation 98:

98 Atomics International Document, N704FDP990006 Rev. A., “Building T024 (SETF) Facilities Dismantling 

Plan,” July 31, 1981. 

3 EPA/HGL, Final HSA 5-A, p. 28, citations 99 & 100:

99 Rocketdyne Environmental Affairs, Building 4024 – SNAP Environmental Test Facility, Undated.

100 Lords, R.E., Westinghouse Idaho Nuclear Company, Inc., SNAP and AI Fuel Summary Report, Report

WINCO- 1222, August 1994. 

4 EPA/HGL, Final HSA 5-A, p. 28, citation 101:

101 Ketzlach, N., Internal Letter Re: New MBA for SNAP 10FS-5 Storage, January 30, 1973. 

HDMSP001840067. 

5 EPA/HGL, Final HSA 5-A, p. 28, citation 102:

102 Heneveld, W.H., NSA-652-160-001 Transfer of SNAP 10A FS-5 from Bldg 019 to Bldg 024, December 

13, 1972, HDMSP001856104. 

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analysis (NSA) should be prepared for handling the elements outside the reactor and should be 

reviewed by the Fuels Committee. However, information regarding the final transfer of the 

system to, and the removal of the system from, Building 4024 could not be located.1 

According to a December 1976 radiological survey plan for Building 4024, the Building 4024 test 

vaults became radioactively contaminated during the dismantling of the reactors upon

completion of operations. The plan stated that the test vaults were lined with aluminum and 

were decontaminated by site personnel to safe limits of less than 50 dpm. Radiation surveys 

performed after decontamination of the test vaults indicated activation of the vault walls and 

equipment within the vaults, including TV camera traverse rails, vault light fixtures, and the 

environmental test vacuum vessel for S10FS3. In addition, beryllium metal contamination was 

also detected to be present during the SNAPTRAN reactor operations. According to the plan, 

the beryllium contamination resulted from the use of bare (not anodized) beryllium metal plates 

as neutron reflectors for the SNAPTRAN. The plan stated the beryllium contamination was 

2

cleaned up to less than 0.1 μg/100 cm following SNAPTRAN operations.2 

According to the 1976 plan, no significant radioactive contamination was ever detected in the 

radioactive liquid waste retention tanks; radioactive gaseous waste retention tanks contained 

atmosphere from the test vaults. Rockwell International stated in the plan that analysis of the 

gas within the tanks identified only argon-41, which has a decay half-life of 1.83 hours. Rockwell 

International indicated that no other long-lived isotopes were reported to have been detected 

during reactor operations. 3 However, it bears noting that several of the assertions made by 

Rockwell International border on the absurd, and directly contradict the standardized regulations 

of the time. (For instance, according to a report authored by Rockwell International, a reading of 

7,500 dpm / 100 cpm2 was considered “acceptable” for surface contamination of beta and 

gamma emitters. This report is discussed in detail, below). 

Following the termination of the SNAP program, Building 4024 was used intermittently on 

different Liquid Metal Fast Breeder Reactor (LMFBR) tasks. In 1987, the most recent program to 

utilize the building was the testing of a mockup of the Clinch River fuel handling systems. SSFL 4

worker records indicate that Rocketdyne employees previously thought to be confined within 

SSFL Areas I, II and III routinely performed Clinch River-related job tasks at Building 4024.5 

There are also indications that the Space Shuttle Main Engine (SSME, NASA) may have been 

1 EPA/HGL, Final HSA 5-A, p. 28, citation 103:

103 Ketzlach, N., Internal Letter Re: New MBA for SNAP 10FS-5 Storage, January 30, 1973. HDMSP001840067.
 

2 EPA/HGL, Final HSA 5-A, p. 28, citation 104:

104 Rockwell International Document, N704TP99009, “Radiological Survey Plan, Support of D&D Program

Operations – T-024 (SNAP 2 and 10), December 9, 1976. 

3 EPA/HGL, Final HSA 5-A, p. 29, citation 105:

105 Rockwell International Document, N704TP99009, “Radiological Survey Plan, Support of D&D Program

Operations – T-024 (SNAP 2 and 10), December 9, 1976. 

4 EPA/HGL, Final HSA 5-A, p. 29, citation 108: 

108 ETEC, Site Consolidation Assessment, April 16, 1987. p. 41. 

5 Worker Records provided upon request, redacted unless authorized to review. 

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developed in the Building 4024 High Bay, which involved Rocketdyne employees with Area I, II 

or III “Time Clock Locations” at SSFL. 

3.4.5.1 Radiation Surveys & Surveillance - Building 4024 

Progress reports for maintenance and surveillance of facilities in 1987 show that monthly 

maintenance and surveillance activities included radiation surveys and groundwater sampling. 

In April 1987, it was reported in the progress report that no significant removable activity was 

found during the monthly radiation survey, and that sampling of water showed to radioactivity 

above background. However, in January 1987 samples of water from the pipe chase room at 

Building 4024 showed low levels of Co-60. It is important to note the monthly report did not 

provide information to indicate what other radionuclides were included in the sampling program, 

how frequently samples were taken, or what the reportable levels were.1 In 1989, DOE’s Office 

of Environmental Audit indicated insufficient monitoring and sampling practices at SSFL.2 

Weekly surveillance and maintenance activities were conducted in 1988 and 1989. According 3

to a 1989 factual perspective of the SSFL facilities, Building 4024 contained approximately 15 

mCi of confined activation radioactivity in concrete. Because of the design of the facility, the cost 

of D&D of the facility was determined to be expensive. According to the report, “the confined 

radioactivity is decaying and will meet release criteria when the SSFL is released for 

unrestricted use.” Accordingly, the report stated Rockwell would continue surveillance of the 

facility in the interim. 

According to Boeing, between 1997-1999 Building 4024 was used as a staging and 

decontamination area for the Hot Lab (Building 4020) concrete blocks. Weekly monitoring 

measured airborne releases to the environment. Monitoring included, but was not limited to, the 

following radionuclides that were determined to possibly be present in the process of 

decontaminating the blocks: Sr-90, Cs-137, Th-230, U-234, U-235, U- 238, Pu-238, Pu-239, 

Pu-240, and Am-241. 

As of 2011, D&D of Building 4024 was incomplete. A chronology of radiological investigations at 

Building 4024 follows: 

• A radiation survey was carried out in test cell 2 of Building 4024 on April 1, 1966, to establish 

the dose rates surrounding the S10FS-3 reactor core. The purpose of the survey was to provide 

measured dose rate data for comparison with calculated dose rate data prior to the removal of 

the test cell shield blocks. At the time of the survey, the reactor had been shut down for a period 

of 15 days. The data were obtained by taking measurements in 1-foot increments by lowering 

the detection instrument into the #3 reactor instrument thimble. The radiation dose rates various 

1 EPA/HGL, Final HSA 5-A, p. 30, citations 110 & 111:

110 Wieseneck, H.C., ETEC Monthly Progress Report-April 1987, May 20, 1987.

111 Wieseneck, H.C., ETEC Monthly Progress Report-January 1987, February 20, 1987. 

2 U.S. DOE Office of Environmental Audit, “Environmental Survey Preliminary Report, U.S. DOE Activities at SSFL,” 

1989. 

3 EPA/HGL, Final HSA 5-A, p. 30, citation 112:

112 Gaylord, G.G., SFMP Weekly Reports, 1988 through 1989. 

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from 24.0 R/hr 8 feet lowered from the high bay floor, to 87.0 R/hr at the midline of the reactor 

core 15 feet lowered from the high bay floor, to 36.0 R/hr 21 feet lowered from the high bay 

floor, at the floor level of reactor test cell 2.1 

• A remote radiation survey of the transverse corridor of test cell 2 was performed on April 30, 

1966. The purpose of the survey was to evaluate the radiation levels at the entrance to the 

transverse corridor with door B-104 open. The levels at three additional points were also 

evaluated. At the four points where measurements were taken, the radiation intensity varied 

from .055 R/hr to 22.5 R/hr. The survey concluded that operations personnel could enter the 

transfer lock with door B-104 open, if necessary.2 

• On May 3, 1966, an Atomics International internal letter provided the results of a second 

radiation survey of SNAP 10FS-3, located in test cell 2 of building 4024. The survey, 

performed on April 30, 1966, measured radiation of the reactor after the reactor had been shut 

down for a period of 45 days. The data were obtained by lowering a Victoreen Radacon Model 

510 detector into the #3 reactor instrument thimble, approximately 8 feet from the axis of the 

reactor core. The results measured that, if the reactor core is assumed to be the point source, 

the maximum dose rate 1 foot from the core was approximated to be 3.07 x 103 R/hr. If the

source geometry was assumed to be a 1 foot line source when the detector is located at the 

midline of the core, the dose rate was approximated to be 2.6 x 103 R/hr at 1 foot from the 

core.3 The letter did not provide information to indicate the purpose of the survey; however, it 

should be noted that the tests on the reactor ceased in March 1966 and decontamination of 

the cells had begun by July 14, 1966.4 It should also be reiterated that maximum permissible 

contamination levels were sought, in order to leave the cells in “standby” status.5

• A February 11, 1977 internal letter from R.K. Owen to W.F. Heine of Rockwell International 

provided preliminary survey results of Building 4024, noting that the “complete radiological 

survey of building 024 appears to require more extensive effort than originally expected.” 

According to the letter, test vault 1 (S2DR and SNAPTRAN-1), test vault 2 (SNAP10FS-3), the 

vacuum equipment room, and radioactive liquid waste system were marked with a 1-meter 

square grid for a detailed smear and instrument survey. The letter did not indicate the type of 

1 EPA/HGL, Final HSA 5-A, p. 33, citation 128: 

.

128 Owen, R.K. Atomics International Letter Re: Radiation Survey of SNAP 10FS-3 Reactor in Test 

Cell 2, Building 024, April 12, 1966. 

2 EPA/HGL, Final HSA 5-A, p. 33, citation 129: 

.	 129 Owen, R.K., Atomics International Letter Re: Remote Radiation Survey of Transverse Corridor of Test 

Cell 2, Building 024, May 3, 1966. 

3 EPA/HGL, Final HSA 5-A, p. 33, citation 130: 

.	 130 Owen, R.K., Atomics International Letter, Re: Radiation Survey #2, SNAP 10FS-3 Reactor in Test Cell 2, 

Building 024, May 4, 1966. 

4 EPA/HGL, Final HSA 5-A, p. 33, citation 131: 

.	 131 Wilson R.F., Atomics International Letter Re: Building 024 – FS-3 Vault and Transfer Cell Contamination 

Levels, July 14, 1966. 

5 EPA/HGL, Final HSA 5-A, p. 27, citation 96: 

.	 96 Wilson R.F., Atomics International Letter Re: Building 024 – FS-3 Vault and Transfer Cell Contamination 

Levels, July 14, 1966. 

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instrument used. In test vault 1, the floor showed 55 dpm/100 cm2 in one location, while the 

remainder of the survey grids showed less than 20 dpm. The letter does not indicate whether 

the contamination was fixed or removable. A light fixture that was surveyed on the south wall 

measured 4 mR/hr. According to the letter, “one storage hold lid was lifted. The liner appears 

to be floating, suggesting that the lower part of the hole is filled with water, but this could not 

be checked at the time.” 

The EPA HSA research team did not find any additional information on the “storage hold lid” or 

its location, and, as a result, could not (at the time the HSA was written) determine if it was in 

contact with groundwater.1 In addition, the floor of test vault 2 was randomly smear-checked 

following removal of the vacuum vessel. The letter does not provide information to indicate the 

number of smears performed; but the smears found no contamination about 20 dpm. 

Radiation levels in the vacuum equipment room were reported to be at the 1977 background 

levels. The letter did not provide 1977 background levels. The radioactive liquid waste system 

measurements comprised the “low level waste tank water,” the “suspect tank water,” and the 

“sump water.” The letter did not provide information to indicate the location of these liquid 

wastes. The results as reported in the letter are presented as follows:2 

• Low level waste tank water

o alpha – 1.2 x 10-9 

μCi/mL 

o beta – 1.7 x 10 8 

μCi/mL 

•	 Suspect tank water
 

-9 o alpha – 2.1 x 10 μCi/mL 

o beta – 1.4 x 10-8 

μCi/mL 

•	 Sump water
 

-8 o alpha – 8.8 x 10-10 

μCi/mL o beta – 2.5 x 10 μCi/mL 

• According to an undated report titled, “Long-Range Plan for Decommissioning Surplus 

Facilities at the Santa Susana Field Laboratories,” a partial decontamination project of 

Building 4024 was performed in 1978. The source document does not indicate how the items 

were decontaminated or where the decontamination took place. The acceptable limits for 

residual radioactivity for surface contamination were reported by Rockwell 

International to be 7,500 dpm/100 cm2 maximum beta-gamma emitters and 300 dpm/100 

cm2 maximum alpha emitters. Accessible surfaces within the cell and support areas 

were cleaned to acceptable removable contamination limits of 100 dpm/100 cm2 beta1 EPA/HGL, Final HSA 5-A, p. 33, citation 132:

132 Owen, R.K., Rockwell International, Re: Preliminary Survey Results – Building 024, February 11, 1977. 

2 EPA/HGL, Final HSA 5-A, p. 33, citations 133 & 134:

133 Owen, R.K., Rockwell International, Re: Preliminary Survey Results – Building 024, February 11, 1977.

134 The alpha and beta measurements are those reported in the letter; however, their values are very low. It 

is important to note that the micro were hand written on the document, which may have been a result of the

original document having been drafted on a type writer, or may have been added in error. 

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gamma and 20 dpm/100 cm2 alpha.1

• A November 1978 radiological survey report presents the findings of a September 1978 

survey to ensure that the facility met unrestricted release criteria. The survey included 

800 smears from the walls, floors, and remaining piping and equipment. No beta 

contamination in excess of 50 dpm/100 cm2 was reported; no alpha activity was reported 

for any areas of Building 4024 including the test vaults, transfer cell, operating gallery, high 

bay, and equipment rooms and support area. The acceptable limits for residual 

radioactivity for surface contamination were reported by Rockwell International to be 

7,500 dpm/100 cm2 maximum beta-gamma emitters and 300 dpm/100 cm2 maximum 

alpha emitters.

According to the report, smears were counted for alpha and beta activity on a Nuclear 

Measurements Corporation automatic counting system with an average background of 25 

cpm for beta and a counting efficiency factor of 2.35 dpm/cpm for beta. The report stated that 

alpha contamination was not suspected; however, had there been alpha contamination it 

would have been detected by the Nuclear Measurements Corporation system, and during the 

course of the survey of removable contamination no alpha activity was detected.

Rockwell International also performed a survey of surface radiation. This survey was 

conducted throughout the building and surrounding area using a beta-gamma ion chamber. 

Beta-gamma surface dose rate limits were 0.1 mrad/hr (average) and 0.5 mrad/hr (maximum). 

The maximum beta-gamma surface contamination detected outside of the power vaults was 

0.07 mrad/hr with an average background of 0.05 mrad/hr. Inside the power vaults, betagamma surface contamination was found to range from 0.5 mrad/hr to 2.5 mrad/hr. Inside the 

corridor to the power vaults, beta-gamma surface contamination ranged from 0.02 mrad/hr to 

1.8 mrad/hr. According to the report, the center floor drain in the lower level operating gallery 

was found to be contaminated during decommissioning and disposal operations. Following an 

acid cleaning, water and smear samples indicated 0.08 mrad/hr using the ion chamber and 

200 cpm using a Technical Associates Model Pug-1. The report summarized that 

contamination was “probably located in the threads at the bottom of the vertical pipe.”

Soil samples were also collected in the yard during and following the removal of the liquid and 

gas holdup tanks. According to the report, all samples were less than 30 pico curies per gram 

(pCi/g). The soil samples were counted on a Nuclear Chicago automatic counting system with 

a KCI standard, with an average background of 30 cpm and a counting efficiency factor of 

3.29 dpm/cpm. The report approximated the natural activity of uncontaminated soil in this area 

to be 20 to 30 pCi/g; however, the report did not identify or provide a reference to how these 

natural activity values were determined.

Concrete cores drilled in the power vault walls and corridors were found to have a maximum 

specific activity of 818 pCi/g. The average specific activity was 103 pCi/g. Rockwell 

International analyzed fourteen cores at 1-inch increments to a depth of between 10 and 11 

inches, for a total of 144 samples. The average specific activity was 103 pCi/g.

1 EPA/HGL, Final HSA 5-A, p. 35, citation 136:

Rockwell International Document N001T1000200, Long-Range Plans for Decommissioning Surplus

Facilities at the Santa Susana Field Laboratories, Date Unknown. 

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Rockwell International collected water samples from the drain pipe in the operating gallery, hot 

waste storage vaults, cooling system water waste holdup tanks, groundwater during the 

removal of the waste tanks, and the vacuum cleaning line in Power Vault 2. The report 

indicated all the samples were below the limit for Sr-90. However, some water was found to be 

contaminated above the limit during decommissioning and disposal work, and it was 

transferred to the RMHF for disposal. The report did not indicate the volume of water that was 

found to be contaminated above the limit during decommissioning and disposal.1 

• On March 26, 1981, additional concrete sampling in the power vaults began to determine the 

amount of concrete needed to be removed to meet 1981 unrestricted release criteria. 

Potential radiological hazards were identified as being limited to the high bay area (including 

cell complex), electrical/mechanical support, and yard areas. Two general areas of concern in 

the high bay were the cells and the S10FS-3 reactor support equipment room. Additionally, the 

electrical/mechanical support area contained other equipment and systems that were 

potentially contaminated.

As a result of exposure to neutrons escaping from the two operating reactors, the walls, 

ceiling, floor and remote handling equipment of the test cells were activated. The survey 

indicated that 12 to 22 inches of concrete would need to be removed for surface radiation to 

meet the acceptable dose rate of 0.1 mR/hr. Only two radionuclides, Co-60 and Eu-152, were 

identified as the principal contaminants found to contribute significantly to radiation greater 

than background. The report identified the following activation products as contributing to 

contamination at Building 4024: Fe-55, Mn-54, and Eu-154.2 

• According to a December 1992 safety review report regarding tritium production and release 

at SSFL, production of tritium by the operation of the S2DR, and later the S10FS3, occurred 

by the following ways 3:
 


 

o Neutron absorption by lithium-6 present in the concrete biological shield – The concrete 

used for the inner 2 feet of the shield differed from the concrete usually used at SSFL in that 

limestone aggregate was substituted for the normally used granitic gravel. The natural lithium 

content of this concrete was measured in 1992 to be 6.2 ppm. Using this concentration and an 

average flux of 1.2 x 108 n/cm2/sec at 37 kWt for the first 112 cm of concrete (and considering 

1 EPA/HGL, Final HSA 5-A, p. 36, citations 137 & 138: 

.	 137 Rockwell International Document N001T1000200, Long-Range Plans for Decommissioning Surplus

Facilities at the Santa Susana Field Laboratories, Date Unknown.
 138 Rockwell International Document, N704TI990044, “Radiological Survey Results – Release to

Unrestricted Use, Building 024, SSFL,” November 28, 1978. 

2 EPA/HGL, Final HSA 5-A, p. 36, citations 139 & 140: 

.	 139 Atomics International Document, N704FDP990006 Rev. A., “Building T024 (SETF) Facilities Dismantling 

Plan,” July 31, 1981. (This document was referenced in the May 2005 HSA but has not yet been obtained.)

140 Rockwell International Document N001T1000200, Long-Range Plans for Decommissioning Surplus

Facilities at the Santa Susana Field Laboratories, Date Unknown. 

3 EPA/HGL, Final HSA 5-A, p. 36, citation 141: 

.	 141 Tuttle, R.J., Rockwell International Report RI/RD92-186, Tritium Production and Release to Groundwater 

at SSFL, December 1, 1992. 

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that this flux was uniform over all the walls, floor, and ceiling in the vault) the activity of tritium 

produced in the concrete of Vault 1 by operation of S2DR is estimated to be 0.16 Ci at 

shutdown in December 1962. In a similar manner, adjusting for power level and operating 

time, the tritium activity produced by operation of S10FS3 in Vault 2 is estimated to be 0.20 Ci 

at March 1966.

o Neutron absorption by lithium-6 present in soil surrounding the biological shield – The thick 

concrete biological shield with the added boron content absorbed nearly all the neutron that 

escaped from the reactor (less than 1 neutron per million produced by the reactor reached the 

soil), and so the production of tritium in the soil was negligible.

o Ternary fission – This was calculated to produce 0.21 Ci in the fuel of the S2DR by 

December 1962, and 0.26 Ci in S10FS3 by March 1966.

o Neutron absorption by lithium-6 present as an impurity in NaK coolant – Tritium produced in 

this manner was negligible for both reactors.

o Neutron absorption by lithium-6 in the lithium hydride shield – This activity was estimated 

from the calculation for S8DR at Building 4059, adjusting for power and time, for the S10FS3 

operation. The natural isotopic fraction of 7.42 atom percent was used to estimate a 

production of 2,200 Ci in the S10FS3 shield.

• In September 1995, ORISE conducted an independent verification survey at Building 4024. 

Surface scans were performed over 50 to 100% of accessible floors and lower walls (up to 2 

meters) for alpha, beta and gamma activity. In the fan room, elevated direct beta radiation was 

identified. While, in all other areas, alpha, beta and gamma radiation was within the 1995 

range of ambient site background. Acceptable contamination limits and gamma exposure 

rates for releasing a facility for unrestricted use are prescribed in the U.S. Department of 

Energy (DOE), the U.S. Nuclear Regulatory Commission (NRC), and the State of California 

guidelines. The lowest, most conservative limits were chosen from these guidelines and 

incorporated into the final survey criteria for Building 4024. The surface contamination limits 

for alpha and beta were excerpted from DOE Order 5400.5 and NRC Regulatory Guide 1.86 

(see Table below). The ambient gamma exposure rate limits at 1 meter were excerpted from 

an NRC Dismantling Order because at 5 microroentgens per hour (μR/hr) it was more 

conservative than the DOE value of 20 μR/hr, and more consistent with as low as reasonably 

achievable principles.1 

1 EPA/HGL, Final HSA 5-A, p. 36, citation 142:

142 ORISE Report, 96/C-5, “Verification Survey of Buildings T019 and T024, Santa Susana Field Laboratory, 

Rockwell International, Ventura County, California,” February 1996. 

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6.1.4.2

Surface Contamination Guidelines from DOE Order 5400.5 (1990) and NRC Regulatory Guide 1.86 (1974)

Allowable Total Residual Surface Contamination (dpm/100 cm2) 

External Gamma Radiation: The average level of gamma radiation inside a building or habitable structure on a site that has 

no radiological restriction on its use shall not exceed the background level by more than 20 μR/hr. Source: U.S. Atomic 

Energy Commission (now NRC) Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors, June 1974, p. 

1.86-5 U.S. Department of Energy Order 5400.5, Radiation Protection of the Public and the Environment, February 8, 1990, p. IV-6. 

Radionuclides Average Maximum Removable

Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa-231, 

Ac-227, I-125, and I-129

100 300 20

Th-Natural, Th-232, Sr-90, Ra-223, Ra-224, U-232, I-126, 

I-131, and I-133

1,000 3,000 200

U-Natural, U-235, U-238, and associated decay products 5000α 15,000α 1,000α

Beta-gamma emitters (radionuclides with decay modes 

other than alpha emission or spontaneous fission) except 

Sr-90 and others noted above

5,000β-γ 15,000β-γ 1,000β-γ

Surface activity measurements were conducted at 76 floor and wall locations. Excluding the 

power vaults, surface activity levels were less than 55 dpm/100 cm2 for alpha and ranged from 

less than 1,400 to 33,000 dpm/100 cm2 for beta. Removable alpha measured less than 12 dpm/ 

100 cm2, and removable beta measured less than 16 dpm/100 cm2. The maximum beta-gamma 

total surface activity guidelines (15,000 dpm/100 cm2) as exceeded in the hot gas compression 

room of Building 4024. Exposure rate measurements were made at four locations in Building 

4024, but none were made in the power vaults. Excluding the power vaults, exposure rates 

ranged from 11 to 13 micro roentgen per hour (μR/hr). Background was 8 μR/hr. Interior 

exposure rates satisfy DOE and NRC exposure rate guidelines. ORISE determined that existing 

documentation for Building 4024 was inadequate to support the determination that DOE 

guidelines for unrestricted release were met. Deficiencies noted included inadequate final status 

survey methods, no discussion of specific contaminants, inconsistent specification of all 

applicable guidelines and presentation of data that may be compared to the guidelines, absence 

of quantitative laboratory data, and absence of adequate figures documenting remediated areas 

and measurement and sampling locations.1 

The EPA-HSA contains more details about subsequent, chronological radiological investigations 

of Building 4024 and recommendations for the final D&D activities. 

6.1.4.3 Radiological Use Authorizations 

A review of a database documenting the use authorizations related to SSFL, the EPA research 

team identified two use authorizations relating to Building 4020 (Hot Lab). Authorization No. 46 

relates to the use of Cs-137 and Pu-Be sources in Building 4024, and Authorization No. 112 

relates to the “surveillance of Building 4024.” 

1 EPA/HGL, Final HSA 5-A, p. 38, citation 143: 

143 ORISE Report, 96/C-5, “Verification Survey of Buildings T019 and T024, Santa Susana Field Laboratory, 

Rockwell International, Ventura County, California,” February 1996. 

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Authorization No. 46 was issued to Atomics International August 5, 1971 and expired on 

September 5, 1971. As referenced above, the authorization was for the use of a Cs-137 source 

and a Pu-Be neutron source in the high bay section of Building 4024. The authorized materials 

included 4.5 Ci of sealed source Cs-137, 5 Ci or sealed Pu-Be source identified as Pu-239, and 

an additional 10 mCi of sealed source Cs-137. According to the user application dated August 1

3, 1971, the purpose of the test was to observe the effect of gamma rays and neutrons on an 

infrared sensor. The application noted that there would be no contaminated materials since the 

sources were encapsulated. 

Authorization No. 112 relates to the surveillance of Building 4024 under the Surplus Facilities 

Management Program. The first available use authorization was issued on January 16, 1978 

and expired January 16, 1979. The authorization operation was listed as “decontamination and 

disposition of Building 024.” the authorized materials were listed as being, “activation product 

radioactivity” of an unknown quantity in the building’s structure and concrete.2 The authorization 

was renewed yearly; and in March 1992, the authorization was modified to include the following 

operations; surveillance and maintenance activities for contaminated areas, future 

decontamination and decommissioning activities (however not scheduled for the duration of the 

1992 annual authorization), and storage of empty radioactive materials containers. By 1994, 

Authorization No. 112F included required environmental monitoring at Building 4024. This 

monitoring included quarterly radiation surveys and quarterly contamination surveys.3 

Subsequent use authorizations after Boeing took over site operations (1996) added the storage 

and survey of contaminated waste materials from Building 4020 Hot Lab decontamination and 

decommissioning operations. The requirement was that a running inventory of Building 4020 

materials stored in Building 4024 would be maintained, but no additional controls for these 

operations were identified in the authorizations. In 1997, added operations included the 

decontamination of Building 4020 materials. 

While active decontamination of Building 4020 materials occurred in Building 4024, personnel 

were to perform weekly routine surveys of radiation and contamination levels. During periods 

when decontamination activities did not occur, personnel were instructed to perform monthly 

surveys. By 1999, use authorizations expanded to include the storage of radioactive sources 

pending transfer to a third party recycling or disposal vendor, and the authorization continued to 

1 EPA/HGL, Final HSA 5-A, p. 40, citation 153:

153 Baumesh, L., and Heine, W., Authorization No. 46., August 5, 1971. 

2 EPA/HGL, Final HSA 5-A, p. 40, citation 155:

155 Tuttle, R., Authorization No. 112, January 16, 1978. 

3 EPA/HGL, Final HSA 5-A, p. 41, citations 156-161:

156 Barnes, J.G., and Rutherford, P.D., Authorization No. 112D, March 18, 1992.

157 Barnes, J.G., and Rutherford, P.D., Authorization No. 112F, April 20, 1994.

158 Barnes, J.G., and Rutherford, P.D., Authorization No. 112H, January 30, 1996.

159 Barnes, J.G., and Rutherford, P.D., Authorization No. 112I, January 14, 1997.

160 Barnes, J.G., and Rutherford, P.D., Authorization No. 112I, January 14, 1997.

161 Barnes, J.G., and Rutherford, P.D., Authorization No. 112L, March 15, 1999. 

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list the storage and survey of contaminated waste material from Building 4020 as an “operation.” 

However, the decontamination of building blocks was expanded to include “L-85 lay down area,” 

but did not provide additional detail to describe what was meant by, “L-85 lay down area.” 

6.1.4.4 Building 4024 Open Storage Areas 

Aerial photographs over the years of Building 4024 operations show open storage areas 

between 1967-1972, container leakage and open storage areas in 1980, 1983, and potentially 

1985. 

The preliminary MARSSIM Classification for the building 4024 area is Class 1, due to its site 

operations, results of previous radiological investigations, and current building status. EPA’s 

research team emphasized that there were radiological incidents at Building 4024 and 

documented evidence of radiological releases; significant information is lacking regarding 

excavation activities at Building 4024. Additionally, EPA emphasized that past characterization 

studies for Building 4024 were focused on delineating the extent of contamination to standards 

that were applicable at the time, and EPA made additional recommendations for appropriate 

characterization of the Building 4024 area. 

6.1.4.5 Building 4024 Radiological Incident Reports 

There have been several incidents associated with Building 4024 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

6.1.4.6 Building 4024 Radiological Incident Report Summary: Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0535 3/6/1962 SETF High Bay Active 

Corrosion 

Products

Welder burned with NaK in High Bay

A0634 2/19/1970 Building 4024 

General Mills

None Identified Maintenance Workers Unknowingly 

Worked on Contaminated General Mills

A0686 4/28/1998 Building 4024 Yard None Identified Non-Contaminated Skin Rash

• On March 6, 1962 a welder began grinding on the upper weld of the “PCS” in the high bay. 

The welder stopped, removed the tape over the lower weld and NaK ran out onto his hands 

and left leg. The welder “flipped” the NaK off his hands and his coveralls started to burn. The 

welder removed the coveralls and entered the shower. The Fire Department cleaned up the 

approximately 1/4 cup of NaK. According to the incident report, health and safety personnel 

found no detectable beta and gamma contamination in the high bay or on the welder 

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(A0535).1

• On February 19, 1970, maintenance workers unknowingly worked on contaminated general 

mills (A0634). 2

• On April 28, 1998, an employee working at Building 4024 developed a skin rash on left thigh 

and both ankles. The employee believed the rash to be “beta burns” caused by beta radiation. 

A thorough investigation of the worker and the area found no radiation or contamination. The 

employee was diagnosed with poison oak at the West Hills Medical Center (A0686). 3

1 EPA/HGL, Final HSA 5-A, p.32, citation 122:

122 Sessions, S.D., “Radiological Safety Incident Report A0533, SETF High Bay, Building 24,” March 21, 

1962. 

2 EPA/HGL, Final HSA 5-A, p.32, citation 123:

123 The research team has not yet received the incident report for this incident. The incident was 

documented in Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV 

Santa Susana Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, 

p. V-3. 

3 EPA/HGL, Final HSA 5-A, p.32, citation 124:

124 Liddy, Patricia, “Incident Report File A0686, Worker Concern Over Skin Rash,” May 1, 1998. 

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6.2 SNAP COMPLEX: ADDITIONAL LOCATIONS / OPERATIONS 

Several important SNAP buildings/facilities were not included in the 2006 Site Profile. The 

SNAP Complex at SSFL consisted of numerous structures that functioned to support SNAP

operations. As such, some were designated as “non-radiological” facilities. However, in some 

cases, evidence of radiological processes and radiological incidents at these locations has been 

discovered. 

This section a partial list of SNAP Complex buildings/facilities that should be included in the 

2006 Site Profile, based on their contributions to the SNAP program, involvement in radiological 

processes, and proximity to sources of radiation. It is not possible to obtain a comprehensive 

understanding of facility operations without a robust description of the SNAP Program and 

SNAP Complex. 

>> Interesting Fact: Many are unaware of NASA operations in Area IV, particularly those 

involving the SNAP program (a joint DOE-NASA endeavor). SNAP Building 4373 (Section 6.1) 

provides an excellent example of NASA-DOE building operations; prior to 1958 the location was 

operated by Rocketdyne Space & Information Division for the testing of high-powered, solid 

rocket fuel. NASA operated a nuclear reactor (HMRFSR) in SNAP Building 4012, the world’s 

largest CT-Scanner in AETR Building 4100, and developed the Space Shuttle Main Engine and 

International Space Station in SNAP Building 4057, under LMEC/ETEC programs. In many 

instances, NASA’s Area IV collaborations with DOE relied on the participation of Rocketdyne 

Space & Information workers commonly excluded from EEOICPA due to their “Time Clock 

Location” assignments in Areas I, II and III. 1

1 According to DEEOIC, North American Aviation’s original contract with the AEC permitted the company’s 

discretionary use of all its facilities in the fulfillment of its contractual obligations with the government. As such, 

workers of all areas rotated as needed.It is not always possible to determine worker locations or establish eligibility.

See file: Turcic_Long_ETEC.pdf 

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Section 6.2.1 

Building Number: 4013 

Building Alias: 013/T013 

Building Name/Function: SNAP Component Assembly & Performance Building

X-Ray Booth

ETEC Thermal Transient Facility (TTF) 

Notes: DOE-NASA 

Associated Buildings: 4713 Electrical Substation 

4823 Time Clock Location 

4413 Uninterruptible Power Supply (UPS) 

Radionuclides of Concern: The HSA research team found no evidence that nuclear or 

radioactive materials were known to have been handled in Building 4013. However, 

radionuclides associated with potential migration from SNAP Buildings 4012 and 4019 include: 

Am-241, Cs-137, Pu-238, Pu-239, Pu-240, Pu-241, Sr-90, U-234, U-235 and U-238 (natural and 

enriched). In addition, H-3 and Be were used in proximity to Building 4013, and may have 

migrated to this location via surface water flow or airborne release. 

6.2.1.1 Description of Operations & Processes: 

Building 4013 was constructed in 1962 as a non-nuclear Systems for Nuclear Auxiliary Power 

(SNAP) component assembly and performance testing building. It was situated near the 

northwest property line of Area IV where the terrain is very rugged and there is a 40’ drop in 

elevation. Building 4013 was situated between the SNAP Critical Test Facility Buildings 4012 

and 4019. 

Assembly and checkout of several SNAP units were accomplished in the 1960’s at Building 

4013: PSM-1, PSM-3, PSM-1A, FSEM-2, PSM-1B, FSM-1, FSEM-2A, FS-3, FSEM-4, FS-4 and 

FS-5. Building 4013 featured a general assembly area, electronic assembly area, “T/C” 

assembly area with X-Ray booth, mechanical area, shop support area with welding booth, and 

receiving/holdup area. 

In 1970, Building 4013 was re-designated the ETEC Thermal Transient Facility. Half the high 

bay was used for thermal testing since that time, and the other half was used for seismic test 

equipment.1 Authorizations for building modifications and improvements to support SNAP 

programs was also granted in 1970. Improvements included the installation of new partitions, 

1 EPA/HGL, Final HSA 5B, p. 34, citation 1: 

1 Chapman, J. A., Radiological Survey of Buildings T019 and T013; An Area Northwest of T059, T019, T013, 

and T012; And a Storage Yard West of Buildings T626 and T038, GEN-ZR-0010, Rockwell International, 

Rocketdyne Division, August 26, 1988, pgs. 2, 10, 17, 20, 24, 92. 

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floor tile, ceilings, doors, air conditioning equipment, ducting, utilities, diffusers, and equipment 

pads for clean room and low-bay cooling requirements, in addition to a new paint job.1 

Although Building 4013 was designated a “non-radiological” facility, its proximity to the SNAP

Critical Facility (Building 4012) and storage of radiological materials necessitates its 

consideration for potential worker exposure and its inclusion to the SSFL Site Profile. 

In 1988, Rockwell International performed a final status survey of Building 4013 along with 

Building 4019 and two adjacent areas to determine if any radioactive material remained that 

required further surveying or decontamination, thus acknowledging that Building 4013 may have 

been impacted. The HSA provides a detailed chronology of the radiological investigations and 

decontamination/cleanup and release of Building 4013. Rockwell International adopted its own 

“acceptable limits” for this survey, based on enriched uranium used for SNAP programs. 

Due to Building 4013’s location between SNAP Critical Test Facility Buildings 4012 and 4019, 

potential radioactive material migration via surface water flow or airborne release from these 

facilities may affect Building 4013. In addition, previous characterization studies for the building 

4013 were focused on delineating the extent of contamination to standards that were applicable 

at the time, and not to the standards required during the EPA Area IV Study. Moreover, leakage 

from Building 4019 was observed in aerial photographs, which could have affected Building 

4013, and storage containers of unknown contents were present at Building 4013. 

The preliminary MARSSIM classification of Building 4013 is Class 1 due to the potential for 

radioactive material migration via surface water flow and airborne releases from SNAP Critical 

Test Facility Buildings 4012 and 4019. 

6.2.1.2 Building 4013 Radiological Incident Reports 

There has been one incident associated with Building 4013 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing. Summaries of all available incident reports are 

provided in Attachment A. 

6.2.1.3 Building 4013 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0369 4/9/1963 X-RAY BOOTH Investigation of possible over-exposure 

appears to be a medical exposure.

• No Incident Report summary was provided by the HSA research team.

1 EPA/HGL, Final HSA 5B, p. 34, citation 2: 

2 Correspondence from Pollman, A.P., U.S. Atomic Energy Commission, to Petersen, R.C., Atomics 

International, Reference: Building 013 Modifications and Improvements, SNAP, Contract AT(04-3)-701, 

dated October 16, 1970. 

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Section 6.2.2 

Building Number: 4025 

Building Alias: 025/T025 

Building Name/Function: SNAP Remote Handling Mock-Up Building 

Notes: DOE-NASA

“Non-Radiological” Designation

Radiological processes documented

Possible Tritium 

Associated Buildings: 4725 & 4924 Electrical Substations 

4925 Mechanical Equipment Slab

4926 SRE Mock-up Equipment Bldg. 

Radionuclides of Concern: Co-60 (Suggesting another sealed source in the building); 

Radionuclides associated with potential migration from the RMHF Building 4075 include 

isotopes of uranium, thorium, plutonium, and mixed fission products. Radionuclides associated 

with nearby SNAP Building 4024 include Am-241, Cs-137, Co-60, Eu-152, Eu-154, Pu-238, 

Pu-239, Pu-240, Pu-241, Sr-90, H-3, U-234, U-235, U-238 

6.2.2.1 Description of Operations & Processes 

Building 4025 was constructed in 1959 as the Remote Handling Mock-Up Building to support 

the Systems for Nuclear Auxiliary Power (SNAP) reactor tests. It was designated a “nonnuclear” facility.1 However, there is a radiological incident associated with this building, which 

involved Co-60. According to the HSA research team, this suggests other sealed sources used 

in the building. 

In addition, Building 4025 was located approximately 200 feet from the Radioactive Materials 

Handling Facility (RMHF) Building 4075, which stored and processed high levels of radioactive 

waste, and in close proximity to Building 4024. Proximity to the RMHF and Building 4024 likely 

impacted ambient radiation conditions, and this location was likely subject to radioactive 

material migration from both nearby buildings. Moreover, given site activities during the 1980’s 

when Building 4025 was re-designated as ETEC’s Instrumentation and Inventory Building, and 

the outside area surrounding it was used as a storage yard for salvageable materials and scrap 

components, it is likely that materials stored at this location were contaminated with 

radioactivity. 

The HSA contains detailed information about the building’s features and schematics, and 

references a floor-plan that depicts a pit at the southern end of the middle bay, as well as other 

1 EPA/HGL, Final HSA 5B, p. 48, citations 1 & 2: 

1 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992.

2 Chapman, J. A., Radiological Survey of Buildings T049, T042, T027, T032, and T025, GEN-ZR-0013, 

Rockwell International, Rocketdyne Division, August 8, 1988, pgs. 9, 17, 22, 24, 26, 31, 35, 109-114, 

119-120. 

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documentation that references two 10’x10’ pits. No other information on the pits has been 

located at this time. 

Building 4025 was used as the Remote Handling Mock-Up Building through 1970 for remote 

handling and viewing of mock-up work in support of SNAP2/10A tests, SNAP 8 tests, and other 

nuclear reactor tests. Following cancellation of the SNAP program in 1970, Building 4025 was 

designated the ETEC Instrumentation and Inventory Building used for component assembly, 

storage, a salvage location, and instrumentation work at least through 1988. 

The preliminary MARSSIM classification for Building 4025 is Class 1 because of its proximity to 

SNAP Building 4024, RMHF Building 4075, and OS-2, and the potential of radioactive material 

migration via surface flow or airborne release from the nearby buildings. 

6.2.2.2 Building 4025 Radiological Incident Reports 

There has been one incident associated with Building 4025 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing. Summaries of all available incident reports are 

provided in Attachment A. 

6.2.2.3 Building 4025 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0306 11/16/1979 Shielded Area Co-60 Off-scale dosimeter during gammagraph 

operation.

The HSA provides a detailed chronology of radiological investigations and decontamination/

cleanup of releases associated with Building 4025. However, during the 1988 Rockwell 

International radiological survey of Building 4025, background radiation levels were found to 

vary considerably and had to be statistically corrected. The most important factor influencing 

ambient background radiation conditions around Building 4025 was its close proximity to the 

Radioactive Materials Handling Facility (RMHF), which stored high levels of radioactive material. 

The exterior storage yard was approximately 200 feet from RMHF Building 4075. After 

correcting for background radiation, the survey concluded that no residual radioactivity was 

found above Nuclear Regulatory Commission acceptance limits of 5 μR/h in 1988.1 

1 EPA/HGL, Final HSA 5B, p. 49, citation 4: 

4 Chapman, J. A., Radiological Survey of Buildings T049, T042, T027, T032, and T025, GEN-ZR-0013, 

Rockwell International, Rocketdyne Division, August 8, 1988. 

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Section 6.2.3 

Building Number:

Building Alias: 

4027 

027/T027 

Building Name/Function: SNAP Mechanical Vibration/Shock Testing Facility

S10FS-1 Acceleration Tests 

Storage of Gamma Radiography Equipment

Storage of Hazardous Waste

Offices, Shop Support, Darkroom

Radiological Uses & Incidents 

Notes: SNAP was a DOE-NASA project involving North American Aviation Space & 

Information Division, Rocketdyne. 

Radionuclides of Concern: Building 4027 is located in proximity to the RMHF; therefore direct 

radiation and sky shine from RMHF affects ambient radiation conditions in the area.1 

Radionuclides of concern at the RMHF include all radionuclides that are included in the 

background study plus any additional radionuclides identified during the HSA. Additionally, 

sealed, fully-encapsulated, radiography sources and/or equipment may have been stored in 

Building 4027; however, no known “loose radioactive material” was handled in the building.2 

6.2.3.1 Description of Operations & Processes: 

Building 4027 Area includes Building 4027, Building 4625, Substation Building 4727, and the 

surrounding area. Building 4027 was one of a series of buildings that supported the 1950-

and-1960 era SNAP Program, used as engineering development and test sites. It was 

constructed in 1961. The building’s operational history is one of many that provides insight on 

the need for Rocketdyne employees from the Space & Information Division (S&ID) that may 

have been associated with SSFL Areas I, II or III to participate in SNAP operations on an asneeded, often undocumented basis. 

Despite the building’s designation as a “non-radiological” facility, according to an April 1987 Site 

Consolidation Assessment, Building 4027 was used for storage purposes, including sealed 

sources for gamma-radiography and hazardous waste. Additionally, an undated listing of SSFL

Site Waste Generators, as well as a 2003 Environmental Assessment for Cleanup and Closure 

1 EPA/HGL, Final HSA 5-A, p. 49, citation 215:

215 ETEC Document, GEN-ZR-0013, “Radiological Survey of Buildings T049, T042, T027, T032, and T025,” 

August 26, 1988. 

2 EPA/HGL, Final HSA 5-A, p. 49, citation 216: Ibid. 

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identifies Building 4027 as a “weld shop.”1 Weld shops at SSFL routinely handled a variety of 

tasks, from all areas and types of facilities. When one weld shop was not available, others 

would be used as needed. The handling of radiological materials or radioactively contaminated 

equipment at this location cannot be ruled out, if it was used as a weld shop. 

In 1963 the building was expanded to include Building 4625, which operated as a non-nuclear 

Component Storage Building. Building 4027 featured a high bay laboratory area, offices, shop 2

support, and at one time a large darkroom. Vibration and shock tests were performed in Building 

4027 in support of the SNAP launch schedule.3 

According to an August 15, 1963 incident report, the building conducted acceleration tests on 

the S10FS-1 system, which included beryllium and NaNO2 to simulate NaK in the S10FS-1 

system. The acceleration tests typically involved vibration and shock tests to evaluate the 

effects of high acceleration experienced during Atlas rocket missions. 

According to an August 6, 1964 Atomics International letter the use of cyclohexane and/or 

4 cycloheptane in the NaK system of the SNAP 10FS-3 in Building 4027 was authorized. In 1965, 

such tests included the testing of the SNAP 8 drum drive mechanism and the SNAP 10FS-5 

reactor.5 

On August 13, 1963 an Atomics International employee notified the Health and Safety 

Department at Santa Susana that a bottle of water that had been exposed to possible 

radioactive and beryllium contamination had been used for a drinking water supply in Building 

4027. The contents of the bottle had been used during acceleration tests related to the S10FS-1 

in Building 4027. The employee removed the bottle from the water cooler because employees 

1 EPA/HGL, Final HSA 5-A, p. 47, citations 202-207:

202 ETEC, Site Consolidation Assessment, April 16, 1987. p. 22.

203 ETEC: GEN-ZR-0013, “Rad. Survey of Buildings T049, T042, T027, T032, & T025,” Aug. 26, 1988.

204 Tuttle, R.J., Listing of Locations in Area IV Associated With Radioactive Materials,” Sept. 1989.

205 Boeing, R/A Waste Certification Plan (WCP) for ETEC Facilities, EID-04758, Feb. 21, 2001. p. 5. 

206 Unknown, SSFL Waste Generators, Undated.

207 DOE, Environmental Assessment for Cleanup and Closure of ETEC, DOE/EA-1345, March 2003. p. 2-10 

2 EPA/HGL, Final HSA 5-A, p. 46, citation 196:

196 Drawing 303-027-A4, “Expansion of Non-Nuclear Mechanical Vibration and Shock Testing, Building 027 

Expansion, Floor Plan, as built,” 1964. 

3 EPA/HGL, Final HSA 5-A, p. 46, citation 197:

197 ETEC Document, GEN-ZR-0013, “Radiological Survey of Buildings T049, T042, T027, T032, and T025,” 

August 26, 1988. 

4 EPA/HGL, Final HSA 5-A, p. 46, citation 199:

199 Atomics International Internal Letter, Letter Re: SNAP 10FS-3, Building 027, Santa Susana, August 6, 

1964. 

5 EPA/HGL, Final HSA 5-A, p. 47, citation 200:

200 Wilmes, R.F., Atomics International Internal Letter Re: Weekly Progress Report for Industrial Hygiene 

and Safety Unit, Santa Susana, Period Ended 1-16-65, January 19, 1965. 

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were complaining about the taste of the water. Upon doing so, the employee noted a strip of 

masking tape that had written on it: “DO NOT DRINK, WATER CONTAMINATED WITH NaNO2.” 

According to the investigation, on July 19, 1963 personnel drained the water from the S10FS-1 

assembly into two 5-gallon water bottles of the type used for drinking water supply at SSFL. The 

bottles were tagged and sealed for chemical analysis. An unknown person moved the bottles to 

the water cooler, and one of the bottles was placed on the water cooler on August 12, 1963. By 

August 13, 1963, approximately one half of the content of the water showed beta and 

gamma contamination of less than 5 x 10-8 uc/cc, alpha contamination of less than 1 x 10k 

uc/cc, and beryllium contamination of 0.21 ug/cc. The incident report did not provide the 

results of the NaNO2 concentrations, and did not indicate the ultimate disposal of the remainder 

of the contaminated water (Incident Report A0023).1 

Building 4027 was included in the DOE SSFL Site Survey to determine if any residual 

radioactivity was accidentally left behind as a result of operations in support of the SNAP

program. However, due only to the building’s designation as a non-radiological facility, it was 

assumed no radioactive materials were handled and according to the 1988 survey report, “only 

gamma exposure rate measurements were acquired in order to assess radiological condition 

because no radioactive or nuclear materials are known to have been handled here.” In the high 

bay, the maximum net gamma measured 3.8 μR/hr (corrected for background and statistically 

tested against an NRC acceptance limit of 5 μR/hr).The average net gamma measured 0.46 

μR/hr. Based on the median value of exposure rate measurements in the vicinity of Building 

4027, the ambient background value for gross gamma was determined to be 9.09 μR/hr. All 

beta surface activity measurements made “for indication” showed no detectable activity. Based 

on the results of the interior survey of Building 4027, the conclusion was made that this area 

passed the NRC criteria at the time for unrestricted use. 

Within the storage area, the maximum net gamma measured 2.8 μR/hr (corrected for 

background and statistically tested against an NRC acceptance limit of 5μR/hr). The average 

net gamma measured –1.26 μR/hr. Based on the median value of exposure rate measurements 

in the vicinity of Building 4027, the ambient background value for gross gamma was determined 

to be 17.40 μR/hr. All beta surface activity measurements made “for indication” showed no 

detectable activity. Based on the results of the exterior survey of Building 4027 and the storage 

yard, the conclusion was made that this area was not contaminated and passed the NRC 

criteria at the time for unrestricted use. 

Because of limited information regarding the possible storage of sealed sources for gammaradiography in Building 4027, and the presence of staining indicative of a possible leak of 

unknown material from Building 4024 toward Building 4027 in 1980 aerial photographs, a 

preliminary MARSSIM Class of 2 has been assigned to Building 4027. 

6.2.3.2 Building 4027 Radiological Incident Reports 

There have been several incidents associated with Building 4027 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

1 EPA/HGL, Final HSA 5-A, p. 48, citation 209:

209 Busick, D.D., Internal Letter North American Aviation, Inc., Re: Water Supply, Building 027, Aug. 15, 

1963. 

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presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

6.2.3.3 Building 4027 Radiological Incident Report Summary: Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0023 8/13/1963 Building 4027 None Identified 

- Beta, 

Gamma, Alpha 

contamination 

identified

Approximately 30 people drank 

contaminated water from a bottle placed 

onto the water cooler, which had been 

filled with contaminated water drained 

from the S10FS-1.

• On August 13, 1963 an Atomics International employee notified the Health and Safety 

Department at SSFL that a bottle of water had been exposed to possible radioactive and 

beryllium contamination had been used for a drinking water supply in Building 4027. The 

contents of the bottle had been used to simulate NaK in the S10FS-1 system during 

acceleration tests in Building 4027. The incident is described in detail above.

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Section 6.2.4 

Building Number:

Building Alias: 

4032 

032/T032 

Building Name/Function: SNAP Thermal & Vacuum Environmental Test Facility

Component Testing / Radiography

Static Sodium Test Facility (SSTF)

Liquid Metal Fast Breeder (LMFBR) Sodium Tests

Clinch River Reactor Programs

Liquid Metal Development Laboratory

Inherently Safe Shutdown Systems (ISSS)

Radiological Uses & Incidents 

Notes: SNAP was a DOE-NASA project involving North American Aviation Space & 

Information Division, Rocketdyne. 

Radionuclides of Concern: Based on incident reports alone, Co-60 and Ir-192 would be the 

radionuclides of concern at Building 4032. 

6.2.4.1 Description of Operations & Processes 

Building 4032 was constructed between 1962-1965 in support of SNAP. It was used for 

component and instrumentation testing, and radiography. In 1975, the building housed small 

sodium test facilities associated with the “FFTF program,” associated with the Liquid Metal Fast 

Breeder Reactor (LMFBR) and Clinch River reactor programs. Between 1978-1983, radiological 

sources were used to determine the positioning of non-radioactive rods for use in developing 

the fuel rod control system in mock-up reactors. As of 1988, the building was still an active 

sodium test loop and sodium component/instrumentation test facility. It housed the Liquid Metal 

Development Laboratory (LMDL-1), which carried out applied research in support of liquid metal 

cooled reactors. The ISSS (Inherently Safe Shutdown System) Test Program was developed at 

Building 4037. 

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In 1996 during environmental remediation, approximately 15,390 pounds of sodium were 

removed from a below-ground drain tank. By 1998, the facility was being used for storage and 

component size reduction in support of water-vapor-nitrogen (WVN) cleaning.1 

The preliminary MARSSIM Classification for Building 4032 is Class 2 based on the building’s 

location within ETEC and incident history. 

6.2.4.2 Building 4032 Radiological Incident Reports 

There have been several incidents associated with Building 4032 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

6.2.4.3 Building 4032 Radiological Incident Report Summary: Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0299 3/27/1971 Outside Ir-192 Radiographer made exposure adjacent 

to building with employee inside the 

building

A0586 4/20/1974 Welded Pipe Ir-192 Unattached collimator allowed gamma 

graph source to fall out of guide tube

A0319 8/13/1981 Test Area Co-60 Gamma graph source failed to return and 

went undetected

A0098 2/16/1982 High Bay Ir-92 Off-Scale Dosimeter Dose

A0310 3/17/1987 High Bay None Identified Security, on rounds, entered radiation 

area of field radiography set up

1 EPA/HGL, Final HSA 5-A, p. 51, citations 223-229: 

223 ETEC Document, GEN-ZR-0013, “Radiological Survey of Buildings T049, T042, T027, T032, and T025,” 

August 26, 1988.

224 The Boeing Company, Rocketdyne Propulsion and Power DOE Operations Annual Site Environmental 

Report 1998, RD99-115, September 22, 1999.

225 Rockwell Aerospace, Rocketdyne Division Annual Site Environmental Report Santa Susana Field 

Laboratory and De Soto Sites 1993, RI/RD94-126, October 21, 1994.

226 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005.

227 ETEC Document, GEN-ZR-0013, “Radiological Survey of Buildings T049, T042, T027, T032, and T025,” 

August 26, 1988.

228 Rockwell International Corporation, Site Development Plan 1977-1981, United States Energy Research

and Development Administration Liquid Metal Engineering Center, June 1975.

229 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005. 

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• On March 27, 1971 during the performance of field radiography on pipe welds located just 

outside Building 4032, a Liquid Metal Engineering Center (LMEC) employee entered the 

building through a posted door. The employee was not observed by the radiographer who 

exposed the 27 Ci or Ir-192 in a “60 degree collimator” for a period of 11 minutes. The 

employee was discovered in the building by the radiographer after the source had been 

returned to the source safe. An investigation of the incident determine the LMEC employee 

was exposed to a maximum radiation level of 6 mR/hr (A0299).1

• On April 20, 1974 while gamma graphing pipe welds in Building 4032, the 65 Cu iridium-192 

(Ir-192) source was driven out for a routine exposure. The source went completely out the end 

of the guide tube and fell to the floor. When the remote hand crank was operated in an attempt 

to retrieve the source, it was discovered that the drive cable was disconnected from the hand 

crank mechanism. Recovery of the source was done by disassembling the guide tube on the 

remote hand crank assembly and exposing the end of the drive cable. The drive cable and 

source were pulled back through the guide tube by hand until the source was replaced in the 

source storage container. A contamination survey of the area, guide tubes, and the source 

storage container found no detectable contamination. The maximum radiation exposure 

received by personnel involved in the recovery was documented as 12 mrem as indicated by 

pocket dosimeter readings. The area was secured approximately 2 hours following the 

incident. (A0586).2

• On August 13, 1981 an employee conducted a field radiography of the “SASS-ACA” test 

article in Building 4032. After completing the fourth exposure, the employee retracted the 

source and set the brake on the control unit. He approached the source with his radiation 

monitor and detected no activity strength on his meter, which was set on the X1 scale. At a 

distance of approximately 4 feet from the exposed source, the meter went off-scale. The meter 

continued to go off-scale when the meter was switched to X-10 and X-100 scale. Following 

investigation, it was found the source “did not crank completely into the projector after the 

exposure. It has apparently hung up approximately 1/4 to 1/2 inch outside the projector in the 

guide tube. The monitor did not register a reading until being 4 feet from the exposed source 

because the projector orientation was such that it provided shielding from his approach 

direction. The source project contained approximately 56 curies of Co-60 (A0319). 3

• On February 16, 1982 an employee checked his dosimeter and found the gamma exposure 

meter to be off-scale while performing a fourth exposure on a test article in Building 4032. The 

1 EPA/HGL, Final HSA 5-A, p. 53, citation 236: 

. 236 Heine, W.F., Internal Letter Re: Radiography Incident Report A-0299, April 2, 1971. 

2 EPA/HGL, Final HSA 5-A, p. 53, citation 237: 

. 237 Klostermann, J.P., Rockwell International Internal Letter Re: Gammagraph Source Incident Building 

T032, April 24, 1974.
 

3 EPA/HGL, Final HSA 5-A, p. 53, citation 239: 

. 239 Breese, J.W., Rockwell International Internal Letter Re: Radiation Exposure Occurrence, August 20, 

1981. 

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employee shut down radiographic operations immediately. Investigation of the occurrence 

found the dosimeter to be “defective” (A0098).1 

• On March 17, 1987 security personnel entered Building 4032 while the building was being “XRayed.” The film badge of the employee was analyzed for exposure; however, the findings of 

the analysis are not presented in the incident report. The incident did not result in 

environmental contamination (A0310).2 

The discovery of a “defective dosimeter” (A0098) and failure to document worker dosimeter 

readings (A0310) following an exposure incident raises serious questions about the integrity of 

all worker dosimetry records. If personnel was unable to rely on functional dosimeters and if 

dosimeter readings were not recorded following exposure incidents, it is unlikely that reliable or 

relevant dose reconstructions can be achieved. These examples of poor and inconsistent 

worker monitoring practices support the determination of an expansive Special Exposure Cohort 

(SEC). 

1 EPA/HGL, Final HSA 5-A, p. 54, citation 240: 

. 239 Breese, J.W., Rockwell International Internal Letter Re: Off-Scale Reading of Personal Dosimeter, 

February 19, 1982. 

2 EPA/HGL, Final HSA 5-A, p. 54, citation 241: 

241 Unknown Author, Industrial Security Preliminary Investigation Report, Incident A-0310, March 17, 1987. 

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Section 6.2.5 

Building Number:

Building Alias: 

4036 

036/T036 

Building Name SNAP Offices 1-3 

Other Buildings: 4037 (037/T037)

4038 (038/T038) 

Notes: SNAP: DOE-NASA / Rocketdyne S&ID

Proximity to RMHF - Ambient exposure risk

Workers unlikely to have been monitored 

Radionuclides of Concern: The EPA research team did not find evidence that high activity 

radioactive materials were used or stored within Building 4036. However, its proximity to RMHF 

may have resulted in direct radiation and sky-shine from the RMHF, and affected ambient 

radiation conditions in the area. Radionuclides of concern at the RMHF include all radionuclides 

addressed in the background study plus any additional radionuclides identified during the EPA

HSA. EPA expressed concern about radioactive materials release and potential drainage from 

the RMHF, which may have resulted in residual contamination. 

6.2.5.1 Description of Operations & Processes 

Although this location is designated as a “non-radiological” location, its proximity to the 

Radioactive Materials Handling Facility (RMHF) and Buildings 4021 / 4022 presented a potential 

for direct radiation and sky-shine, which may have affected ambient radiation conditions in the 

area.1 In addition, releases of radioactive materials and drainage from the RMHF may have 

resulted in contamination. 

Building 4036 was constructed between 1962 and 1965 to support SNAP operations and 

included an ambient radiation dosimeter at the east side of the building. The presence of the 

ambient radiation dosimeter on the outside of the building suggests awareness of potential 

radiation in the area. However, given the building’s designation as a “non-radiological” location, 

it is likely that workers assigned to the location were not individually monitored. 

Employment Summaries provided for employees assigned to this location may reference the 

“non-radiological” designation of the building and the worker’s lack of dosimetry data to imply 

that the worker was not at risk of radiation exposure. However, the building’s proximity to the 

RMHF and potential for increased ambient radiation levels should be considered. 

1 EPA/HGL, Final HSA 5-A, p. 58, citation 268: 

.	 268 Unknown, Room and Area Numbering Plan Santa Susana Building 0036-0037, Snap Office Building, 

March 15, 1962. 

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A March 15, 1962 layout of the building shows there to be approximately 40 offices in the 

building.1 The building appears to have been demolished in 1999. Based on available 

information, it appears radiological controls were not required during demolition. The EPA

research team did not locate any radiological investigations at or near Building 4036. While 

there were no radiological burial or disposal locations identified in relation to Building 4036 

operations, EPA noted its proximity to the RMHF and Buildings 4021 and 4022. 

The preliminary MARSSIM Classification for the Building 4036 area is Class 2, due to its 

location within ETEC, proximity to Building 4021 and 4022 and the RMHF, and because no site 

investigation has been conducted.

1 EPA/HGL, Final HSA 5-A, p. 59, citation 276: 

.	 276 ETEC Document, GEN-ZR-0013, “Radiological Survey of Buildings T049, T042, T027, T032, and T025,” 

August 26, 1988. 

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Section 6.2.6 

Building Number: 4042 

Building Alias: 042/T042 

Building Name: SNAP Thermal Test Facility

SNAP Shield Casting Facility

SNAP Lithium Hydride Shield Fabrication

LMFBR Development Testing 

Notes: SNAP: DOE-NASA / Rocketdyne S&ID

“Non-Radiological” Designation

Workers may not have been monitored for radiation exposure

AEC licensed UO2 (15,000 lbs. of UO2 powder) at this location 

Radionuclides of Concern: Possible radionuclides at Building 4042 include U-235 and U-238. 

6.2.6.1 Description of Operations & Processes 

Building 4042 was constructed in 1963. The HSA research team could only located limited 

information regarding the building’s features. It contained approximately 4,100 square feet of 

laboratory space, and had 38’ ceilings. According to a 1988 plot plan, the building included three 

trenches (drainage could not be determined). Special features of this building included a lithium 

hydride casting furnace. The building also had other environmental chambers for testing in inert 

and vacuum environments. Building 4042 also contained facilities used for the development of 

an alcohol cleaning process for the removal of sodium from large components. 

Building 4042 is identified as a SNAP program thermal and structure test facility, identified as a 

non-nuclear and test support facility. The SNAP Shield Casting Facility was used as a general 

test and lithium hydride shield fabrication building in support of the SNAP program. It was also 

used for sodium aerosol and related technology tests. After support work for the SNAP tests 

ceased, Building 4042 was used for testing liquid metal systems, also known as LMFBR 

Development Testing. 

In 1973, the AEC SNAP Project Office granted Atomics International permission to use Building 

4042 and the associated equipment for loading UO2 into a “lower axial blanket shield.” While the 

permission letter does not provide the amount of UO2 used, the authorization permitted 15,000 

pounds of uranium in the form of UO2 powder. Permission was granted with the understanding 

that the work would not interfere with the SNAP Program closeout and the equipment and 

building would be cleaned of any UO2 deposited during the operation.1 Additional information 

regarding these operations could not be located. 

While the 1986 plan of long-range D&D of facilities at SSFL identifies hazardous waste as being 

present at this location, no radiological waste is referenced. However, a December 20, 1989 

1 EPA/HGL, Final HSA 5-A, p. 114, citation 616 & 617: 

616 Stamp, S.R., Letter Re: Temporary Use of Building 4042 at Santa Susana, February 14, 1973.

617 Rockwell International Document, Use Authorization 62, February 6, 1973. 

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report titled, “Nuclear Operations at Rockwell’s Santa Susana Field Laboratory - A Factual 

Perspective” identifies Building 4042 as having contained a radioactive test loop.1 

The EPA HSA research team did not located any radiological incident reports involving Building 

4042, but did reference an incident involving mercury on February 24, 1965. 

The preliminary MARSSIM Classification for the Building 4042 area is Class 1 due to its location 

within ETEC, the possible unconfirmed use of uranium in the building, the unknown nature of 

the materials that entered the trenches located within the building, and the presence of standing 

water south of the building.

1 EPA/HGL, Final HSA 5-A, p. 145, citation 619: 

619 Rockwell International, N001ER000017, Nuclear Operations at Rockwell’s Santa Susana Field 

Laboratory – A Factual Perspective, December 20, 1989. 

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Section 6.2.7 

Building Number:

Building Alias: 

4057 

057/T057 

Building Name SNAP Launch Handling (DOE-NASA)

Mobile Equipment Development (DOE-NASA)

Liquid Metal Engineering Center (LMEC) Lab

Static Sodium Test Facility

NASA: International Space Station Development

ETEC General Test Building 

Notes: DOE-NASA / Rocketdyne S&ID

Proximity to RMHF - Ambient exposure risk

Proximity to Building 4059 - Radionuclide migration risk

Workers may not have been adequately monitored

Tritium 

Radionuclides of Concern: U-238, U-234, U-235, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, 

Cs-137, Sr-90, H-3, Fe-55, Co-58, Co-60, Ni-63, Ba-133, Eu-152, Eu-154, Eu- 155, Pm-147, 

Ta-182 

6.2.7.1 Description of Operations & Processes 

Building 4057 was constructed in approximately 1963 to house two sodium test rigs. It was used 

as the Launch Handling and Mobile Equipment Development Building, Liquid Metal Engineering 

Center (LMEC) Laboratory, and the Static Sodium Test Facility Building. In 1975, it was used for 

the research and development of the International Space Station (ISS) power generation 

system, before becoming the ETEC General Test Building. It was decommissioned in 1998 for 

laboratory use, and became a records room. 

This location further illustrates shifting building uses and joint projects between DOE and NASA, 

a characteristic of SSFL site operations at various locations throughout Areas I, II, III and IV. 

Currently, Building 4057 is used to store Boeing records. 

Because the building was used as a research laboratory, there is a possibility that radioactive 

materials may have been used. In addition, Building 4057 is located approximately 130 feet 

south of Building 4059; consequently, there is some potential that other radionuclides 

associated with Building 4059 may have drained / migrated to the area surrounding Building 

4057, which include U-238, U-234, U-235, Pu-238, Pu-239, Pu-240, Pu- 241, Am-241, Cs-137, 

Sr-90, H-3, Fe-55, Co-58, Co-60, Ni-63, Ba-133, Eu-152, Eu-154, Eu- 155, Pm-147, Ta-182. 

The preliminary MARSSIM Classification for Building 4057 is Class 1 due to its location 

within ETEC, close proximity to SNAP reactor Building 4059, and because no site 

investigation has been conducted. 

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6.2.7.2 Building 4057 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0164 02/26/2003 Cabinet Drawers Six folders in cabinet drawers contain air/

wipe samples with low-level 

contamination.

• In 2003, Boeing found 16 folders containing air/wipe samples in cabinet drawers. Six were 

found to contain low levels of residual contamination, but this contamination had not escaped 

from the envelopes containing the filters. The majority of the contaminated samples measured 

less than the NRC Regulatory Guide 1.86 release limit of 1,000 dpm/100 cm2 for removable 

contamination. Based on the survey results, the incident was deemed to be an insignificant 

hazard.1 However, the incident report did not identify specific radionuclides tested for, and did 

not speculate on the source of the contaminated air/wipe samples. 

1 EPA/HGL, Final HSA 5-C, p. 33, citation 1: 

1 The Boeing Company, Incident Report No. 01684, February 26, 2003. 

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Section 6.2.8 

Building Number:

Building Alias: 

4228 

228/T228 

Building Name/Function: (SNAP Critical Test Facility, Formerly Bldg. 4012)

SCTI Power Pak Facility

SCTI Co-Generation Plant 

Notes: “Non Radiological” designation

Formerly a Radiological Facility 

Associated Buildings: Building 4708 Electrical Substation

Building 4807 / 4808 Electrical Equipment Pads

Building 4809 Air Blast Heat Exchanger

Building 4710 SCTI Power Pak Cooling Tower 

Radionuclides of Concern: The HSA research team did not find evidence that radioactive 

materials were used in Building 4228. However, radionuclides resulting from former reactor 

operations in Building 4012, which supported the northwest end of Building 4228, include 

Am-241, Cs-137, Pu-238, Pu-239, Pu-240, Pu-241, Sr-90, U-234, U-235, U-238. In addition, 

H-3 due to Building 4012 operations. 

6.2.8.1 Description of Operations & Processes 

Building 4228 was constructed in the early 1980’s as the Sodium Components Test Installation 

(SCTI) Power Pak Facility. The northwest end of Building 4228 was built on top of the SNAP

Critical Test Facility, Building 4012. 1 Because the new building was designated as “nonradiological,” workers did not have to wear dosimetry badges. However, as referenced, this 

building was constructed on top of the former SNAP Critical Test Facility Building 4012, where 

numerous radiological incidents were documented in the Health Physics Dept. Log Book. For 

more information on the operations of Building 4012, please see its section in this document. 

The HSA provides a detailed description of the building’s features and schematics. 

Building 4228 has been identified as the Power Pak Facility and the SCTI Co-Generation Plant. 

The SCTI Power Pak Facility was designed to harness the steam produced through SCTI’s 

sodium experiments, and to generate commercial electric power. The system operated from 

1988 through 1993. 

The preliminary MARSSIM classification of Building 4228 is Class 1 because Building 4228 is 

located on top of and south of former SNAP Critical Test Facility Building 4012, which is also 

identified as Class 1.

1 EPA/HGL, Final HSA 5-A, p. 52, citations 1 & 2: 

1 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. L-21.

2 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. 

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7.0 Civilian Nuclear Power Test Facilities 

According to the 2006 Site Description, critical test facilities supported the development of 

civilian nuclear power and included Buildings 4009 and 4100. 

The 2006 Site Description provides an incomplete definition of operations, processes, and 

incidents associated with Building 4009, including the second Van de Graaf Particle Accelerator 

used in Building 4009. 

The 2006 Site Description vaguely referenced the CAT-Scan located in Building 4100 and used 

between 1980-2008 to inspect rocket engine components. It should be noted that the CATScan, (the largest in the world) was operated by NASA and that Building 4100 (along with 

several other Area IV locations) involved NASA operations. Space Shuttle Main Engine (SSME) 

and International Space Station (ISS) operations occurred in Building 4100, which involved 

NASA aerospace employees of Rockwell (DOE contractor) typically associated with Area I, II 

and III work locations. NASA operations in Area IV, inclusive of the SNAP program, often 

required the presence and participation of DOE-contracted employees often thought to be 

disqualified from EEOICPA, based solely on the location of assigned “time clocks” in Areas I, II 

and III. The inclusion of all employees to EEOICPA - Areas I-IV - is strongly supported. 

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Section 7.1 

BuildingNumber:

Building Alias: 

4009 

09 / T09 

Building Name: Organic Moderated Reactor (OMR)

Sodium Graphite Reactor (SGR)

SGR “Hot Lab”

Van de Graaf Particle Accelerator

Engineering Development Facility 

High Energy-Rate Forging (HERF) Program 

Building Function: Reactor Operations / Criticality Testing / Hot Lab Activities 

Notes: Enriched uranium and thorium were stored in the OMR and SGR fuel vaults and were 

used in the two critical assembly rooms during their operation. Up to 800 pounds of depleted 

uranium were stored in the OMR counting room under the Accident Debris program. In addition, 

a Van de Graaff accelerator was housed in the SGR graphite storage area during the early 

1960s. In the 1980s, the OMR high bay was later used for HERF, which included the handling of 

depleted uranium. In addition: Tritium. 

Radionuclides of Concern: Decay products from U-235, U-238, and Th-232 decay chains 

include Th-234, Th-228, actinium-228 (Ac-228), radium-226 (Ra-226), lead- 214 (Pb-214), 

bismuth-214 (Bi-214), Pb-212, Bi212, and thallium-208 (Tl-208). In addition, H-3, Sr-90 and 

Cs-137, Na-24, Co-57, Co-60, europium-152 (Eu-152), and americium-243 (Am-243) would 

have been formed. 

7.1.1 Description of Operations & Processes 

Building 4009 was constructed in two phases to first house the Organic Moderated Reactor 

(OMR) and later, the Sodium Graphite Reactor (SGR). The OMR and SGR facilities were pilot 

plants for developing large nuclear reactor power plants for commercial power generation, such 

as the Hallam nuclear power plant in Nebraska and the Piqua nuclear power plant in Ohio. 

The OMR was used for testing uranium fueled reactors moderated and cooled by organic 

liquids. The critical assembly core used “slightly” enriched uranium fuel in a heterogeneous 

organic moderated lattice. The degree of uranium enrichment that “slightly” represents is 

unclear. Various types and configurations of fuel elements and core geometries were tested. 

The SGR was used to determine the operating characteristics of reactors with cores cooled by 

sodium and moderated with graphite. The basic critical assembly of the SGR was a cylindrical 

array of hexagonal graphite cylinders into which various amounts and configurations of fuel and 

sodium (in cans or simulated by aluminum) could be inserted. The fuel used in the SGR was 

uranium metal and uranium carbide enriched up to 2.5%. 

Both the OMR and SGR reactors were low-powered (less than 200 watts) critical assemblies. 

The OMR and SGR operated from 1959 and 1960, respectively, until 1965 and 1967, 

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respectively. 1 In addition, as-built Drawing No. 303-009-E17 indicates that a Van de Graaff 

generator was installed in the SGR graphite storage area in August, 1960 and was moved to 

Building 4030 in about May, 1962. 

According to Rockwell International’s Final Decontamination and Radiological Survey 1990 

report, after the OMR and SGR programs were terminated in 1967, all materials from the two 

critical assemblies were removed and transferred to “other laboratories” or were disposed of at 

“authorized sites” in the late 1960’s and early 1970’s. In June, 1966 Atomics International 

reported that it was sending OMR irradiated fuel to the Savannah River site for “recovery.”2 

Building 4009 was designated the Engineering Development Facility after decommissioning in 

the early 1970’s. Sodium fire experiments intentionally exposing air to sodium were conducted 

in the OMR high bay, for finding new ways of extinguishing sodium fires. Also, during this time 

period, up to 800 pounds of depleted uranium were stored in the OMR counting room under the 

Accident Debris Program. The depleted uranium was received from the Los Alamos National 

Laboratory.3 

The LANL depleted uranium was shipped off-site in the early 1990’s. From the early 1980’s until 

the 1990’s, the former SGR high bay was used for the storage and underwater testing of 

Rocketdyne’s In-Service Inspection (ISI) equipment, which was used for inspecting commercial 

power reactors off-site. This equipment sometimes became contaminated off-site with low levels 

of radioactivity. Containers of contaminated equipment were stored in a “controlled area” in 

Building 4009.4 In May 1983, Rockwell ultrasonically inspected a contaminated and activated 

pipe specimen using ISI equipment.5 

In October 1988, Rockwell requested a new authorization for the use of radioactive materials 

and radiation producing devices for the forging of depleted uranium in the former OMR critical 

assembly room. This was known as the high-energy rate forging (HERF) program, conducted 

under LANL; it was allowed under sub-item F of the State of California license, which authorized 

30,000 pounds of source material in any form for research and development. 

1 EPA/HGL, Final HSA 8, p.26, citation 3: 

3 Energy Technology Engineering Center, Organic Moderated Reactor and the Sodium Graphite Reactor, at 

www.etec.energy.gov/History/Major-Operations/Organic-moderated.html. 

2 EPA/HGL, Final HSA 8, p.26, citation 4: 

4 Internal letter from V. J. Schaubert to F. W. Schlapp, Re: OMR Irradiated Fuel Shipment to Savannah

River, dated June 24, 1966. 

3 EPA/HGL, Final HSA 8, p.26, citation 5: 

5 Memorandum from R. S. Frazier, Rockwell International, Re: LANL DU Received at T009, dated 

September 22, 1989. 

4 EPA/HGL, Final HSA 8, p.27, citation 1: 

1 Rockwell Internal Letter from W. R. Johnson to W. E. Nagel, Re: Request for Radioactive Material and 

Radiation Producing Device User Authorization for ISI Operations at Building T009, June 10, 1987. 

5 EPA/HGL, Final HSA 8, p.27, citation 2: 

2 Rockwell Internal Letter from R. J. Tuttle to Isotope Committee, Re: Short-Term Approval for Work with 

Radioactive Sample in Building 009, Santa Susana, May 9, 1983. 

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HERF involved a series of approximately 114 depleted uranium blanks (some alloyed with 

titanium) of 1-inch in diameter, inserted into the HERF press that used compressed air to 

accelerate a piston that hammered the blanks at approximately 300  ̊C. Reportedly, this process 

did not involve machining, sawing, filing or other processes that produced fines. All material was 

stored in wooden boxes in the former OMR critical assembly room, with a maximum of 1,000 kg 

of depleted uranium in Building 4009 at any one time. An inventory of shipping drums inidicates 

that 120 drums of uranium and depleted uranium were shipped from the EG&G facility at Rocky 

Flats, Colorado to building 4009 for the HERF project. The material was to be returned to Rocky 

Flats at the conclusion of the study.1 

A Rockwell International internal letter, dated April 20, 1990 indicates that the HERF program 

has ended and that the former OMR critical assembly room is needed for the ISI program to 

store three cargo containers of radioactive equipment; each container having an external dose 

rate of approximately 2mrad/h.2 An internal letter dated November 13, 1990 indicates that an 

effort is underway to upgrade Building 4009 to make it suitable for office use, which includes 

moving the HERF equipment from the former OMR critical assembly room to storage.3 In 1995,

Rockwell reported plans to upgrade Building 4009 to support a new proprietary sensor program, 

which would involve the high bays and removal of half the existing room, replaced by a new 

rolling roof, plus other installations. The use authorization licenses for Building 4009 were 

changed 102 times, before Building 4009 was released for unrestricted use in 1999. 

In September, 1969 North American Rockwell conducted an analysis of the levels of induced 

radioactivity in the SGR structural steel and related components to assess the potential hazards 

associated with the demolition of the facility. Cobalt-60 (Co-60) was the only gamma emitting 

radionuclides detected in the steel with a high reading of 16 dpm/gram of steel. 

In 1988, Rockwell performed a radiological survey to identify areas needing further radiological 

inspections or requiring remedial action. The survey covered the interior of the OMR side of the 

building, the SGR radioactive liquid hold-up tank and pit, the old sanitary leach field, and grease 

/ sludge from sink clean-outs, shower drains, and matching equipment. A few locations on the 

OMR side of the facility were contaminated at levels below Rockwell’s acceptance limits. The 

OMR fuel vault was contaminated with a maximum alpha activity of 92 dpm/100 cm2 while 

Rockwell’s “acceptable limit” in 1988 was 5,000 dpm/100cm2. 

1 EPA/HGL, Final HSA 8, p.27, citations 4 & 5: 

4 Rockwell Internal Letter from P. H. Horton to L. J. Auge, Re: Start Up of High Energy Rate Forge (HERF) 

Depleted Uranium (DU) Operations at Building 009, SSFL, September 29, 1989.

5 Schaubert, V. J., Nuclear Materials Management Plan for the HERF Program, Rockwell International 

Report No. N001NMP000003, October 31, 1989, pp. 3-4. 

2 EPA/HGL, Final HSA 8, p.27, citation 6: 

6 Rockwell Internal Letter from J. M. Harris to R. Tuttle, Re: Temporary RA Equipment Storage – T009, April 

20, 1990. 

3 EPA/HGL, Final HSA 8, p.27, citation 7: 

7 Rockwell Internal Letter from F. C. Schrag to P. H. Horton, Re: Plans and Work in Progress for Building 

009, November 13, 1990. 

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Maximum removable activity was 15 dpm/100 cm2 while the acceptable limit was 1,000 dpm/

100cm2. A sludge sample collected from inside the inactive SGR radioactive liquid hold up tank 

contained fission products, uranium-238 (U-238) (18 pCi/g), thorium-232 (Th-232) (25.2 pCi/g), 

and possibly U-235 (0.81 pCi/g). No contamination was reportedly detected in the hold-up tank 

pit. Rockwell recommended that the tank and drain lines leading to it be removed and disposed 

of as radioactive waste. The OMR hold-up tank was not included in the survey. It is unclear why 

this was not surveyed, but the HERF program was in operation in the OMR high bay at the this 

time. Ambient exposure data in the area of the old sanitary leach field (partially paved) showed 

a mean exposure rate of 13 μR/h. In late 1989 and early 1990, all contaminated material was 

packaged and initially moved to the RMHF.1 

The HSA contains a detailed chronology of radiological investigations and decommissioning/

cleanup of Building 4009. In 2002, EPA contractor Tetra Tech EM, Inc. conducted oversight, 

sampling, and an independent technical review of documents for Building 4009. Tetra Tech 

noted that because of Boeing’s security considerations, Building 4009 had not been accessible 

to EPA or Tetra Tech to conduct confirmation surveys. Tetra Tech reviewed the results from 

Rockwell’s 1988 survey of the OMR and noted that rockwell had collected over 280 swipe 

samples in the SGR portion of Building 4009, and all results were below 5,000 dpm/100 cm2 for 

beta-gamma activity, and were below 500 dpm/100 cm2 for alpha activity. Tetra Tech reported 

that Building 4009 had been released from radiological controls by CDPHE and recommended 

no further survey action.2 

The preliminary MARSSIM classification for the Building 4009 area is Class 1, due to its former 

use as housing two-critical assembly test facilities. 

7.1.2 Building 4009 Radiological Incident Reports - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0378 07/05/1961 SGR Fuel Vault Uranium 

Carbide

Can of powered UC ignited when 

opened.

A0372 06/11/1964 OMR Critical 

Assembly Room

Tritium - H3 Tritium target material for accelerator 

changed without monitoring.

A0157 06/05/1986 High Bay (does not 

specify OMR or 

SGR)

Am-241 Smoke detector source leaked after 

hose-down by the fire department

• On July 5, 1961 Atomics International personnel in the SGR fuel vault noticed that two 1-

gallon containers of uranium-carbide fuel slugs were bulging at the top and bottom. Atomics 

International personnel transferred the fuel to the SGR Hot Lab vented hood. (It was 

1 EPA/HGL, Final HSA 8, p.31, citation 6: 

6 Harris, J. M., SGR Liquid Holdup Tank Decontamination and Decommissioning, T009, Rockwell 

International Report No. N001DWP000025, August 1, 1989, pp. 3, 14. 

2 EPA/HGL, Final HSA 8, p.33, citation 5: 

5 Tetra Tech EM, Inc., Final Rocketdyne Technical Support and Field Oversight Document Review for 

Buildings T009, T011, T019, T055, and T100, December 20, 2002, pp. 5-7. 

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previously learned that this fume hood had HEPA filters). The Fire Department was notified to 

observe the transfer. The container with the bigger protrusions was placed in a 5-gallon can 

with its lid secured tightly for the transfer from the vault to the Hot Lab hood. Then, the 1-

gallon fuel container was removed from the 5-gallon container in the vented hood. When its lid 

was removed in the vented hood, a violent reaction and fire occurred with the release of 

gases. Uranium powder on 12 fuel slugs ignited, but the fire was put out by smothering it with 

G-1 powder. The second can was opened in the same manner with a less violent release of 

pressure. This can contained 18 slugs. All of the slugs were subsequently transferred to pans 

of cutting oil in the Hot Lab hood. According the Atomics International, all contamination was in 

the SGR Hot Lab hood; there was no damage to personnel, equipment, or property (A0378).1

• On June 11, 1964 between 3 and 4 curies of tritium target material were replaced in the OMR 

critical assembly room without the required monitoring by the Atomics International Health and 

Safety Department. The incident report states that the target material was for the Department 

741 accelerator and that an outside vendor for the accelerator may have been involved. The 

employee was subsequently informed to contact the Health & Safety Department prior to any 

operations involving the disassembly or replacement of tritium targets (A0372).2

• On June 4, 1986 the high bay was washed down with a fire hose. This incident report does not 

state whether the location was in the OMR or SGR, but both high bays had sumps connected 

to individual radioactive hold-up tanks connected to the leach field. A bag placed over one of 

the smoke detectors was not sealed and filled with water during washdown activities. Plant 

Services was called to repair the damage to the detector. When an electrician empties the 

bag, water fell onto his hands, which subsequently registered 120 cpm direct reading of alpha 

contamination. The electrician was decontaminated. A gamma scan of both of his hands 

showed a very small amount of Am-241 (4.5E-4 microcurie) (450 pCi). This was considered 

“acceptable” by Atomics International, Health Physics Department, and the electrician was 

released. It was recommended that on future occasions, the Radiation & Nuclear Safety 

Branch be consulted to ensure that precautions are adequate to perform the work (A0157).3

1 EPA/HGL, Final HSA 8, p.29, citation 1: 

1 Atomics International Inter-office Letter from J. P. Klostermann to E. C. Hickey re: Incident in Building 009 

OMR on July 5, 1961, dated July 13, 1961. 

2 EPA/HGL, Final HSA 8, p.29, citation 2: 

2 Atomics International Internal Letter from E. E. Owens to R. M. Hill, Re: Incident in Building 009 OMR on

June 11, 1964, dated June 17, 1964. 

3 EPA/HGL, Final HSA 8, p.29, citation 3: 

3 Rockwell International Radiological Safety Incident Report from R. McGinnis to J. Chapman, dated June

10, 1986. 

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Section 7.2 

Building Number:

Building Alias: 

4100 

100 / T100 

Building Name: Advanced Epithermal Thorium Reactor (AETR)

Fast Critical Experiment Laboratory (FCEL)

Radiation Safety / Computed Tomography Laboratory

NASA Space Shuttle - Int’l Space Station 

Building Function: Reactor configuration tests: Thorium, Uranium, Fast-Neutron 

Notes: • DOE-NASA*

* Computerized Tomography = Inspection of Rocket Engine Components

Drainage to Area III Silvernale Pond / Area II R-2 Pond

• Tritium / Neptunium 

Radionuclides of Concern: U-233, U-234, U-235, U-236, U-238, Th-232, Np-237, Pu-238, 

Pu-239, Pu-240, and Pu-241 in the AETR and in January 1972 for use in the FCEL. Decay 

products would include Th-228, Ra-228, Th-230, Ra-226, Pb-210, Pa-231, and Ac-227. 

7.2.1 Description of Operations & Processes

Building 4100 was constructed after 1962 for the Southwest Atomic Power Association; that use 

was terminated in 1974. It housed the Advanced Epithermal Thorium Reactor (AETR), which 

was a separable-half critical experiment operating at less than 200 watts (thermal), Fast Critical 

Experiment Laboratory (FCEL), and Radiation Safety and Computed Tomography Laboratory. 

Twenty reactor core configurations were studied including thorium, uranium, and later, highenergy fast neutrons in the FCEL. The laboratories were used for radioactive sample counting 1

and instrument calibration.

From 1980-2008, the high bay was used for high energy Computer Aided Tomography (CAT) 

scanners in the world - owned by NASA. NASA programs conducted at SSFL after 1980 

included the Space Shuttle Main Engine (SSME), and the International Space Station (ISS). 

Due to NASA programs that relied on Rocketdyne Space and Information Division personnel 

associated with Areas I, II and III it is likely that worker rotation into Area IV occurred on a 

routine basis, with a lack of documentation. 

Building 4001 contained an experimental critical assembly located within a high bay (Room 

110), special nuclear material stored within a vault (Room 112), supporting rooms, and a control 

room (Room 109). Reportedly, racks in Room 112 were spaced accordingly to avoid criticality 

from stored items. 

1 EPA/HGL, Final HSA 5-C, p.47, citation 3: 

3 U.S. Nuclear Regulatory Commission, Rockwell International Corporation Docket No. 50-147, Order

Terminating Facility License, October 1, 1980. 

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Photographs of Building 4100 from 1960 show the presence of debris and drums in a trench 

(80’x20’), which was used by contractors for burning and disposal of construction debris and 

possible hazardous substances. 

The preliminary MARSSIM Classification of Building 4001 is Class 1 because of its former use. 

7.2.2 Building 4100 AETR Radiological Incident Reports 

There have been two incidents associated with Building 4100 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

7.2.3 Building 4100 Radiological Incident Report Summary - Data Provided by Boeing 

*Incidents are written as they are presented in the incident database. The HSA research team believes that MFP is an 

acronym for Mixed Fission Products. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0217 12/17/1991 Respirator 

Laboratory

MFP Respirator lab Washington Machine 

contaminated

A0220 10/23/1992 CAT X-Ray Cell Contractor service rep. activated X-Ray 

system while employee was in cell.

• On December 17, 1991 Rockwell performed a radiological survey on the internal respirator 

laboratory washing machine in Building 4100. A smear wipe indicated 173 dpm/100cm2. 

Gamma spectroscopy of a 4,000-second scan indicated 254 dpm/100cm2. Rockwell 

recommended that the respirator laboratory decontaminate and routinely survey the washing 

machine on a monthly basis (A0217).1

• On October 23, 1992 a contract service technician activated NASA’s 2.5 MVp CAT X-ray 

system while the cell was occupied by another individual. This was in violation of Rocketdyne 

procedure and California state law. The individual’s exposure was calculated to be 

approximately 5 mrem. Rockwell recommended that operating personnel be reminded to 

inspect the cell area thoroughly prior to activation of an X-ray system (A0220).2 

1 EPA/HGL, Final HSA 5-C, p.48, citation 2: 

2 Saba, V. B., Rockwell Internal Letter to Radiation Protection and Health Physics Services, re: Radiological

Safety Report, Contaminated Respirator Laboratory Washer, December 17, 1991. 

2 EPA/HGL, Final HSA 5-C, p.48, citation 3: 

3 Barnes, J., Rockwell Internal Letter to Distribution, re: CAT Scanner Operation with Individual in Cell, 

October 23, 1992. 

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8.0 Nuclear Support Operations 

According to the 2006 SSFL Site Description, beginning in 1956 several operations were 

conducted in Area IV to support nuclear programs. These operations included the manufacture, 

management, and disassembly of fuel for reactor operations as well as the operation of nuclear 

waste management facilities for off-site disposal. 

However, it should be noted that not all disposal of radioactive materials occurred off-site. Onsite disposal (through a variety of methods) is well documented. Onsite disposal of nuclear and 

non-nuclear materials included incineration, burning in open burn-pits, burial, combustion, 

dispersal, and submersion into unlined ponds. In addition, the 2006 Site Description does not 

reference the Site-Wide Water Reclaim System constructed by the AEC and Atomics 

International, which carried waste water, industrial effluent, and reactor blowdown to open 

reclaim ponds used by Area I, II and III personnel. 

The 2006 SSFL Site Description indicates that, with the exception of the Radioactive Materials 

Handling Facility (RMHF), the Fuel Storage Facility (FSF), and the Radiation Instrument 

Calibration Laboratory, these nuclear support operations were terminated in 1988. 

Several buildings and facilities associated with the RMHF have been excluded from the 2006 

Site Description; among them one of the most significant contributors to onsite ambient 

radioactivity, the Building 4664 Low-Level Radioactive Waste Incinerator. Descriptions of the 

buildings and facilities associated with the RMHF, which are in need of inclusion to the Site 

Profile, can be found in Section 16.0. 

8.1 Reactor Fuel Manufacturing 

According to the 2006 Site Description, as part of the nuclear reactor development work 

performed for the government, three different reactor fuel manufacturing operations occurred at 

SSFL in Buildings 4003, 4055, and Building 4064. 

The first operation was the assembly of fuel elements for the Sodium Reactor Experiment 

(SRE), the second was a plutonium fuel manufacturing facility, and the third was a uranium 

carbide fuel pilot plant. There was also a Fuel Storage Facility, used to store the Special Nuclear 

Materials (enriched uranium and plutonium) used to make the fuels. 

However, the 2006 Site Description has excluded several operations and programs associated 

with Buildings 4003, 4055 and 4064. Building 4005 (UCPFF) alone engaged in numerous 

processes that were excluded from the 2006 Site Description, including the Apollo (NASA) 

Counterweight Coating Program, Molten Salt Test Facility (Coal Gasification), and other coal 

processes. 

As a result, relevant data associated with incidents at these locations, including environmental 

releases and worker exposure, have been excluded from the SSFL Site Profile. Descriptions of 

additional buildings and facilities can be found in Section 16.0.

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Section 8.2 

Building Number:

Building Alias: 

4003 

03 / T03 

Building Name: Engineering Test Building (ETB) “Hot Cave” 

Building Function: Fuel Element Assembly (SRE)

Molten Uranium Research

Fuel Decladding

Irradiation Experiments

SNAP Fuel Analysis & Corrosion Tests

Other Related Buildings: Substation Building 4693

Support Building 4825 

Notes: Tritium 

Radionuclides of Concern: Uranium, thorium, transuranic elements, mixed fission products, 

and activation products. Possible radionuclides include natural and enriched uranium (U-238, 

U-234, U-235, U-236), isotopes of plutonium (Pu-239, Pu-240, Pu-241, Pu-242), americium-241 

(Am-241), thorium-228 (Th-228), Th-232, Th-234, activation products (tritium (H-3), carbon -14 

(C-14), sodium-22 (Na-22), Na-24, chromium-51 (Cr-51), manganese-54 (Mn-54), nickel-59 

(Ni-59), Ni-63, iron-59 (Fe-59), cobalt-60 (Co-60)) and fission products (krypton-85 (Kr-85), 

strontium-89 (Sr-89), Sr-90, antimony-125 (Sb-125), iodine-129 (I-129), I-131, cesium-134 

(Cs-134), Cs-137, cerium-144 (Ce-144), barium (lanthanum)-140 (Ba (La)-140), niobium-95 

(Nb-95), ruthenium-103 (Ru-103), Ru-106, xenon-133 (Xn-133), Xe-135, promethium-147 

(Pm-147), samarium-151 (Sm-151), europium-152 (Eu-152), radium-226 (Ra-226),

actinium-228 (Ac- 228). 

8.2.1 Description of Operations & Processes 

Building 4003 was constructed in 1957.1 It was known as the Engineering Test Building (ETB) 

and the “Hot Cave.” Between 1957-1964 it was used to assemble fuel elements for the SRE. In 

this operation, uranium and thorium metal slugs were loaded into metal tubes, the remaining 

tube space was filled with sodium and the tubes were sealed. The Hot Cave portion of the 

building was designed and constructed to investigate the chemistry of molten uranium, and to 

study the separation of fission products and plutonium from uranium systems. 

The HSA provides aerial photographs and a detailed description of building features and 

specifications. Building 4003 contained offices, a change room, several laboratories, a Freon 

1 EPA/HGL, Final HSA 6, p.12, citation 1: 

1

ETEC website at: www.etec.energy.gov/History/Major-Operations/SRE.html 

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compressor, a concrete pit, a tank pit, a sump, an expansion tank, a boiler, and a cooling tower.1 

It should be noted that because Building 4003 contained offices, employees that may have been 

assigned to the offices may have been considered “non-radiological” employees. However, 

given the operations at the facility, such a designation may grossly misrepresent their exposure 

risks. 

Building 4003 also contained an exhaust system that included nine fume hoods, valves, ducts, 

blowers, filters, filter plenum chambers, and a stack. In addition, Building 4003 contained two 

radioactive waste sinks connected to 5-gallon bottles, and a highly shielded area designed for 

remote manipulation of radioactive materials, known as the “Hot Cave.” The Hot Cave contained 

identical east and west “hot cells,” a pair of mechanical manipulators for each of the two cells, a 

transfer tunnel, liquid waste lines, east and west cell “small” radioactive liquid waste holdup 

tanks located below the Hot Cave, and a hot cell exhaust system with HEPA filters and two 

blowers located on the roof of Building 4003. 

The west hot cell of the Hot Cave was used primarily for kilogram-scale processing experiments 

involving chemical reactions of irradiated reactor fuels at temperatures up to 1,800 ̊ C. The east 

hot cell was used primarily for mechanical operations, such as decladding, component 

disassembly, and inspection. The transfer tunnel was used for sample handling, waste removal, 

inter-cave transfer, and the storage of kilocurie sources.2 According to DOE, later activities 

involved bench-scale research into reprocessing of used (or irradiated) fuel assemblies. The 

research involved the removal of fission products from used fuel.3

Building 4003 was also used for the analysis of Systems for Nuclear Auxiliary Power (SNAP) 

fuel burn-up samples and the evaluation of irradiation experiments. Use of the Hot Cave ended 

when the SNAP program was terminated in 1973. 

The 2006 Site Profile indicates that “some traces of radioactivity were later found in the drain 

line, which was removed.” However, according to Atomics International (1973), the inner 

surfaces of the Hot Cave were found to be “grossly” contaminated with mixed fission products. 

According to the HSA, containment of this high-level contamination required continuous 

operation of the radioactive exhaust system as well as filtering and sampling activities. 

Radioactive contamination was found to be on the internal surfaces of the hot cells, in the liquid 

waste lines, the liquid waste holdup tanks, the radioactive exhaust system ducting and on filter 

1 EPA/HGL, Final HSA 6, p.12, citations 3 & 4: 

3 Montgomery Watson Harza, Table 1, List of Historical Document Map Features at SRE, August 1, 2003.

4 Rockwell internal letter from C. D. Bingham to W. F. Heine, Re: Application to Perform Radiochemical 

Operations in Building 003A (ETB Annex), December 10, 1970. 

2 EPA/HGL, Final HSA 6, p.13, citation 10: 

10 Strausberg, S., Gardner, W. J., Guon, J., Luebben, T. E., and Mills, T. H., Modified Hot Cave Facility for

Reprocessing Experiments, Atomics International Report No. NAA-SR-2687, June 30, 1958, p.3. 

3 EPA/HGL, Final HSA 6, p.14, citation 1: 

1 ETEC website at: www.etec.energy.gov/History/Major-Operations/SRE.html. 

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plenums. In 1974, the building 4003 facilities were declared “excess” and a facilities dismantling 

plan was prepared.1 

Atomics International informed the AEC that the alpha-emitting radionuclides present were from 

enriched uranium; some transuranics may also have been present. The beta-gamma emitting 

radionuclides resulted from fission products and stainless steel activation products.2 Atomics 

International commenced D&D in January 1975 and ended in June 1975. Dismantling of the Hot 

Cave was completed in April 1975. 

After 1961, Building 4003 was connected to a site-wide (Areas I, II, III, IV) sewage treatment 

system. In 1982, the sewer lines were found to be radioactively contaminated.3 The sewage 

treatment plant began in the 1950’s to collect sanitary waste from Areas II, III, and IV. In addition 

to sanitary waste, the treatment plant received cooling tower discharges from non-chromated 

cooling tower systems and treated groundwater from various site groundwater recovery 

systems. The treated discharge was allowed to flow into the R-2A Discharge Pond in Area II. As 

part of the Site Wide Water Reclaim System, the Area II R-2A Pond serviced the Coca Rocket 

Engine Test Stands and Saturn V personnel. In July, 1982 Rockwell reported enriched uranium 

was identified in the sewer sump outside Building 4003 in the range of 23-32 μg/g. Rockwell 

concluded that ANL would object to the unrestricted use of Building 4003 but asserted that 

cleanup was impractical and almost impossible.4 

The HSA provides a detailed chronology of radiological investigations and D&D activities 

associated with the Building 4003 area. The preliminary MARSSIM classification for the Building 

4003 area is Class 1, due to its former use as an SRE support building and its proximity to SRE 

Building 4143. 

8.2.3 Building 4003 Radiological Incident Reports 

There have been several incidents associated with Building 4003 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

1 EPA/HGL, Final HSA 6, p.14, citation 2: 

2 Graves, A. W., Facilities Dismantling Plan for Building (D+D) T003 Hot Cave, Atomics International Report 

No. FDP-704-990-001, October 10, 1974, p. 2. 

2 EPA/HGL, Final HSA 6, p.14, citation 3: 

3 Atomics International letter from W. F. Heine to R. L. Westby, U.S. Atomic Energy Commission, re: 

Contamination Limits for Release of KEWB and Building 003 for Unrestricted Use, November 21, 1974. 

3 EPA/HGL, Final HSA 6, p.22, citation 1: 

1 Letter from B. D. Sujata, The Boeing Company, to J. Evans, County of Ventura, re: Information Regarding

Permit – Septic Tank and Leach Field, October 23, 2001. 

4 EPA/HGL, Final HSA 6, p.14, citation 4: 

4 Rockwell International Telephone Conversation Record from W. Smith to B. F. Ureda, Re: Building 003

Contamination, July 2, 1982. 

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8.2.4 Building 4003 Radiological Incident Report Summary - Data Provided by Boeing 

*Isotopes are written as they are presented in the incident database. The research team believes that MFP is an 

acronym for mixed fission products. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0530 10/7/1959 ETB Maybe Employee contaminated his hands when 

he failed to wear prescribed protective 

clothing.

A0423 12/22/1959 ETB MFP Contractor removed R/A exhaust stack 

without health physics cognizance.

A0549 7/22/1960 ETB Rm 160 Employee cut hand in controlled area.

A0295 2/1/1965 ETB & D004 Varied work location & late return of film 

badge resulted in exceeding guidelines.

Employee exposure: 2,000 mrem gamma 

and 640 mrem beta

A0438 9/15/1969 ETB Hot Cell MFP Multiple projects resulted in employee 

exposure above guidelines.

Employee exposure: 6,065 mrem

A0048 9/2/1969 ETB Chem Labs ETB Lab equipment and floor found 

contaminated.

• On October 7, 1959, an employee performed maintenance on a radioactive contaminated 

sodium line with his bare hands. The employee was aware the line was contaminated. Both of 

his hands became contaminated to a level of 1.5 mrad/h. After 35 minutes of cleaning, the 

employee’s hands were successfully decontaminated (A0530). 1

• On December 22, 1959 contract personnel were working on the exhaust stack of the 

radioactive vent systems at the ETB without the knowledge of the health physics department, 

without the proper permit, and without film badges. The contract crew was unaware the 

exhaust system was radioactive. No personnel contamination was discovered. Work was 

immediately stopped until the proper permit was filled out and film badges assigned (A0423).2

• On July 22, 1960 an employee cut his hand during a drilling operation. No activity was found 

above background on his hand (A0549).3

1 EPA/HGL, Final HSA 6, p.16, citation 2: 

2 Bell, C. E., Internal Letter, re: Violation of Health Phys. Practices, October 8, 1959. 

2 EPA/HGL, Final HSA 6, p.16, citation 3: 

3 Lang, J. C., Atomics International Internal Letter, re: Notice of Rule Infraction, December 31, 1959. 

3 EPA/HGL, Final HSA 6, p.17, citation 1: 

1 Illegible Author, Atomics International Internal Letter, re: Radiological Safety Incident Report, Room 160 

ETB Annex, July 22, 1960, July 27, 1960. 

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• In October 1965, a missing film badge belonging to a chemist was found. The badge had been 

missing since January 1965 and when processed, indicated an exposure 2,000 mrem gamma 

and 640 mrem beta. When added to the chemist’s February and March exposures, the 

quarterly total exposure was 2,600 mrem gamma and 1,870 mrem beta. Review of the 

chemist’s duties in January suggested that he did in fact receive the exposure indicated on the 

film badge, despite the possibility of other factors affecting the reading. The chemist’s lifetime 

accumulated exposure through September 1965 was 28.4 rem, below the permissible 

exposure of 65 rem at that time (A0295).1

• On September 15, 1969 film badges worn by an employee over the course of a calendar 

quarter were evaluated and the combined exposure from the badges was 6,065 mrem. This 

overexposure was found to be the result of multiple different operations at the ETB (A0438). 2

• On September 2, 1969 a smear survey was requested for the upstairs laboratory at the ETB 

because contamination had been found on the floor. The contamination was first noted on an 

employee’s shoe (200 cpm). The employee smeared the stairs and found up areas of 

contamination up to 1,000 dpm/100 cm2. This prompted him to smear the floor where he was 

setting up equipment. Repetitive smears of the floor brought forth “hotter” and “hotter” results. 

The employee eventually roped off the area and requested further surveying of the lab. 

Additional smear surveys found gross contamination on the bench top and covered sink. A

couple of spots read up to 1.5 rad/h. This material was easily removable. Parts of two ring 

stands on the work bench indicated 0.5 rad/h and were bagged up. Another “hot” area of the 

lab was located on an adjacent bench top where a survey detected 30,000 cpm. Smear 

surveys taken on September 4 and 5 1969 found contamination greater than 150,000 dpm/

100 cm2 (the counter detectibility limit) on the outside of a furnace, a balance, and a hot plate 

in the hood. The hot plate was measured to be 330 mrad/h. Also found in the hood, behind two 

lead bricks, was a bottle of liquid waste from earlier equipment, which read 5 rad/h. According 

to the incident report, the “episode was somewhat mysterious.” It appears that there may have 

been two sources of contamination because gamma activity was missing from one set of 

smears. It was unknown whether a chemical or physical process had removed gamma 

emitters, such as cesium, from this area. As of September 8, 1969, the lab had been fairly well 

decontaminated and most items outside the hood were less than 100 dpm / 100 cm2 betagamma (A0048).3

As part of the dismantling operations, Atomics International opened the west cell and conducted 

a radioactivity survey; the cell was found to contain significantly higher levels of radioactivity 

than the east cell. The west cell contained test analysis equipment and experimental residue. 

Miscellaneous waste was removed including three trays (7 rad/h) and two 1-gallon paint cans 

1 EPA/HGL, Final HSA 6, p.16, citation 2: 

2 Correspondence from Remley, M. E., Atomics International, to Levy, J., U.S. Atomic Energy Commission, 

Re: Apparent Type B Radiation Exposure, November 5, 1965. 

2 EPA/HGL, Final HSA 6, p.16, citation 3: 

3 Bresson, J. F., North American Aviation Internal Letter, re: Film Badge Exposure – Contamination at the 

ETC, SS003, Bresson to Heine, 9/13/69, October 15, 1969. 

3 EPA/HGL, Final HSA 6, p.16, citation 4: 

4 Bresson, J. F., North American Rockwell Corporation Internal Letter, re: Contamination at the ETC, S003, 

September 15, 1969. 

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(approximately 25 rad/h). These items were sent to the RMHF (Radioactive Materials Handling 

Facility) for disposal. Five prefilters (250 mrad/h) from the lower section of the cell and a 30-

gallon bag of solid waste (2 rad/h) were then removed and the cell was vacuum cleaned again 

to remove most loose contamination. The roof shield blocks and associated manipulators were 

removed while the openings in the roof were covered with plastic sheeting to contain loose 

contamination. The remaining items in the cell were then removed. These included a Lucite 

enclosure, a shelf, a table, and 90 gallons of solid radioactive waste. Dry uranyl salt solution (25 

rad/h at 10 cm) were removed from the floor. A container of SNAP burnup samples was located 

in the transfer tunnel between two cells. Using tongs, this container was transferred to a 5-

gallon can and then into a lead cask, which was then transferred to RMHF.1 

The HSA provides a detailed chronology of radiological investigations and decontamination / 

cleanup procedures, including Atomics International’s excavation of the Hot Cave. 

1 EPA/HGL, Final HSA 6, p.17, citation 3: 

3 Ureda, B. F., Building 003Decontamination and Disposition Final Report, AI-ERDA-13158, February 25, 

1976, pp. 13-15. 

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Section 8.3 

Building Number:

Building Alias: 

Visitor Log Entry Code: 

4055 

055 / T055

6013 / T055 

Building Name: Nuclear Materials Development Facility (NMDF)

Building Function: • Research, Development, Production of

 Nuclear Fuels and Radioactive Sources

• Advanced Fuel Systems Program (AFSP - LMFBR)

• Transuranic Waste Reduction Processes

• Molten Salt Combustion

• Support of the Fast Flux Test Facility (FFTF)

• Fission Research

• Fuel Fabrication 

• Fabrication of U-Pu Sources for Irradiation Testing 

Associated Buildings: Building 4155 - Guard Shack

Building 4755 - Substation 

Notes: Tritium 

Discharge to the Area IIl Silvernale Pond

(Site-Wide Water Reclaim System) 

Radionuclides of Concern: The primary special nuclear materials handled in the Building 4055 

were plutonium and uranium. Accordingly the radionuclides of concern are uranium, Pu-239, 

and their decay and daughter products, primarily Am-241, decay products from U-235, U-238, 

and Th-232 decay chains include Th-234, Th-228, Ac-228, Ra-226, Pb-214, Bi-214, Pb- 212, 

Bi212, and Tl-208. In addition, H-3, Sr-90, Cs-137, Na-24, Co-57, Co-60, Eu-152, and Am-243 

would have been formed. 

8.3.1 Description of Operations & Processes 

Building 4055 was constructed in 1967, specifically for developmental work involving plutonium. 

According to the 2006 Site Profile, the NMDF “incorporated all of the safety systems and 

safeguards required for such work,” and no incidents associated with the NMDF are referenced. 

However, the HSA research team provided abundant detail on facility operations, processes, 

and incidents, which are summarized below. It should be noted that many of the incident 

reports, which are contained in the Incident Report Database, document stack failure at the 

Building 4055 Plutonium Facility, which may call into question emissions data NIOSH is 

currently using for SSFL environmental / ambient exposure. 

The NDMF was used for a wide variety of purposes and programs. Based on historical 

documents, the major effort at Building 4055 involved plutonium-bearing fuels, primarily 

plutonium-239. These, according to a 1967 radiation and nuclear safety document, included 

Po-210, Pu-238, Pu-239, and Cm-242. For “feed material,” any of the programs used various 

forms of plutonium, uranium, and “other radionuclides.” 

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Following its construction, operations began in 1968 on two programs in support of the US 

Government’s Fast Flux Test Facility. In 1970, the facility initiated a fabrication effort for ANL. 

Both programs, according to Rockwell’s 1992 summary of plutonium capabilities, involved 

mixed uranium-plutonium oxide fuel pellets. In July 1968, a new program was initiated to 

develop a technique to “introduce micro-quantities of tungsten into mixed uranium-plutonium 

carbide” with the goal of the tungsten acting as a nucleating site for fission gas.1 

According to a July 22, 1968 internal letter regarding the in-box filter systems at Building 4055, 

the high-efficiency in-box filter systems for the glove box trains that were involved in the mixed 

oxide and carbide programs had some deficiencies with regard to nuclear material recovery, 

accountability, and safety. The memo indicates that the in-box high-efficiency filter in those 

boxes that may house particularly dusty operations needed to be modified; the amount of 

“inaccessible areas and crevices” near the recessed filter unit needed to be reduced to minimize 

the accumulation of particulates; and identified a need to improve secondary inbox containment 

capabilities for the operations that generated significant quantifies of airborne particulates. Prior 

to 1968, employees engaging in these activities may have been at elevated risk. 

A 1968 Atomics International internal letter regarding 12.3 gm of Pu introduced into Glove Box 

15 describes modification to the Radeco air monitors to “prevent false alarms in the high level 

alarm circuit” that would result in building evacuation. The memo specifies the need to 

recalibrate alarm circuitry to avoid the sounding of stack monitor alarms, which could result in 

“unnecessary building evacuation.” The memo described the modification of Radeco alarm 

plugs, so that pulling them would not cause an alarm to sound in the building (but a signal would 

still be received by the Control Center and only actuated at Building 4055 if the Control Center 

was notified by Health and Safety). New procedures for operations of counting equipment were 

introduced; including procedural changes so that determining alpha and beta air activity in the 

building would be determined by counting for alpha only. The memo specified that, based on the 

new schedule for daily health physics personnel and the “non-routine events that occur” at 

Building 4055, full-time coverage by the health physics department would be required.2 

During the mid 1970’s the NMDF conducted two programs concurrently. One involved the 

process development of synthesizing mixed plutonium-uranium carbide into high-density pellets 

to be used in the Advanced Fuel System Program for liquid metal fast breeder reactors. The 

second “demonstrated the use of molten salt oxidizer to reduce the volume of transuranic solid 

waste.” 

A 1981 onsite radiological contingency plan indicated that the procedures for handling analytical 

chemistry samples of plutonium-bearing fuel and other highly toxic material samples varied, 

according to the size of the sample and the nature of the analysis being performed 

1 EPA/HGL, Final HSA 5-D, p. 77, citation 2:

2 Litwin, R.Z., Rockwell International Letter Re: RI Background and Capability to Develop a Weapons-Grade 

Plutonium Fuel Cycle and Disposal Evaluation for the PDR, October 5, 1992. 

2 Atomics International, Internal Letter to R.E. Remley from R.E. Alexander, Re: Operational Safety Unit Weekly 

Newsletter for the Period Ending June 1, 1968,” June 6, 1988. File: HDMSp01640028.pdf 

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1 

Page 234 of 418 

A summary of contamination in the facility was presented in the final decommissioning report. It

was noted that the Chemistry Laboratory was suspected of low-level contamination, and only 

the most easily accessible locations in the Glove Box Room were suitably cleaned after one 

incident, while a portion remained contaminated following another incident. The radioactive filter 

system for the glove box exhaust system was contaminated, and the liquid radioactive waste 

holdup system was contaminated. In addition, it was noted that contamination existed inside the 

glove boxes, transfer tunnels between glove boxes, hoods in the controlled area, and above the 

glove box system. The decontamination plan noted that approximately 98% of all the plutonium 

contamination in Building 4055 was contained within the 38 glove boxes and ranged from 100 to 

10,000,000 dpm/cm2 (removable, per smear paper swipes). The high-volume air exhaust 

system ducts had suspected contamination levels of 0 to 100 dpm / 100cm2, mostly alpha, from 

plutonium and its decay products. 

The entire building was reportedly stripped to the walls, including contaminated equipment and 

surfaces, drain lines, and ventilation ducts and the final decommissioning report indicated the 

facility was decontaminated to levels that were “as low as reasonably achievable.” 

On May 23, 1991, Rockwell International submitted a letter to the Ventura County Air Pollution 

control District regarding the activation of Building 4055 in support of new lasers and advanced 

technology research programs, including diamond coating technology. The letter indicated that 

the programs were operated out of the Van Owen Building at Canoga Park, but were being 

relocated to Building 4055. 

8.3.2 Air Monitoring Data: Building 4055 

According to the 1973 annual report for environmental and radioactive effluent monitoring, the 

gaseous effluent released to unrestricted areas was monitored at Building 4055. Monitoring 

included the volume of effluent releases, the average and maximum concentration (μCi/mL), 

1 EPA/HGL, Final HSA 5-D, p. 81, citation 6:

6 

Rockwell International Report, AI-DOE-13559, Nuclear Materials Development Facility Decommissioning 

Final Report, March 31, 1987. 

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and the total activity released (Ci). The results of this yearly data are presented in the annual 

environmental monitoring and facility effluence reports from 1973 through 1986.1 

The preliminary MARSSIM Classification for Building 4055 is Class 1 based on previous

activities and incident reports. 

1 EPA/HGL, Final HSA 5-D, p. 77, citations 1-18, 1-3: 

1 Moore, J.D., Environmental and Radioactive Effluent Monitoring Annual Report 1973, Undated.

2 Rockwell International, Report No. ESG-79-7, Environmental Monitoring and Facility Effluent Annual 

Report 1978, April 1979.

3 Rockwell International, Report No. ESG-81-17, Environmental Monitoring and Facility Effluent Annual 

Report 1980, May 27, 1981.

4 Rockwell International, Report No. ESG-82-21, Environmental Monitoring and Facility Effluent Annual 

Report 1981, July 15, 1982.

5 Rockwell International, Report No. ESG-83-17, Environmental Monitoring and Facility Effluent Annual 

Report 1982, June 1983.

6 Rockwell International, Report No. ESG-84-9, Environmental Monitoring and Facility Effluent Annual 

Report 1983, March 1984.

7 Rockwell International, Report No. RI/RD85-123, Environmental Monitoring and Facility Effluent Annual 

Report De Soto and Santa Susana Field Laboratories Sites 1984, March 1985.

8 Rockwell International, Report No. ESG-84-9, Environmental Monitoring and Facility Effluent Annual 

Report 1983, March 1984.

9 Rockwell International, Report No. RI/RD85-123, Environmental Monitoring and Facility Effluent Annual 

Report De Soto and Santa Susana Field Laboratories Sites 1984, March 1985.

10 Rockwell International, Report No. RI/RD86-140, Environmental Monitoring and Facility Effluent Annual 

Report De Soto and Santa Susana Field Laboratories Sites 1985,April 1986.

11 Rockwell International, Report No. ESG-79-7, Environmental Monitoring and Facility Effluent Annual 

Report 1978, April 1979.

12 Rockwell International, Report No. ESG-81-17, Environmental Monitoring and Facility Effluent Annual 

Report 1980, May 27, 1981.

13 Rockwell International, Report No. ESG-82-21, Environmental Monitoring and Facility Effluent Annual 

Report 1981, July 15, 1982.

14 Rockwell International, Report No. ESG-83-17, Environmental Monitoring and Facility Effluent Annual 

Report 1982, June 1983.

15 Rockwell International, Report No. ESG-84-9, Environmental Monitoring and Facility Effluent Annual 

Report 1983, March 1984.

16 Rockwell International, Report No. RI/RD85-123, Environmental Monitoring and Facility Effluent Annual 

Report De Soto and Santa Susana Field Laboratories Sites 1984, March 1985.

17 Rockwell International, Report No. RI/RD86-140, Environmental Monitoring and Facility Effluent Annual 

Report De Soto and Santa Susana Field Laboratories Sites 1985,April 1986.

18 Rockwell International, Report No. RI/RD87-133, Environmental Monitoring and Facility Effluent Annual 

Report De Soto and Santa Susana Field Laboratories Sites 1986, March 1987.

1 Rockwell International, Report No. RI/RD88-144, Environmental Monitoring and Facility Effluent Annual 

Report De Soto and Santa Susana Field Laboratories Sites 1987, March 1988.

2 Rockwell International, Report No. RI/RD89-139, Environmental Monitoring and Facility Effluent Annual 

Report De Soto and Santa Susana Field Laboratories Sites 1988, May 1989.

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8.3.3 Building 4055 NMDF Radiological Incident Reports 

There have been several incidents associated with Building 4055 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

8.3.4 Building 4055 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0286 5/9/1970 Glove Box Room Pu Hole in Pu glove box glove contaminated 

employee.

A0222 6/26/1973 Glove Box Room Pu Glove box controller failure pressurized 

box releasing contamination to area

A0224 12/21/1977 Glove Box Room Pu Contaminated roll of green tape 

discovered in glove box room

A0063 1/10/1978 Glove Box 26 Pu Employee contaminated during foam 

cleaning of glove box.

A0068* 5/4/1978 Glove Box Room Pu Leaking glove contaminated employee 

lab coat and surgeon gloves.

A0335 5/10/1978 Glove Box Room Pu Lost seal during change-out caused loss 

of vacuum and contamination.

A0071 6/15/1978 Glove Box Room Pu Radioactive Waste Compacted in 

“Suspect” Waste Compactor

A0225 6/30/1975 Stack Monitor Pu Stack monitor vacuum line disconnected 

in plutonium facility.

A0072 7/16/1978 Facility Failure of air sampling pump and backup 

pump, shut-down operations

A0226 7/21/1978 Stack Monitor Pu Stack monitor in plutonium facility out of 

service for 84 hours.

A0073 7/24/1978 Support Lab Pu Floor contamination found in waste 

handling area.

A00582 6/26/1979 B-55 Glove Box Pu Am Maintenance of glove box furnace 

caused airborne activity.

A0081 5/10/1980 Stack Monitor Pu Failure of stack monitor system in 

plutonium facility.

A0250 10/31/1980 Support Lab Contaminated green coveralls returned 

with deconned “blueline” coveralls.

A0085 5/31/1981 Stack Monitor Pu Failure of stack monitor in plutonium 

facility.

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A0091 11/7/1981 Glove Box Room False airborne alarm in glove box room 

at plutonium facility.

A0100 3/24/1982 Compactor Area Pu Waste package with glass broke cutting 

hand of employee

A0113 2/7/1983 Glove Box Room Pu Plutonium oxide incised into the left 

index finger of an employee

A0114 3/27/1983 Stack Monitor Response to Stack Monitor Alarm

A0164 8/29/1986 Outside Overfilled portable radioactive liquid tank.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

8.3.5 Incidents of Known Environmental Releases 

• On January 4, 1971 a water line to the air-conditioning unit in the attic of Building 4055 leaked, 

causing water to run into office areas, change room, support area, and glove box room. 

Maintenance personnel attempted to pick up the water that had leaked with a vacuum cleaner. 

According to the internal letter documenting the incident, the radioactive waste system sump 

tank and the sample tank were both full, causing the water to back up in the support area 

drains. Personnel opened the valves to the 1,000-gallon holdup tank, which cleared the 

backed up water condition. According to the letter, a total of 800 gallons was collected. The 

water was sampled and measured to be 2.8 x 10 9 μC/cc. The letter indicated that the “water 

was released to the Rocketdyne [Area III Silvernale, Site-Wide Water Reclaim System] pond.” 1

• A December 8, 1976 Rockwell International internal letter documented the release of 

radioactive sludge during the replacement of the radioactive waste line to the Building 4055 

holdup tanks. According to the letter, removal of the old radioactive waste line was started by a 

contractor on December 6, 1976. The waste line was broken open where it exits Building 4055 

a four smears were taken inside the pipe. The smears measured 0 dpm and the contractor 

was allowed to break out the remainder of the waste line. Approximately 2 hours after the 

initial start of work, water and sludge were observed running out of the line. Two sludge 

samples were taken and measured 15 and 53 dpm. Direct readings from inside the removed 

pipe read 700 dpm.

Rockwell International surveyed the contractor personnel and instructed the contractor to 

cease work. Rockwell International staff removed the pipe and dirt where the sludge had run 

out of the pipes on December 7, 1976. Contaminated materials were put in “burial boxes” for 

disposal. A survey of the trench on December 7, 1976 indicated that no contamination was 

detected; however, the internal letter did not provide the results of the survey. The contractor 

1 EPA/HGL, Final HSA 5-D, p. 90, citation 1:

1 

Lane, W.D., North American Rockwell Internal Letter Re: Water Leak at the NMDF, January 14, 1971. 

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was informed that the trench was released, and was allowed to resume work.1

• On June 30, 1978 it was discovered that a stack monitor vacuum line had not been monitored 

for 23 days. Air samples taken directly from the room where the stack monitor was located 

were about twice the normal activity of the filtered air; however the incident report did not 

indicate what these levels were (A0225).2

• On July 21, 1978 it was discovered that a stack monitor in the plutonium facility was out of 

service for 84 hours due to an electrical failure. The incident report stated that airflow through 

a filter was maintained and no uncontrolled release of material occurred (A0226).3

• On March 27, 1983 Protective Services Control Center notified the facility health physicist of a 

stack monitor alarm. Upon investigation, the stack monitor chart showed an increasing count 

trend for about 1 hour to 26 dpm. The filter was removed and counted for alpha radioactivity. 

The result was 7 dpm alpha immediate and 3 dpm alpha after 19-hours of decay. After a 72-

hour decay, the final count on the stack sample was 0 dpm for a 100-minute counting time. 

The event was attributed to “the effect of atmospheric inversion conditions on naturally

occurring airborne radioactivity.” The alarm was reset and remained at normal levels. The 

incident report did not indicate whether the filter was reinstalled in the stack or if it was 

replaced with a new filter and disposed of (A0114). 4

• On August 29, 1986 an unknown amount of water overflowed from the RMHF transfer tank 

onto asphalt at Building 4055. A summary of the incident was reported to have been included 

on an attached flysheet; however, the flysheet was not attached to the file obtained by the 

research team. As a result, additional information regarding this incident, including the 

location, cannot be provided at this time. Additional information may become available upon 

receipt of additional documents from Boeing’s 1.4 million document database. 5

1 EPA/HGL, Final HSA 5-D, p. 91, citation 1:

1 Bradbury, S.M., Rockwell International Internal Letter Re: Replacement of R/A Waste Line to Hold-up 

Tanks at Bldg. 055, December 8, 1976. 

2 EPA/HGL, Final HSA 5-D, p. 91, citation 2:

2 Owens, D.E., Internal Letter Re: Incident Report, July 26, 1978. 

3 EPA/HGL, Final HSA 5-D, p. 91, citation 3:

3 Owens, D.E., Internal Letter Re: Stack Monitor Failure at NMDF, August 1, 1978. 

4 EPA/HGL, Final HSA 5-D, p. 91, citations 4 & 5:

4 Moore, J.D., Internal Letter Re: Quarterly Review of NMDF (T055) for Radiation Safety – First Calendar

Quarter, 1983, July 7, 1983.

5 Bradbury, S.M., Internal Letter Re: Radiological Safety Incident Report, March 29, 1983. 

5 EPA/HGL, Final HSA 5-D, p. 91, citation 1:

1 Begley, F.E., Internal Letter Re: Radiological Safety Incident Report, October 21, 1986. 

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Section 8.4 

Building Number: 4005 

Building Alias: 05 / T05 

Building Name: 1950’s - OMRE Support

1960’s - Uranium Carbide Fuel Manufacturing Pilot Plant (UCPFF)

Apollo Counterweight Coating Program (DOE-NASA)

1970’s - Molten Salt Test Facility (Coal Gasification)

 Decontamination of Sodium Components

1980’s - NOx-SOx Burner / Coal Processes 

Building Function: Thermodynamic Testing related to the OMRE / Piqua Reactors, and 

Uranium Carbide Fuel Manufacturing / Coal Gasification Support 

Note: “Non-Radiological” designation

Workers may not have been monitored

Radiological uses & incidents documented

1987 Contamination: 107,954 dpm/100 cm2 (beta) 2,467 dpm/100 cm2 (alpha)

DOE would not release for unrestricted use until 1994 

Radionuclides of Concern: As the UPCFF, Building 4005 was used to react uranium oxide 

with graphite to convert it to uranium carbide. The facility operated for a period of 9 months 

during 1966 and 1967, using depleted uranium and enriched uranium.1 Drain lines, duct work, 

walls and floor were found to be contaminated with natural and enriched uranium (U-238, 

U-234, and U-235) and fission products (C-14, Mn-54, S-35, Th-231, Th-234, P-32, Fe-59, and 

Co-60). Protactinium-234m (Pa-234m) is excluded from the list of radionuclides of concern as a 2

result of its short half-life of 1.17 minutes. However, for the purposes of dose reconstruction, its 

potential ingestion and/or inhalation by employees deserves consideration, given its properties 

and potential health hazards. 

8.4.1 Description of Operations & Processes 

Building 4005 was constructed in 1958 for non-nuclear testing of thermodynamic characteristics 

of proposed coolants for the Organic Moderated Reactor Experiment (OMRE) and Piqua 

1 EPA/HGL, Final HSA 5-A, p. 142, citation 600-602:

600 Rockwell International, Decontamination and Decommissioning of the Uranium Carbide Fuel Facility –

Building T005, 005-AN-002, September 28, 1993.

601 Rocketdyne, Final Radiological Survey of Building 005, 005-ZR-0001, September 21, 1993.

602 Rockwell International, Nuclear Operations at Rockwell’s Santa Susana Field Laboratory – A Factual 

Perspective, N001ER000017, December 20, 1989. p. 19 

2 EPA/HGL, Final HSA 5-A, p. 142, citation 603 & 604:

603 Rocketdyne, Final Radiological Survey of Building 005, 005-ZR-0001, September 21, 1993.

604 Unknown Author, Laboratory Status Report, Building 005, circa 1959, 3 pages. 

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reactors.1 During the mid-1960’s, Building 4005 was converted into a small-scale production 

facility, the Uranium Carbide Fuel Facility, to study the operations associated with manufacturing 

reactor fuel assemblies out of uranium carbide for the AEC Heavy-Water Organic Cooled 

Reactor. The work was performed under the authority of the AEC’s Chicago Operations Office, 

whose policy at the time exempted the operations in Building 4005 from licensure. 

A 1968 Operational Safety Unit Weekly Newsletter 2 references the Apollo Counterweight 

Coating Program. Given SSFL’s involvement in the Apollo program and SNAP, shield irradiation 

tests being conducted at the STR/STIR facilities nearby, and requirements of the Apollo 

shielding that were being developed, it is reasonable to deduce that this process involved NASA

and North American Aviation Space & Information Division Rocketdyne employees. According to 

the memo, extensive contamination levels in the furnace room of Building 4005 during the 

Apollo Counterweight Coating Program ranged “from 40,000 dpm/100 cm2 after breakout. 

[Employees] made reasonable efforts to decontaminate between melts. Contamination levels 

were usually reduced to less than 5,000 dpm/cm2. Spread of contamination from the work area 

was minimal, since levels in the hallway and change area were always less than 200 dpm/100 

cm2.” According to the document, finger ring badges were issued; extremity doses were incurred 

during the operation that were significant “but less than six rad for the entire six-week operation. 

Dose rates at contact with the freshly-cast counterweights ranged to 10 rad/hr, as did the 

crucible and mold.” The document explains that decommissioning efforts at Building 4005 were 

underway; equipment leaving the facility was tagged as uncontaminated (less than 200 dpm/ 

cm2 beta, less than 50 dpm/100 cm2 alpha). Equipment that was contaminated to higher levels 

was wrapped in plastic and sent to RMHF or Building 4063. In addition, the memo explains that 

Building 4005’s furnace room (108 and 110) with an average radiation level of about 10,000 

dpm/100 cm2 beta and 500 dpm/100cm2 alpha. 

In the pilot plant, uranium oxide was reacted with graphite to convert it to uranium carbide, 

which was cast into pellets, machined, and assembled into cladding tubes to make fuel 

assemblies. During this time approximately 700 UCx cylinders (.25” in diameter x 3” long) were 

fabricated. In 1967, equipment was removed and surfaces were reportedly decontaminated to 

permit non-radiological use of the building. However, the 2006 Site Profile specified that 

uranium contamination at this location had not yet been “cleaned up,” and several incidents 

documented after 1967 involved radioactive materials. 

Beginning in 1972, Building 4005 was used as the Molten Salt Test Facility, a non-nuclear test 

facility consisting of the Molten Salt Test Bed and the Process Demonstration Unit (PDU). 

According to the 1992 environmental monitoring program plan, the molten salt gasification plant 

(PDU) was designed to demonstrate the technical feasibility of producing sulfur-free, low-Btu 

product gas by partial combustion of Illinois No. 6 coal in a sparged bed of molten, sodium 

carbonate salt (“make-up salt”). The process resulted in “green liquor,” similar to that formed in a 

Kraft paper mill. This liquor (2 to 3 gallon per minute stream) was filtered and processed to 

regenerate sodium carbonate for recycle to the gasifier. According to the 1992, plan, these 

operations were first started in November 1978 and operated for a total of nine test runs until 

1 EPA/HGL, Final HSA 5-A, p. 131, citation 552:

552 Rockwell International, Decontamination and Decommissioning (D&D) of the Uranium Carbide Fuel

Facility – Building T005, 005-AN-002, September 28, 1993. 

2 Atomics International, Internal Letter to R.E. Remley from R.E. Alexander, Re: Operational Safety Unit Weekly 

Newsletter for the Period Ending June 1, 1968,” June 6, 1988. File: HDMSp01640028.pdf 

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final shutdown in June 1981. The total operating time was approximately 1,500 hours at an 

average coal feed rate of 0.25 tons per hour. 

In addition to the above, in 1982, Building 4005 also contained a low NOx-SOx burner to 

determine the possibility of burning high-sulfur coal with reduced emissions of NOx and sufur 

dioxide. 48 test runs were completed between 1982 to 1988. According to the 1992 

environmental monitoring program plan, the location had been partially remediated by removing 

bulk quantities of coal and fly ash. 

It should be noted that throughout the building’s function as a “non-radiological” facility devoted 

to coal gasification processes (from the early 1970’s onward), according to the 2006 Site 

Description, the ventilation ducts, filters and piping that were contaminated with uranium still had 

not been “cleaned up.” This suggests that the building, despite any designation as a “nonradiological facility” from the standpoint of personnel and potential exposure, was contaminated 

with radioactive substances for the entirety of its use as a Molten Salt Testing and NOx-SOx 

facility. 

The building included a number of concrete pads that held equipment used in the Molten Salt 

Oxidation Project and radioactive filter plenums from the fuel fabrication project.1 

Building 4005 was divided into several portions, including a small administration area, change 

rooms, chemistry laboratories, storage rooms and a large high-bay area. The building was 

connected to two underground radioactive liquid holding tanks by drain lines that extended from 

various laboratories and work areas within Building 4005, and were made of cast iron. 

The preliminary MARSSIM Classification of Building 4005 is Class 1 becasue of prior history as 

a uranium carbide fuel facility, known site contamination, reported incidents, and unknown 

building demolition operations. 

8.4.2 Building 4005 UCPFF Radiological Incident Reports 

There have been several incidents associated with Building 4005 that could have resulted in a 

release to the environment and worker exposure. It should be noted that while the building was 

considered a “non-radiological” facility, incidents involving radioactive materials are 

documented. The following table provides information presented in an incidents database 

provided by Boeing. Summaries of all available incident reports are provided in Attachment A. 

8.4.3 Building 4005 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0494 12/17/1959 Low Bay N/A Employee did not wear prescribed 

protective clothing in a controlled area

A0473 6/3/1960 High Bay S-35 HBR and kerosene caught fire.

1 EPA/HGL, Final HSA 5-A, p. 131, citation 554:

554 ORISE, Verification Survey of Buildings 005, 023, and 064, Santa Susana Field Laboratory, Rockwell 

International, Ventura County, California, 94/K-14, October 1994. 

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A0430 8/15/60 HBR Low Bay N/A Hot OMRE coolant spilled in hood, 

burning employee

A0603* 12/31/1966 UC Conv. Room Uranium 

Carbide

Failed coolant line in UC conversion 

room

A0606 1/30/1967 Building 5 Pad Uranium 

Carbide

Fire in roto cone duct work on service 

pad

A0605 2/13/1967 Room 115 Uranium 

Carbide

Fire reported in duct between “Queen 

City” and scrubber

A0649 7/18/1967 UCPFF Uranium A small fire occurred in the large particle 

collection tank.

A0215 8/8/1991 North Pad Depleted 

Uranium

Contaminated oil dripped from 

radioactive exhaust duct to pad.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

*Incident file name A0603 does not appear to be an incident report. Rather, the document is an internal letter dated 

Januaray 5, 1967, with the subject, “Radiation Safety Unit Weekly Highlights for the Week Ending December 31, 

1966.” A portion of this letter has been redacted and it is possible the referenced incident description has been 

redacted and, as a result, cannot be summarized. 

• On December 17, 1959 an employee was observed working the South Low Bay lab of Building 

4005 without a “red line” lab coat or film badge in two separate instances.1

• On June 3, 1960 while high boiler residue was being drained from the impurities loop system, 

a fire broke out within its enclosure. The incident report indicated the fire occurred in the north 

high bay of the building. Smears of the outside surfaces of the enclosure and nose swipes at 

the time of the incident determined no detectable contamination. On June 6, 1960 the Building 

4005 Health Physicist surveyed the entire area and found contamination of 80 dpm/100 cm2 

inside the enclosure on burned areas. According to the report, the highest radiation intensity, 

also on the burned areas, was approximately .9 mR/hr. The nature and quantity of the material 

could not be deciphered from the scanned document (gallons of 40% high boiler residue 

containing mainly S-35 and C-14?) (A0473).2

• On August 15, 1960, an employee was filtering treated “OMRE” coolant in a hood in the south 

low bay of Building 4005. The employee used a vacuum procedure in an attempt to improve 

the filtration. During this procedure, some of the liquid spilled onto the protective cotton gloves, 

seared through the gloves, and caused a second degree burn. The incident report stated that 

no contamination was detected on the worker’s skin (A0430). 3

1 EPA/HGL, Final HSA 5-A, p. 136, citation 570:

570 Atomics International, Notice of [Illegible] Infraction, A0494, December 18, 1959. 

2 EPA/HGL, Final HSA 5-A, p. 136, citation 571:

571 Warren, J.W., A0473, Building 4005 North High Bay, July 1, 1960. 

3 EPA/HGL, Final HSA 5-A, p. 136, citation 572:

572 Warren, J.W., A0430, Radiological Safety Incident Report, Bldg 005 Low Bay, August 15, 1960. 

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• On January 30, 1967 a uranium fire occurred in a retention tank of a vacuum system. Tank 

ducting was burned through, allowing a release of contaminated smoke to the building. 

According to the radiation safety unit weekly newsletter summarizing the incident, no largescale dispersal of contamination and no personnel exposure occurred as a result of the 

uranium fire. Samples of the residue removed from the tanks following the fire were submitted 

for analysis; however, the results of this analysis were not provided in the documentation 

relating to this incident (A0606).1

• A February 23, 1967 radiation safety unit weekly newsletter stated that on February 13, 1967 

a small fire occurred in a metal exhaust duct connecting a grinder with an air scrubber. A small 

amount of uranium carbide material had collected in an elbow of the duct and was ignited by a 

hot chip. No unfiltered release occurred, and damage was limited to burning of the paint on 

the surface of the duct. The newsletter states at the end of the same paragraph that “a total of 

700 gallons of radioactive liquid effluent with a concentration of 3.7x10 7 uc/cc were released 

from tank #4 to the surface drain.” (A0605).2

• According to a radiation safety unit weekly newsletter for the week ending on July 8, 1967 an 

inspection of Building 4005 resulted in the discovery that an undiscovered fire occurred in the 

large particle collection tank resulting in the blistering and peeling of the facility vacuum 

system. The inspection determined that no release to the atmosphere resulted (A0649). 3

• On August 8, 1991, personnel conducting a facility surveillance inspection observed an oil spill 

on the concrete pad under the radioactive exhaust duct. The area was surveyed with a GM 

beta-gamma probe that measured 8,000 dpm/100 cm2 beta-gamma. All removable 

contamination was bagged and removed from the area. The spill was painted and identified as 

fixed contamination. The exhaust duct was decontaminated at the suspect area and contained 

with tape and plastic. A “smear swipe” of the spill was collected and analyzed. Analysis 

determined that the material was depleted uranium with a total activity of approximately 2nCi 

(2x10 9 Ci). The total activity of the spill was approximately 4 nCi, below the radioactive 

material labeling limit for natural uranium of 100 μCi (A0215).4 

8.4.4 Radiological Investigations 

Following the fuel fabrication program in 1967, the building was decontaminated to permit nonradiological use of Building 4005. However, only “some” of the radioactive liquid drain lines were 

1 EPA/HGL, Final HSA 5-A, p. 136, citation 573:

573. Unknown, Internal Letter Regarding “Radiation Safety Unit Weekly Newsletter for Period Ending 

February 4, 1967,” February 9, 1967. 

2 EPA/HGL, Final HSA 5-A, p. 136, citation 574:

574 Unknown, Internal Letter Regarding “Radiation Safety Unit Weekly Newsletter for Period Ending 

February 13, 1967,” February 23, 1967. 

3 EPA/HGL, Final HSA 5-A, p. 136, citation 575:

575 Alexander, R.E., Internal Letter Regarding “Radiation Safety Unit Weekly Newsletter for Period Ending 

July 8, 1967,” July 13, 1967. 

4 EPA/HGL, Final HSA 5-A, p. 136, citation 576:

576 Wallace, J.H., Radiological Safety Report A0215, August 12, 1991. 

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removed. Additional decontamination was required in 1987; walls and floors of rooms and 

hallways; radioactive ventilation drops and drain lines, etc. The radioactive exhaust ducting, 

filter plenums, and underground radioactive drain lines were apparently not removed, since 

another characterization survey showed these items still in place and contamination was above 

1987 DOE release limits. The acceptable limits were 1000 dpm/100cm2, but the beta level 

detected was 107,954 dpm / 100 cm2 and the maximum alpha level detected was 2,467 dpm/ 

100 cm2. The EPA HSA provides a detailed chronology of radiological investigations associated 

with Building 4005.

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Section 8.5 

Building Number:

Building Alias: 

4064 

064 - T064 

Building Name: Fuel Storage Facility 

Building Function: Nuclear Material / Source Radioactive Material Storage

Radioactive Waste Storage

Contained materials for TRUMP-S Tests 

Notes: Tritium / Neptunium

Drainage through Area III to Area II Ponds 

Radionuclides of Concern: U-233, U-234, U-235, U-236, U-238, Th-228, Th-232, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Ra-226, Cs-134, Cs-137, Sr-90, H-3, Na-22, K-40, 

Mn-54, Fe-55, Co-60, Ni-59, Ni-63, Eu-152, Eu-154, Pm-147, Ta-182, Neptunium 

8.5.1 Description of Operations & Processes 

Building 4064 was constructed in two phases; the first in 1958 and the second in 1963. It was 

built to meet AEC criteria for vaults for the storage of fissionable material, equipped with 

intrusion alarms, and fenced. 1 The HSA provides a detailed description of the building’s 

features, equipment, schematics, and processes. 

Building 4064 was designed and built as a special nuclear material and source radioactive 

material storage building. It was used for storing packaged items of source material (natural 

uranium, depleted uranium, thorium) and special nuclear materials (enriched uranium, 

plutonium, and U-233) of various forms and configurations. Originally, the north (Room 114) and 

south (Room 110 and 104 vault) contained steel racks for storing material. The south side was 

primarily used for the storage of highly enriched uranium and plutonium bearing items; the north 

side was primarily used for source material and “low” enriched uranium storage.2 

Enriched uranium powders and source material powder packages were split into smaller units or 

combined into larger units in a glove box located in Room 104. The glove box was removed 

from Building 4064 at an unknown time. During shutdown and termination of the SNAP

program, excess Zr-U (enriched uranium) alloy material was sectioned into lengths suitable for 

1 EPA/HGL, Final HSA 6, p. 158, citation 5:

5 Boeing Environmental Affairs, Fact Sheet, Building 4064 – Fuel Storage Facility, February 10, 2000. 

2 EPA/HGL, Final HSA 6, p. 159, citation 4:

4 Remley, M. E. General Storage Data for Buildings 064 and 022, October 7, 1977. 

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packaging for shipment in DOE containers. This was done in Room 104; the floor was covered 

with plastic sheeting and the Zr-U was sectioned using a hack saw.1 

During the early 1960’s, the metal racks in the south half of Room 110 were removed in order to 

store material in “bird cages” and drums. This storage included large quantities of special 

nuclear material recoverable scrap. At this time, the fenced yard areas in the front, side, and 

back of Building 4064 were used to store 55-gallon drums of low-level enriched recoverable 

scrap. This material was shipped off-site in the mid-to-late 1960’s and early 1970’s, and may 

have relied on packaging containers constructed by employees at the Building 4163 Box Shop. 

Rockwell International claims that residual contamination from handling bare metallic pieces 

was from enriched uranium, natural uranium, depleted uranium, and thorium. Most reactor 

contracts had ended by the early 1980’s. After 1980, no special nuclear material powders were 

handled or repackaged at this location. After all fissionable material had been removed, 

miscellaneous equipment and containers of radioactive waste (principally soil) were stored in 

Building 4064. In 1993, all nuclear material was removed and Building 4064 was 

decontaminated. 

The HSA provides a detailed description of previous radiological investigations and 

decontamination/cleanup of releases. In brief, Rockwell International discovered high alpha 

radioactivity on the concrete ramps leading to rooms 110 and 114 in 1988. The alpha activity 

was concluded to arise from natural elements in the concrete that could not be removed. 

Ambient gamma exposure rate measurements showed a contaminated area bordering and 

outside the eastern fence, but according to Rockwell, no contamination was detected inside the 

fenced-in storage yard. Ambient gamma exposure rate measurements made outside the fence 

in the 2-acre area showed an area of about 4,000 square feet contaminated with Cs-137. Two 

soil samples collected from an area of greatest exposure rate showed a Cs-137 radioactivity 

concentration of 2,500 pCi/g, regarded as 2,500 times Rockwell’s background concentration. 

Beta activity was measured at 1,200 pCi/g in the area of greatest contamination.2 

In 1992, Rockwell surveyed a 6,580-square-foot area comprising the fenced in yard that 

surrounded Building 4064 to assess its radiological condition. All alpha surface activity levels 

were found to be below “Rockwell International’s acceptance limit” of 5,000 dpm/100 cm2. All 

beta surface activity levels were found to be below “Rockwell International’s acceptance limit” of 

5,000 dpm/100 cm2. All gamma exposure rates were found to be below “Rockwell’s acceptance 

limit” of 5 μR/h above Rockwell’s background rate of 15.3 5 μR/h. Rockwell concluded the 

fenced-in yard of Building 4064 met its criteria for release for unrestricted use.3 

1 EPA/HGL, Final HSA 6, p. 159, citation 3:

3 Montgomery Watson Harza, DOE Leach Fields (Area IV AOC) RCRA Facility Investigation Report, Santa 

Susana Field Laboratory, Ventura County, California, Draft, October 2003, pp. 2-3. 

2 EPA/HGL, Final HSA 6, p. 164, citation 1:

1 Chapman, J. A., Radiological Survey of the Source and Special Nuclear Material Storage Vault-Building 

T64, GEN-ZR-0005, August 19, 1988, pp. 2, 91-92. 

3 EPA/HGL, Final HSA 6, p. 164, citation 8:

8 Kneff, D. W., Tuttle, R. J., and Subbraman, G., Radiological Assessment of the Building T064 Fenced-in 

Yard, Rockwell International, N704SRR990035, January 12, 1994, pp. 10, 40-41. 

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In 1992 ORISE concluded Rockwell’s cleanup was incomplete upon identifying three Cs-137 

hotspots in the site yard. Rockwell subsequently remediated the hotspots and revised the 

Building 4064 side yard guidelines to meet a more restrictive 10 mrem/yr maximum dose rate 

for a residential scenario.1 

In 1993, Rockwell removed all known radioactively contaminated equipment, components, and 

structures from Building 4064, as well as asbestos containing materials. All radioactive waste 

was packaged and shipped to an unnamed disposal facility. This activity may have relied on the 

construction of shipping containers by employees in the Building 4163 Box Shop.2 

The preliminary MARSSIM classification for the building 4064 area is Class 1, due to its location 

within the ETEC and its former use as storage for radioactive source and special nuclear 

materials and radioactive waste. 

8.5.2 Building 4064 Radiological Incident Reports 

There have been several incidents associated with Building 4064 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

8.5.3 Building 4064 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0523 3/10/1959 SS Vault SSFL Personnel entered vault area without 

required film badges.

A0028 2/18/1963 Yard SS Vault MFP Cask drain plug rusted out, draining 

contaminated water onto the ground.

N/A 5/3/1963 SS Vault U Cans of UC exploded inside SSFL vault.

A0468 10/8/1964 SS Vault U Can of UC exploded inside a “birdcage” 

shipped to SSFL Vault.

A0622 7/20/1967 SS Vault SSFL U308 Outside transfer of U powder caused 

increased activity in vegetation samples.

A0095 1/12/1982 Vault RAS Alarm Response.

A0109 9/21/1982 Vault Response to RAS alarm at SS Vault.

1 EPA/HGL, Final HSA 6, p. 165, citation 1:

1 Vitkus, T. J., Verification Survey of the Old Conservation Yard, Building 064 Side Yard, and Building 028, 

Santa Susana Field Laboratory, Rockwell International, Ventura County, California, ORISE 93/J-107, 

October 1993, pp. 5-11. 

2 EPA/HGL, Final HSA 6, p. 165, citation 3:

3 Dahl, F. C. and Tuttle, R. J., Final Radiological Survey Report of Building 064 Interior, ETEC Report No.

SSWA- ZR-0001, January 14, 1994, pp. 17-19. 

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A0218 6/8/1992 Storage Racks U Employee stacking fuel storage racks cut 

hand.

A0663 8/8/1995 Side Yard Cs-137 Performing the modified Area IV transient 

survey in grid R24S25, a hot spot was 

found.

A0687 9/28/1998 Side Yard Cs-137 Excavated 6 cubic yard of Cs-137 

contaminated soil.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

*Isotopes are written as they are presented in the incident database. The HSA research team believes that “MFP” is 

an acronym for Mixed Fission Products. 

• On March 10, 1959 an employee entered the storage vault without a film badge and it was not 

known how long he was in the vault. The employee was reminded on other occasions to wear 

his film badge (A0523).1

• On February 18. 1963 soil east of the exclusion fence at the SS Vault was found to be 

contaminated during a routine survey. Three additional surveys revealed an area of 700 

square feet of asphalt and soil contaminated with mixed fission products to a maximum of 700 

mrad/h at 2 inches. It was deduced that the contamination resulted from a cask containing 

irradiated Seawolf submarine reactor fuel pins. The cask was received at the vault, stored, 

transferred to the Hot Lab for inspection, and returned to the vault for storage. After storing the 

fuel pins for 1.5-2 years it was shipped back to Westinghouse around May, 1962. Sometime 

during this last storage period, it is believed that the drain plug rusted through, permitting fluid 

contaminated within the cask to spill onto the asphalt. A soil sample showed 8,340 pCi/g betagamma gross radioactivity, including 1x106 dpm/g of Cs-137 and 2x105 dpm/g of Cs-134. 

Alpha activity was negligible in all samples. Approximately 2,365 gallons of soil and asphalt 

were removed at the vault to reduce the maximum contamination level to 0.5 mrad/h. Three 

inches of soil was removed from all the contaminated soil areas and 80 percent of the areas 

were found to be free of contamination upon resurveying. The remaining areas were 

excavated up to 1.5 feet. Fill soil was brought in by the maintenance department to return the 

soil levels to their previous level (A0028).2

• On May 3, 1963 an explosion occurred in the north vault of Building 4064. The sources of the 

explosion were a 1-gallon can containing 8.3 kilograms of uranium carbide (3.7 percent 

enriched U-235) in the form of fines and pieces, and a 1-gallon can containing 2.5 kilograms of 

uranium carbide (4.9% enriched U-235) comprising sintered pieces. The fuel in both cans was 

immersed in Redline 60 oil. The explosion was believed to have been caused by the release 

of hydrogen gas from degrading uranium carbide. After 20 minutes elapsed, the Atomics 

International Fire Department entered the vault and smothered the cans with Metal X. 

1 EPA/HGL, Final HSA 6, p. 161, citation 1:

1 Loba, M. L., Internal Letter, Re: Violation of Health Phys. Practices, March 13, 1959. 

2 EPA/HGL, Final HSA 6, p. 161, citations 2 & 3:

2 

Badger, F. H., North American Aviation Internal Letter, re: Contamination Incident SS Vault Santa Susana, 

November 11, 1965.

3 Moore, J. D., Atomics International Internal Letter, re: Radioanalysis Report, February 22, 1963. 

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According to Atomics International, a smear survey confirmed that contamination had been 

confined to the interior of the vault, however the Incident Report indicates that contamination 

was noted some contamination on the grill of the ventilation exhaust. The cans were removed 

and the vault was decontaminated using kerosene and K-pads. Atomics International planned 

to remove approximately 10 square feet of asphalt floor tiles where the cans had been 

located.1

• On October 8, 1964 a custodian found black oxidized powder deposited on the inside of a 

“bird cage” container. Investigation revealed that uranium carbide had oxidized blowing the lid 

off one 1-gallon can and warping the bottom of the container. The inside of the can was 

surveyed at 5 x 103 dpm/100 cm2 alpha. Contamination levels on the concrete dock of the 

building 4064 increased to 200 dpm / 100 cm2 alpha from a “clean” level. Evidence suggested 

that an inert atmosphere had not been sufficiently established in the 1-gallon container 

(A0468).2

• On July 20, 1967 an investigation was performed to identify reasons for increased alpha 

activity in vegetation samples between Building 4064 and the fire station. It was determined 

that uranium carbide and uranium oxide powder had been transferred from one 55-gallon 

drum to another in the Building 4064 storage yard. The drum from which material had been 

removed was still located in the yard on a piece of plastic sheeting and visible amounts of 

uranium oxide powder remained. The plastic was folded carefully and disposed of as 

radioactive waste. No “significant” surface contamination was found in areas other than the 

plastic sheeting and vegetation samples collected between Building 4064 and the fire station. 

Personnel were cautioned about procedures for transferring radioactive materials or opening 

drums of radioactive materials (A0622).3

• On January 12, 1982 an alarm went off in Building 4064. Crews arrived at the building to 

check it out and the all-clear was given within an hour of arriving on the scene. The cause for 

the alarm was unknown and the alarm was to be checked by maintenance (A0095).4

• On September 21, 1982 an alarm went off in Building 4064. Crews arrived at the building to 

check it out and the all-clear was given within an hour of arriving on the scene. The cause for 

1 EPA/HGL, Final HSA 6, p. 161, citation 4:

4 Coonce, G. L., Atomics International Internal Letter, re: UC Fire at Building 064, May 8, 1963. 

2 EPA/HGL, Final HSA 6, p. 162, citation 1:

1 Owen, D. E., Atomics International Internal Letter, re: Incident Report, Building 064 Vault, 10-8-64, October 

27, 1964. 

3 EPA/HGL, Final HSA 6, p. 162, citation 2:

2 Alexander, R. E., Atomics International, re: Radiation Safety Unit Weekly Newsletter for Period Ending July 

22, 1967, August 2, 1967. 

4 EPA/HGL, Final HSA 6, p. 162, citation 3:

3 Bradbury, S. M., Rockwell International Internal Letter, re: Radiological Safety Incident Report, Building 

064, 1/12/82, Unknown Date. 

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the alarm was unknown and the alarm was to be checked by maintenance (A0109).1

• On June 8, 1992 an employee was stacking storage racks on a pallet when he cut the palm of 

his hand. No detectable activity was noted on the employee’s hand or the storage racks, 

despite concern for fixed U-235 contamination under paint on the fuel storage racks (A0218).2

• On August 8, 1995 a hot spot of approximately 1 square foot was identified during an Area IV 

survey in grid number R24S25. The hot spot was measured at 10,000 cpm or 46 μR/h. Per 

Rockwell’s procedure, a 1 meter ambient survey was performed, but it was below the 5 μR/h

above background limit for classifying as a hot spot. However, the Building 4064 sideyard was 

a formally remediated facility, so the discovery was reported to the Area IV survey project 

manager. The hot spot was roped off and a sample was taken. The sample contained 271 pCi/

g of Cs-137. The Cs-137 concentration was too low to be considered an inhalation hazard and 

the surface radiation was only three times background, so there were no specific safety 

concerns. According to Rockwell, the discovery of contamination did not warrant a formal 

occurrence report because building 4064 was not yet released for unrestricted use. The hot 

spot had been identified in a verification survey by the Oak Ridge Institute for Science and 

Education (ORISE), but had not been properly communicated to Rockwell. The discovery was 

cited as an example of the importance of a walk-about surface gamma surveying used in the 

Area IV characterization survey (A0663).3 The exact location of this hot spot has not been 

identified but it appears to be west of Building 4064.

• On September 28, 1998 ORISE discovered an area of elevated soil contamination in the side 

yard. The area measured approximately 18 feet by 6 feet and was located on the sloping bank 

between 10th street and the Building 4064 parking lot. A total of 6 cubic yard of soil were 

excavated until radiation levels reached background levels. ORISE resurveyed the area, 

verified that radiation levels were normal, and collected two soil samples. Rockwell collected 

additional samples to confirm that the remediation had been successful (A0687).4 

1 EPA/HGL, Final HSA 6, p. 162, citation 4:

4 Bradbury, S. M., Rockwell International Internal Letter, re: Radiological Safety Incident Report, Building 

064, 9/21/82, September 22, 1982. 

2 EPA/HGL, Final HSA 6, p. 162, citation 5:

5 Wallace, J. H., Rockwell International Internal Letter, re: Radiological Safety Report, T064 South Vault, 

6/8/92, June 17, 1992. 

3 EPA/HGL, Final HSA 6, p. 163, citation 1:

1 McGinnis, E. R., Rockwell International Internal Letter, re: Radiological Incident Report, Soil Contamination 

Found in T064 Sideyard, 8/8/95, August 25, 1995. 

4 EPA/HGL, Final HSA 6, p. 163, citation 2:

2 Rutherford, R., Rockwell International Incident Report, re: Verification Survey, September 28, 1998. 

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9.0 Disassembly & Examination of Reactors / Used Reactor Fuel Assemblies 

There are two Hot Laboratories documented at SSFL. The Building 4009 SGR Hot Laboratory 

was not included in the 2006 Site Description. It is described in detail in Section 6.1, “Critical 

Test Facilities.” 

As indicated in the 2006 Site Description, reactor test operations often required examination of 

reactor fuel assemblies and other test specimens to determine how they were performing. 

Reactors were often removed from their operating locations, and moved to the Building 4020 

Hot Lab for examination. Some of the reactors required complete disassembly for the transition. 

Other, smaller reactors, (like the S8DR after its fuel failure) were moved without complete 

disassembly. For instance, the S8DR was transitioned to the STIR Facility for “dry runs” before 

being taken to the Building 4020 Hot Lab for disassembly and analysis. 

The Building 4020 Hot Laboratory was used to examine fuel and/or components from the SRE, 

SER, S2DR, S8ER, S8DR, and the S10FS3 reactors at SSFL, the OMR and SGR criticality test 

facilities, and the Piqua, Ohio, reactor. It was also used to declad fuel from the SRE, EBR-1, 

EBR-II, Hallam, Fermi, and Southwest Experimental Fast Oxide reactors. In addition, it may 

have been used in the examination of the NASA nuclear reactor, the HMRSFR that operated in 

Building 4012. In addition, the Hot Labs were used to manufacture sealed sources. 

The 2006 Site Description indicates that as a result of the work in the Building 4020 Hot Lab, the 

interior of the Hot Cells and equipment they contain “have been contaminated by small amounts 

of uranium, plutonium, thorium, and fission and activation products. Some exterior surfaces 

have become slightly contaminated, as well. The extent of the contamination is approximately 

2.2 Ci in the drain system, ventilation and exhaust system, and inside the shielded cells plus 

traces of contamination on the building walls and surroundings.” 

The 2006 Site Description only references one incident (a fire, May 19, 1971 - Incident Report 

#A0052). The 2006 Site Description states that, “No measurable external or internal exposures 

of radiation were received by personnel involved in the activities before, or during the fire.” 

The following description of the Building 4020 Hot Laboratory, conducted by EPA as part of the 

HSA component of the Area IV Radiological Study, documents 129 incidents associated with 

the facility that could have resulted in environmental releases and worker exposure. 

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Section 9.1 

Building Number:

Building Alias: 

Visitor Log Entry Code: 

4020 

020 - T020 

6012 / 6000C / 6001F / T020 / AIHL / RIHL 

Building Name: Hot Laboratory

Component Development Hot Cell 

Building Function: Research and development, examination, decladding, decontamination, 

analysis, etc. of fuel and highly radioactive materials. Please see detailed list of Hot Cell uses 

and description of building processes below.

Notes: According to Incident Report A0027, airflow in the Hot Lab was below design standards 

in some areas. Tritium use. Numerous incidents resulting in worker exposure and airborne 

releases. 

Radionuclides / Constituents of Concern: Building 4020 handled a number of radionuclides 

including uranium, plutonium, thorium, and promethium. Decay products from U-235, U-238, 

and Th-232 decay chains include Th-234, Th-228, Ac-228, Ra-226, Pb-214, Bi-214, Pb-212, 

Bi-212, and Tl- 208. In addition, H-3, Sr-90, Cs-137, Na-24, Co-57, Co-60, Eu-152, and Am-243. 

Solid reactor fuel materials (including but not limited to) U-Mo, UO2, U-ZrH, plutonium-bearing 

fuels, (in addition to cladding materials: aluminum, sintered aluminum powder, stainless steel, 

zirconium, Hastelloy alloys, Sodium, and NaK bonded capsules). 

9.1.1 Description of Operations & Processes 

The Building 4020 Hot Laboratory is among the most detailed in the EPA HSA. Its diverse 

operations, numerous incidents resulting in releases and worker exposure, and myriad of 

functions were thoroughly investigated by EPA’s research team and explained in detail in the 

HSA. The below description of building operations and processes is only an abbreviated 

synopsis of the information contained in the EPA HSA. In addition, EPA felt it was appropriate to 

include numerous excerpts from employee interviews conducted by EPA and DOE during the 

Area IV Study and HSA component. The interviews detail job processes and incidents at 

Building 4020 over several eras of facility operations. They are contained, in their entirety, on 

the accompanying disk with the EPA HSA report. 

Building 4020 was constructed in 1959 as the Component Development Hot Cell used for the 

remote handling and examination of highly radioactive materials. During initial construction, the 

facility’s radioactive drain was connected to two 500-gallon holdup tanks that were located in the 

north end of the basement. One tank was designed for high-level waste, and the other for lowlevel waste. Following an incident whereby these tanks overflowed (no incident report 

available), the system was modified by installing a 3,000 gallon tank in Building 4468 in 

approximately 1970, and removing the 500-gallon tanks from Building 4020 in 1977. After the 

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new system was installed, the contents of these sumps were pumped into the Building 4468 

holdup tank. Building 4020 was connected to the Area III Sewer Treatment System. 1

According to the 1984 license renewal application, stack sampling was performed to permit the 

measurement of particulate radioactive material discharged from the facility. This monitoring 

was performed quarterly and the results were presented annually in environmental monitoring 

reports. However, documents indicate that emissions were sampled for entrained particulate 

radioactive materials by means of continuous stack samplers that were installed at the point of 

release. In addition, a gas monitor was also installed in the stack to measure radioactive gas 

discharges and to indicate accidental criticality of low-energy releases occurring within a cell. 

According to DOE’s Environmental Audit (1989) radiological monitoring was inadequate. CORE 2

Advocacy has included a disk containing historical documents, with this Proposed 2016 SSFL

Site Description. A folder on the disk contains approximately 30 historical Environmental 

Monitoring Reports authored by Atomics International, between the 1950’s to mid-1980’s. 

Building 4020 was used for the examination and evaluation of reactor fuels and components, 

and in irradiation test materials. This included Sodium Reactor Experiment (SRE) fuel 

assemblies and moderator cans, fuel elements from the Ohio Piqua Nuclear Power Facility and 

the Organic Moderated Reactor Experiment (OMRE) reactors, and fuel test capsules irradiated 

at the Materials Testing, Engineering Test, General Electric Testing, (HWOCS) and Hanford 

reactors. 

Irradiated fuel test capsules, based on available documents, appear to have been used in 

experiments to verify predicted release rates and disposition of fission products from vented fuel 

elements. They required disassembly at Building 4020, similar to fuel rods, to remove 

radioactive material. An april 13, 1971 internal letter documents the preparation of Building 4020 

for the acceptance and disassembly of an “URIPS-1P-RTG” from Aeroget-General Nuclear 

Company. The fuel for the RTG was documented to be a capsule containing 7,700 Ci of Sr-90.3 

Building 4020 was also used for the disassembly and examination of Systems for Nuclear 

Auxiliary Power (SNAP, a DOE-NASA project) Reactor cores, the analysis of irradiated test 

materials, the manufacture and leak-testing of sealed radioactive sources (primarily Pm-147), 

and the machining of radioactive Co-60. At an unknown date, the entire SNAP 8 Experimental 

Reactor core, containing 211 fuel elements, was disassembled and examined at Building 4020. 

1 EPA/HGL, Final HSA 5D, p. 14, citations 4,5,6:

4 Boeing, EID-06141, “Hot Laboratory Decontamination and Dismantlement Final Report,” November 27,

2001.

5 Boeing, Document No. RS-00010, Area 4020 MARSSIM Final Status Survey Report, October 31, 2000.

6 

Badger, F.H., Rockwell International Internal Letter, Re: Radiological Occurrence – Bldg T020, October 3, 

1977. 

2 DOE Office of Environmental Audit, “DOE Activities at SSFL,” 1989. Page 4-37, Paragraph 4, Re: Radiological 

Monitoring. 

3 EPA/HGL, Final HSA 5D, p. 22, citations 1&2:

1 Atomics International, Document No. NAA-SR-9469, Project Proposals Fiscal Years 1965 and 1966, 

Sodium Cooled Reactor Development Programs ̧November 2, 1964.

2 Heine, W.F., North American Rockwell Internal Letter Re: Operational Safety Unit Weekly Highlights – 

Week Ending 4/9/71, April 13, 1971. 

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The SNAP 10FS3 reactor, including the vessel, core, primary pump, and piping, was also 

disassembled and examined at the Hot Lab. 

The Hot Lab involved the use of Dowanol, which is a trade name for ethylene glycol n-butyl 

ether (99%), which was found to be effective for the removal of sodium and NaK from 

components by reaction. Its use at SSFL for the removal of sodium is documented in 

documented as early as 1965,1 and it is reasonable to assume Dowanol was routinely used in 

all SSFL sodium component cleaning and testing facilities. 

Specific operations performed in Building 4020’s Hot Cells varied, depending on the program 

needs and changing program requirements. However, the work was principally categorized as 

“research and development.” The various capabilities of each cell block area was listed in the 

1984 license renewal application as follows: 2

• Cell #1:

- Preparation of samples of irradiated material for metallography

- Micro-hardness testing

- Microscopic measurements

- Preparation and replication of samples for electromicron microscopy

- Autoradiography on mounted samples

• Cell #2
 

- Materials testing, tensile testing, stress-rupture and creep testing, fatigue testing
 

- NaK and sodium distillation
 

- Visual examination
 

- Density measurements
 

- Dimensional measurements
 

- Minor component disassembly
 

- Fission gas collection
 

- Isotope encapsulation
 

• Cell #3
 

- Disassembly cell for irradiated materials
 

- Sample preparation
 

- Elox equipment
 

- Cutoff wheel

- Waste packaging

- Visual examination
 

- Stereomicroscopic examination
 

- Dimensional measurements
 

- Cask Unloading
 

1 EPA/HGL, Final HSA 5D, p. 23, citation 3:

3 Atomics International, NAA-SR-MEMO-11605, Weekly Hightlights for S8ER Operations Week Ending 

August 21, 1965, August 27, 1965. 

2 EPA/HGL, Final HSA 5D, p. 23, citation 6:

6 Rockwell International, ESG-82-33, Health and Safety Sections for Renewal Application of the Special 

Nuclear Materials License SNM-21, Docket 70-25, Issued to Energy Systems Group of Rockwell

International, June 5, 1984. 

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• • Cell #4

- Hydrogen analysis (Tritium)
 

- Profilometer measurements
 

- Annealing studies
 

- Permeation testing
 

- Major component disassembly and repair
 

- Visual examination
 

- Stereomicroscopic examination
 

- Fuel canning
 

- Dimensional measurements
 

- Waste packaging
 

- Cask unloading and loading
 

- Density measurements
 

- Gamma spectroscopy
 

- Autoradiography on capsule assemblies 

The Hot Cells were backed by decontamination rooms for examining high-level radioactive 

material. The areas surrounding these included office space, an operating gallery, operations 

support, a mock-up area, and a service gallery. The operations support facilities included a hot 

storage room (Room 153), air lock (Room 155), hot laboratory (Room 141), manipulator 

maintenance room (Room 128), and glove box laboratory (Room 139). The EPA HSA provides 1

a detailed description of Building 4020’s features, dimensions, and equipment routinely used in 

each part of the Atomic International / Rockwell International Hot Laboratory (AIHL / RIHL). The 

facility also included hot and cold change rooms, a photographic laboratory, and office areas. 

The preliminary MARSSIM Classification for the Building 4020 is Class 1 because of previous 

site use, incident reports, and radioactive material use during building operations. 

9.1.2 Building 4020 Radiological Incident Reports 

There have been several incidents associated with Building 4020 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

9.1.3 Building 4020 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0002 8/26/1959 CDHC Hot cell door ramp fell, striking an 

employee’s hand

A0001 12/5/1959 CDHC Mixed Fission 

Product

Radioactive liquid spilled during vendor 

pickup.

1 EPA/HGL, Final HSA 5D, p. 15, citation 3:

3 Boeing, EID-06141, “Hot Laboratory Decontamination and Dismantlement Final Report,” November 27,

2001. 

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A0392 4/14/1960 CDHC DECON 4 Mixed Fission 

Product

Weekly radiation limits exceeded

A0003 4/18/1960 CDHC Mixed Fission 

Product

Irradiated SRE fuel slug placed in an 

employee’s hand

A0011 5/12/1961 CDHC Mixed Fission 

Product

Unsafe Handling of B of E Cask

A0505 5/20/1961 CDHC Mixed Fission 

Product

Personnel contaminated during waste 

transfer operations.

A0347 

(same 

incident as 

A0505 

above)

5/21/1961 Service Gallery Mixed Fission 

Product

Employee contaminated while loading 

waste box

A0283 6/2/1961 Cell 3 Mixed Fission 

Product

Employee exposed to airborne activity 

when supplied air hose was pinched.

A0012 6/19/1961 CDHC C2 & D2 Mixed Fission 

Product

Employee attempted to enter very high 

radiation area.

A0317 5/9/1962 Cell 4 Mixed Fission 

Product

Irradiated fuel slug burned in Cell 4 

releasing radioactive gas.

A0016 5/31/1962 North Pad / Drain Mixed Fission 

Product

Portable radioactive liquid tank 

overflowed on North Pad flowing to 

surface drainage.

A0018 9/4/1962 Cell 4 Mixed Fission 

Product

Repair of Fission Gas Analyzer caused 

high airborne activity and contaminated 

personnel and lab

A0020 12/5/1962 North Pad / Drain Mixed Fission 

Product

Contaminated shoes from controlled 

area contaminated most of Hot Lab.

A0021 1/25/1963 Cell 3 Mixed Fission 

Product

Employees hand contaminated when 

glove tore in Hot Cell.

A0433 5/8/1963 Cell 3 Mixed Fission 

Product

Waxing mop swung through controlled 

area contaminated the clean area.

A0022 7/2/1963 Radioactive 

Storage Yard

Spill of OMRE HB 40 coolant in 

radioactive storage yard.

A0316 7/8/1963 RIHL & Off Site Mixed Fission 

Product

Employee contaminated clothing during 

cask movement out to storage yard.

A0027 925/1963 Cell 3 Mixed Fission 

Product

A fire occurred during dissolution of NaK 

from fuel decladding in Cell 3.

A0024 9/26/1963 DECON / Cell 3 Mixed Fission 

Product

NaK and alcohol fire during cleaning of 

fission gas monitor.

A0025 9/26/1963 Cell 3 Mixed Fission 

Product

Exposure above guidelines during cell 

cleanup.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

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A0026 10/9/1963 Cell 2 Mixed Fission 

Product

Uncontrolled furnace left on overnight 

burning cell equipment.

A0031 3/20/1964 OP Gallery Mixed Fission 

Product

Fire suppressing nitrogen purge of Hot 

Cells generated airborne activity.

A0551 6/8/1964 CDHC Slave Shop Mixed Fission 

Product

Unauthorized and unprotected employee 

contaminated by contaminated 

equipment.

A0443 6/16/1964 CDHC OP Gallery Mixed Fission 

Product

Fuel element elevated in cell causing 

radiation streaming through slave parts.

A0338 7/10/1964 Service Gallery Outer and inner doors to airlock open at 

the same time.

A0554 7/23/1964 CDHC Cell 4 Mixed Fission 

Product

Contaminated employee was deconned 

by medical department.

A0354 8/27/1964 OP Gallery Mixed Fission 

Product

High airborne activity resulted after loss 

of exhaust controller

A0415 9/3/1964 CDHC Cell 1 Mixed Fission 

Product

In-cell work resulted in exposure above 

guidelines and employee contamination.

A0337 10/15/1964 Service Gallery Mixed Fission 

Product

Employee became contaminated while 

bagging shielding blanket.

A0574 11/18/1964 CDHC Cell 3 Mixed Fission 

Product

Exit survey from cells revealed speck 

contamination.

A0033 11/26/1964 North Pad Mixed Fission 

Product

Radioactive liquid transfer tank was overfilled and contaminated outside area.

A0034 12/7/1964 Storage Yard Radioactive material stored in yard 

creating radiation levels above 

guidelines.

A0035 5/27/1965 OP Gallery Mixed Fission 

Product

High airborne activity during decladding 

NaK bonded uranium carbide fuel.

A0037 7/16/1965 OP Gallery Mixed Fission 

Product

Disintegrated fuel deposited in filter was 

spread during recovery.

A0441 8/12/1965 AIHL Yard Mixed Fission 

Product

Rain water from one way waste cask 

dumped in clean area contaminating 

ground.

A0038 9/21/1965 OP Gallery Mixed Fission 

Product

Cell purging with nitrogen caused high 

airborne activity.

A0039 2/2/1966 Metalograph Room Mixed Fission 

Product

Extremity exposure above guidelines 

from picking up irradiated fuel sample.

A0040 2/24/1966 Service Gallery Mixed Fission 

Product

Maintenance of contaminated Elox 

machine resulted in high airborne activity.

A0042 8/16/1966 ALL OF AIHL Failure of emergency power generator 

caused increased airborne activity.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

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A0044 12/31/1966 Room 139 Pm147 Fabrication of heat sources caused 

extremity exposures above guidelines.

A0607 1/28/1967 Service Gallery Mixed Fission 

Product

Contamination tracked from slave shop 

to service gallery and hot shop.

A0613 4/6/1967 AIHL Cell 3 Mixed Fission 

Product

Fire occurred in Cell 3 during cutting of 

metallurgical sample.

A0617 5/17/1967 AIHL Glove Box PM-147 Promethium glove box transfer caused 

high airborne activity and nasal smears.

A0619 6/10/1967 AIHL Glove Box Pm-147 Wast removal from the promethium glove 

boxes caused high airborne activity.

A0046 8/21/1967 RM 139 Glove Box Pm-147 Extremity exposure above guidelines 

during fabricated of heat sources.

A0047 10/11/1967 RM 139 PM-147 Ingestion of Pm-147 during cleanup 

operation.

A0627 10/30/1967 AIHL Mixed Fission 

Product

AIHL Radioactive Drain System plugged 

up and flooded controlled area.

A0050 7/22/1970 Cell 3 AIHL Mixed Fission 

Product

Small alcohol fire during disassembly of 

NaK bonded fuel element

A0051 10/11/1970 OP Gallery AIHL Ra Th High airborne alarm after-hours due to 

increase in natural activity.

A0052 5/19/1971 DECON 4 AIHL Mixed Fission 

Product

A major fire resulted during draining of 

NaK tank in Decon 4.

A0054 7/29/1975 Cell 2 AIHL Kr-85 SRE fuel slug partially burned releasing 

radioactive gas and contamination.

A0060 9/30/1977 Basement Mixed Fission 

Product

Failure to comply with requirements 

caused contamination of basement floor.

A0067 4/25/1978 DECON 2 Mixed Fission 

Product

Employee contaminated with concentrate 

from sodium digestion.

A0069 5/3/1978 DECON 2 STILL Mixed Fission 

Product

A 5-minute alcohol evaporator fire in 

Decon 2.

A0076 2/9/1979 DECON 2 Mixed Fission 

Product

Alcohol distillation system spilled 

contaminating employee.

A0088 8/21/1981 Hot Storage Room Mixed Fission 

Product

Bags of radioactive waste were broken 

during loading, releasing contamination.

A0092 10/20/1981 AIR LOCK Ta-182 One Rad-Pac pin tossed in radioactive 

waste.

A0101 3/22/1982 Cell 1 Mixed Fission 

Product

Zirconium fuel pin cladding fines ignited 

causing high airborne contamination.

A0104 6/28/1982 D4 and Airlock Mixed Fission 

Product / Ta

Working two projects resulted in 

extremity exposure about guidelines.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

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A0105 7/13/1982 DECON 4 AIHL Mixed Fission 

Product

Faulty crimping seal resulted in alpha 

contamination of employees.

A0325 8/16/1982 Liquid Waste BL RAS alarm response at RIHL radioactive 

liquid waste bldg.

A0110 10/11/1982 DECON 4 AIHL Pu/Mixed 

Fission Product

Leaking glove box glove caused crosscontamination of two employees.

A0112 12/22/1982 DECON 4 AIHL Employee experienced a seizure while 

working in a controlled area.

A0262 4/21/1983 MBA 54 Pu Incorrect I.D. of fuel rod resulted in 

grams of Pu in excess of MBA limit

A0120 7/15/1983 Alpha Box D-4 Pu/Mixed 

Fission Product

Loss of glove seal during change-out on 

alpha box contaminated employee.

A0119 8/8/1983 Decon 3 Mixed Fission 

Product

Leaking transfer tube contaminated 

employee and decon room.

A0118 10/11/1983 Cell 3 Face Mixed Fission 

Product

Radioactive liquid siphoned from cell 

contaminating wide area of building.

A0122 1/30/1984 RIHL Ta-182 Contaminated electrode was worked on 

a clean grinder and tracked out of 

building.

A0124 3/22/1984 Cell 3 Mixed Fission 

Product

Protective clothing failure allowed 

caustics to burn and contaminate 

employee.

A0125 8/20/1984 Cell 3 During in-cell repair an employee 

bumped his head, drawing blood.

A0126 8/21/1984 DECON / Cell 3 Mixed Fission 

Product

Employee contaminated smearing lowlevel tools with high-level contamination.

A0127 10/15/1984 Cell 3 Mixed Fission 

Product

During remote decon the alcohol caught 

fire.

A0128 10/28/1984 Slave Shop Mixed Fission 

Product

Employee contaminated hands handling 

bag of radioactive waste.

A0129 12/6/1984 Cell 1 Mixed Fission 

Product

Failure of protective clothing seal 

contaminated employee.

A0130 12/10/1984 Cell 1 Mixed Fission 

Product

Failure of protective clothing seal 

contaminated employee.

A0131 12/10/1984 Cell 1 Mixed Fission 

Product

Improper unsuiting contaminated 

employee.

A0133 12/10/1984 Cell 1 Mixed Fission 

Product

During decon of Cell 1 extremity 

exposure exceeded administrative 

guidelines.

A0134 1/14/1985 Decon 1 Mixed Fission 

Product

Bag radioactive waste came untaped 

and contaminated employee’s pants.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

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A0132 1/15/1985 Slave Shop Mixed Fission 

Product

Employee cut finger with contaminated 

screwdriver.

A0135 2/20/1985 SHOP Mixed Fission 

Product

Employee cut finger on Cell 4 end door 

gear shaft during repair.

A0136 3/4/1985 Cell 1 Mixed Fission 

Product

Contaminated sharp object punctured 

protective clothing contaminating shoes.

A0137 3/20/1985 Service Gallery Mixed Fission 

Product

Contaminated item from cell 1 

contaminated personnel and area with 

opened.

A0139 6/6/1985 SLAVE SHOP Mixed Fission 

Product

Decon with acid resulted in burning and 

contamination of employee.

A0140 8/12/1985 Decon 1 Mixed Fission 

Product

Contaminated puncture wound occurred 

during packaging radioactive glass 

waste.

A0142 10/4/1985 Service Gallery Mixed Fission 

Product

Personal clothing worn under protective 

clothing became contaminated.

A0141 10/17/1985 Cell 3 Mixed Fission 

Product

Sharp object in cell 3 punctured 

protective clothing contaminating shoes.

A0147 10/25/1985 Cell 3 Mixed Fission 

Product

Protective clothing leached 

contamination onto employee.

A0148 11/4/1985 Cell 3 Mixed Fission 

Product

Protective clothing leached 

contamination onto employee.

A0145 11/13/1985 Cell 3 Mixed Fission 

Product

Protective clothing leached 

contamination onto personal shoes and 

socks.

A0143 11/20/1985 Cell 4 Mixed Fission 

Product

Protective clothing leached 

contamination onto personal shoes and 

socks.

A0144 11/21/1985 Cell 3 Mixed Fission 

Product

Protective clothing leached 

contamination onto personal shoes and 

socks.

A0146 11/22/1985 DECON 4 AIHL Mixed Fission 

Product

Protective clothing leached 

contamination onto personal shoes and 

socks.

A0149 12/10/1985 Hot Storage Room Mixed Fission 

Product

Employee pinched hand while loading 

radioactive waste in box, drawing blood.

A0150 1/14/1986 ALL OF AIHL A review of recent non-reportable 

incidents at AIHL/RIHL.

A0151 2/4/1986 Cell 4 Loss of breathing air supply during cell 

entry.

A0152 2/19/1986 RM 139 Maintenance employee cut finger on 

radioactive exhaust plenum.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

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A0155 3/4/1986 Service Gallery Alconox and water solution splashed in 

eye of employee.

A0156 4/16/1986 Hot Storage Room Sr-90 Inventory revealed check source missing.

A0165 10/28/1986 Cell 4 U/Mixed 

Fission Product

Ignition of Fermi saw chips during 

disassembly in Cell 4.

A0173 3/12/1987 Cell 4 Employee passed out from heat 

exhaustion in radioactive environment of 

cell #4.

A0174 7/10/1987 Cell 2 Mixed Fission 

Product

Employee contaminated wrist during 

doffing protective clothing.

A0175 7/14/1987 Cell 2 Mixed Fission 

Product

Protective clothing seal failure resulting 

in contamination of employee’s neck.

A0177 7/20/1987 Cell 2 Mixed Fission 

Product

Exist survey revealed neck 

contamination on employee.

A0178 10/15/1987 Service Gallery Mixed Fission 

Product

Waste handling exit survey revealed 

contaminated pants.

A0181 12/1/1987 Operating Gallery Employee pinched finger and drew blood 

while working with a radioactive shield 

plug.

A0182 12/17/1987 Sump Pump Employees entered RAAEP area and 

confined space without permits or 

approvals.*

A0185 6/13/1988 Basement Alcove Faulty meter alarmed security guards.

A0190 9/30/1988 Cell 1 Mixed Fission 

Product

Exit survey revealed contamination on 

knee and forearms after cell work.

A0192 11/14/1988 Storage Room 153 Employee received puncture would in 

contaminated area.

A0193 1/4/1989 Cell 1 Mixed Fission 

Product

Improper seal of protective equipment 

allowed radioactive inhalation by 

employee.

A0198 06/6/1989 Cell 1 Mixed Fission 

Product

Employee exit survey revealed 

contamination from in-cell operation

A0199 0629/1989 Cell 2 & DECON 1 Mixed Fission 

Product

Exit survey revealed general point of 

contact contamination

A0201 10/4/1989 Glove Box RM 139 Pu Puncture wound in plutonium glove box 

room

A0587 12/15/1989 RIHL Airlock Filter failure on vacuum cleaner during 

scabbling of contaminated concrete.

A0205 4/18/1990 Cell 2 Mixed Fission 

Product

Contaminated steel chip projectile 

penetrated wrist cutting radial artery.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

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A0207 7/17/1990 Cell 3 Mixed Fission 

Product

Ladder slid out from under employee 

while sledge hammering in cell.

A0210 10/17/1990 Service Gallery Mixed Fission 

Product

A dropped duct section missed 

employees but caused airborne activity.

A0213 12/17/1990 T468 Radioactive 

Liquid BL

False reading on RAS detector caused 

alarm.

A0267 8/26/1991 OP Gallery Contamination barriers blown out into 

clean area by grit blaster in cell.

A0571 6/17/1993 RIHL basement Mixed Fission 

Product

Employee became contaminated during 

D&D of radioactive exhaust duct.

A0575 9/15/1993 RIHL CELL 3 Dosimeter went off scale during work in 

high radiation area.

A0662 5/10/1995 Area IV Decon 3 Dumping of vacuum catch-drum without 

notification.

A0668 4/16/1996 SS T020 Rm 147 Particle on right shoe.

A0675 3/26/1997 Parking Lot Loss of control of radioactive material.

A0677 7/2/1997 Frisked shoe Contaminated pipe.

A0679 8/15/1997 Rubber boot Rubber boot had a hole in it.

A0678 8/15/1997 Left Hand Contractor removed paper suit without 

gloves.

A0689 10/a/1997 To20 Parking Lot Shipment of clean T020 Building debris 

prior to state confirmatory survey.

A0682 3/20/1998 So. Storage Yard Unauthorized move of radioactive 

material.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

*The acronym RAAEP cannot be defined from available documents. On December 17, 1987, an incident occurred at 

a sump pump and employees entered a confined space without the proper permits or approval (Incident Report 

A0182). The original incident report is identified as missing from Boeing’s incident files. No other details are available 

on this incident at this time.

9.1.4 Incidents of Known Environmental Releases

• On December 5, 1959 radioactive liquid spilled during vendor pick up and contaminated the 

vendor’s truck. The Coastwise Marine Disposal Company arrived to pick up approximately 

2,000 gallons of liquid waste from the hold-up tanks in the basement of Building 4020. The truck 

was parked adjacent to the north loading dock of the building and the flow of liquid was 

controlled by a valave near the dock. The van was loaded with 51 empty 55-gallon drums and 

lids. The wooden floor of the van was covered with a white powder used to solidify liquid waste, 

but the floor was not watertight. A long plank was resting on top of the drums for walking from 

the back to the front of the van and a long stick was used for stirring the solidifying agent into 

the wastewater. The first drum of water was pumped out to the 50-gallon mark and then the 

solidifying agent was added to demonstrate its capabilities. 

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When the agent was dumped into the drum, a small amount of radioactive water splattered out 

into the truck body and flooring and also onto some planks in the back of the open van. The 

vendor was directed to put some covering down in the back of the open van. A piece of plastic 

sheeting was provided and the vendors were told not to move off the plastic without removing all 

protective clothing and leaving it on the plastic sheeting. A bit later, a small amount of water was 

observed coming out of the van body and running onto the rear wheels of the van. Solidifying 

agent was immediately thrown on the floor of the van to try and soak up the radioactive water 

that was trickling down onto the wheels and ground. Pumpin of the water was halted until a 

satisfactory means of controlling the water inside the van body was devised so that more 

radioactive water would not be spilled onto the asphalt loading pad. The wastewater transfer 

was stopped until the degree of contamination could be determined. 

Smears were taken of the truck wheels, axle and body and the asphalt area under the truck. 

Twenty smears were taken and averaged approximately 75 dpm / 100 cm2 beta gamma. The 

background radiation intensity had not risen on the survey insrument over normal levels. 

Pumping operations were discontinued until the truck and area were decontaminated to a level 

of 40 dpm/100 cm2 beta gamma, the Atomics International contamination limit for an 

uncontrolled area. The truck was wiped down and allowed to leave the area after smears 

showed the beta gamma measurement was less than 30 dpm/100 cm2 (Incident Report 

A0001).1

The quantity of liquid waste released to the ground surface was not reported in the incident 

report. Additionally, the report does not describe any cleanup efforts that were made on the 

asphalt surface. Resulting contamination on the asphalt surface would have flowed across the 

parking lot and into the drainage ditch. 

• On May 9, 1962 a fire occurred in Cell #4 of building 4020. Prior to the fire, an irradiated 

capsule containing uranium slugs in a stainless steel tube bonded with sodium-potassium 

allow (NaK) was being disassembled in the Hot Lab. The capsule was separated into two 

pieces and during decladding of the longer piece in Call 4 the liquid nitrogen supply necessary 

to maintain the cell atmosphere ran out. Examination of the fuel was discontinued and the fuel 

slugs were stored under amyl alcohol in a metal tray and covered with a metal plate.

It was thought that a piece of partially oxidized NaK may have fallen onto the cell floor during 

the transfer of slugs to the metal tray. When the humidity in the cell increased, the NaK 

oxidized at a rate that generated sufficient heat to ignite the cell floor covering and the alcohol 

vapors outside the metal tray. Alcohol inside the tray also evaporated exposing the fuel slugs 

to air and allowing NaK inside the slugs to oxidize. Three of the nine fuel slugs were destroyed 

completely. According to the incident report, at the time of the fire, the only volatile 

radioisotopes remaining in the fuel in significant quantities were krypton-85 (Kr-85) and 

iodine-131 (I-131). A maximum of 7.8 x 10 5 micro curies per cubic centimeter (μCi/cc) of Kr-85

and 1.3x10 6 μCi/cc of I-131 were estimated to be released through the exhaust stack. The 

incident report summarized that the release could not have resulted in significant exposure to 

1 EPA/HGL, Final HSA 5D, p. 47, citation 1:

1 Clow, H.E., Atomics International Internal Letter, Re: Liquid Waste Removal at CDHC, January 20, 1960. 

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personnel or surrounding areas (Incident Report A0317). 1

• On May 31, 1962, a portable radioactive liquid holdup tank was being filled from the 

radioactive holdup tank inside Building 4020. The operator’s attention was diverted by a 

leaking pipe fitting at the bottom of the portable tank. While the operator was tightening the 

pipe fitting, the portable tank overflowed and a maximum of 50 gallons of liquid was spilled. 

Analysis of the liquid remaining in the tank showed an activity level of 2.2x10-3 μCi/cc,

indicating a total activity of 420 μCi involved in the spill. The portable tank was on a concrete 

pad surrounded by an asphalt apron that sloped to the roadway. The spilled liquid immediately 

flowed down slope to the road and then to a drainage ditch, which connected to the “cold 

sewerage system.” The pattern of the spill is shown in a Building 4020 figure included in this 

report (Figure 2.2.1s).

Decontamination procedures began immediately. Within minutes of the spill the area was 

hosed down with fire hoses to prevent any fixation of particulate matter and to dilute liquid 

already in the drainage ditch. The area was roped off and decontamination of the asphalt 

began with mops and brushes. Prior to decontamination smears of the asphalt and concrete 

pad were as high as 50,000 dpm. After a series of scrubbing procedures the area was 

decontaminated to the point that no detectable removable contamination was found. The area 

was then surveyed with a portable floor monitor and a 2612 G.M. survey instrument to locate 

areas of fixed contamination.

Thirty-seven “hot-spots” approximately 2 inches in diameter were found to have radiation 

levels in excess of 1 millirad per hour (mrad/hr). Levels ranged from 2 mrad/hr to 130 mrad/hr 

beta-gamma. These areas were marked with white paint and chipped away. Following

chipping, no detectable activity remained. A survey of the drainage ditch with the 2612 G.M. 

survey instrument found no activity above normal background. Sixty samples were taken from 

the bed of the drainage ditch from the site of the spill to the sewage treatment facility. 

Contamination levels ranged from less than 30 dpm to greater than 500 dpm beta-gamma, 

with an average of approximately 100 dpm beta- gamma contamination. The ditch was 

sampled from the site of the incident to its entry into the sewage treatment facility. The 

contamination appeared to be evenly distributed throughout this 400 yard length. Surface soil 

samples of the ditch showed contamination in the amount of 0.06 μCi/ft2 (using a factor of 10 

for self absorption of the sample). The entire length of the ditch was approximately 1,500 

square feet (ft2), so it was estimated that the total contamination remaining in the ditch was 90 

μCi. The incident report notes that removal of the contamination was probably not warranted 

by the contamination levels, the “narrow, tortuous and deep” nature of the ditch, and the fact 

that the ditch traversed unoccupied areas. Additional flushing was seen as increasing the rate 

of contamination in the sewer system. It appears from the incident report that no other cleanup 

was performed, although recommendations for prevention of similar incidents were made 

1 EPA/HGL, Final HSA 5D, p. 48, citation 1:

Remley, M.E., Atomics International, to Levy, J. V., U.S. Atomic Energy Commission, Re: Fire in the 

Component Development Hot Cell, May 9, June 7, 1962. 

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(Incident Report A0016). 1

• On July 2, 1963, an area in the radioactive holding area of Building 4020 approximately 15 

feet by 3 feet was contaminated to a level of 7.4x104 μCi/cc. The incident report did not 

provide a detailed description of the location of this contamination. The Health and Safety 

Section requested the area be decontaminated and sealed before the contamination spread to 

other clean areas. This area was contaminated during a spill of “EB 40” from the Organic 

Moderated Reactor Experiment (OMRE) cask (Incident Report A0022). 2

• On July 8, 1963, three Atomics International employees working at Building 4020 were 

contaminated while moving a cask containing radioactive waste material. The cask was 

transferred from Cell #1 to a building airlock using a fork lift and special lifting rig. At the airlock 

the cask was placed in clean plastic bags. The cask was then transferred to a clean fork lift 

outside the building, without changing the contaminated lifting rig that had been used inside 

the building. Three employees working outside the building had contact with the contaminated 

lifting rig. Because the three employees were involved in different stages of the procedure and 

had continued on to other activities in the day, personnel contamination was discovered over 

the course of a day. One employee had to have his car and home surveyed as he had gone 

home before the contaminated lifting rig had been discovered. A spot of contamination was 

noted on the employee’s shirt and his trousers at 0.03 mr/hr and 0.7 mr/hr, respectively; 

however, it was not felt that contamination had spread outside of Atomics International 

premises (units may be incorrectly presented in the source document). The lifting rig was 

decontaminated and personnel surveys indicated no significant external or internal radiation 

exposure (Incident Report A0316). 3

• On September 25, 1963, a fire occurred in Cell #3 of Building 4020. The fire occurred during 

the dissolving of NaK bonding material on fuel cladding using butyl alcohol, a routine 

operation. Concurrent with the start of the fire, the operating gallery continuous air monitor 

increased approximately 400 Ci/m beta-gamma. A high volume air sample taken in front of 

Cell #3 after the fire began indicated 9.4x10 10 μCi/cc beta-gamma, a factor of 10 higher than 

a sample taken a short time before the fire. The fire burned a piece of wood, Tygon tubing, 

butyl alcohol, blotter paper, and the plastic boot that helps keep the contamination levels of 

the manipulator down. The fire was extinguished within 10 minutes using a nitrogen purge. 

The airborne concentrations were significant and bioassays of all personnel in the area at the 

time of the fire indicated a positive and significant intake of mixed fission product. If the cell 

1 EPA/HGL, Final HSA 5D, p. 49, citations 1&2:

1 

Heine, W.F., Atomics International Internal Letter, Re: Radioactive Spill at CDHC Bldg. 20 Santa Susana, 

June 13, 1962.

2 Coonce, G.L., Atomics International Internal Letter, Re: Contamination Survey, June 7, 1962. 

2 EPA/HGL, Final HSA 5D, p. 49, citation 3:

3 Lane, W.D., North American Aviation, Inc. Internal Letter, Re: Decontamination of R/A Holding Area at 

CDHC, July 2, 1963. 

3 EPA/HGL, Final HSA 5D, p. 50, citation 1:

1 

Badger, F.H., North American Aviation Internal Letter, Re: Investigation of In-Cell Fire at CDHC, November 

4, 1963. 

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had not recently undergone a major cleanup, the airborne release could have been more 

serious. During investigation of the incident it was discovered that air flow in the building was 

below design standards in some areas (Incident Report A0027). 1

• On November 24, 1964 a radioactive spill resulting from the overflow of a 500-gallon portable 

liquid holdup tank occurred in the north loading dock area of Building 4020. Transfer of the 

liquid waste from inside of Building 4020 had stopped when the pump malfunctioned leaving 

the transfer tank half full. Repairs were made and pumping was restarted at a rate of 30 

gallons per minute. The operator, unaware the pump was working at full rate, was 

approximately 12 feet from the tank when it overflowed. The shutoff valve was immediately 

closed and a maximum of 25 gallons of liquid was lost at the north loading dock area. 

Analysis of the liquid remaining in the transfer line showed an activity of 2.4x10 1 μCi/cc

indicating a total release of 2.3x104 μCi/cc was involved in the spill. The radioactive liquid 

included in Cs-137, zirconium-95 and niobium-95. Approximately 300 ft.2 of cement pad and 

asphalt were contaminated. The spill area was outlined with spray paint and desiccant was 

placed in the larger puddles to control runoff. A radiation survey indicated general levels of 

10-15 mrad/hr beta-gamma activity and a maximum of 20 mrad/hr beta-gamma at 

approximately 2 inches above the water on the ground. Preliminary decontamination efforts 

included scrubbing and vacuum cleaning. The gross contamination of the cement pad was 

reduced to 1 mrad/hr at the surface and less than 10,000 dpm beta-gamma on a tape smear. 

Levels on the asphalt, which was covered with a primer coat of paint, were 20 mrad/hr betagamma. Final decontamination efforts were completed on December 2, 1964. A final survey 

showed the cement pad to have 0.3 – 0.5 mrad/hr beta-gamma at the surface, less than 150 

dpm beta gamma on tape smear, and less than 30 dpm/cm2 beta gamma on a normal smear, 

with three small surface cracks indicating 1.0 mrad/hr. A second coating of primer was applied 

to the asphalt and a maximum radiation level of 20 mrad/hr beta-gamma was embedded in 

the asphalt. No removable contamination was evident after the second coat of primer. The 

contaminated asphalt was not removed. Health and Safety determined that no further 

decontamination be performed, but the area would be routinely surveyed for future smearable 

contamination (A0033).2

• On May 27, 1965, Cell #3 was under a nitrogen purge to accommodate the decladding of 

irradiated uranium carbide fuel. The cladding was cut and NaK oozed out. An air monitor at 

the north end of the operating gallery started to increase, setting off the primary alarm (1,000 

Ci/m) and then the upper alarm point of 3,000 Ci/m. The 12-minute air sample indicated 

1.5x10 8 μCi/cc. A second air sample taken 30 minutes after the initial alarm was 2.3 x 10 9 

μCi/cc. Smears taken throughout the area were less than 30 dpm. A third air sample taken fifty 

minutes after the initial alarm indicated 7.3 x 10 10 μCi/cc beta-gamma, immediate count. This 

sample was recounted 90 minutes later and was 1.4x10 10 μCi/cc beta gamma. The stack 

monitor indicated two distinct increases in activity. The maximum count rate was 500 Ci/m or 

1 EPA/HGL, Final HSA 5D, p. 49, citation 4:

4 Remley, M.E., Atomics International, to Levy, J.V., U.S. Atomic Energy Commission, Re: Unusual Incident 

at the Component Development Hot Cell, July 15, 1963. 

2 EPA/HGL, Final HSA 5D, p. 50, citation 2:

2 Ericson, G.I., Atomics International Internal Letter, Re: Radioactive Liquid Spill at CDHC Building 020, 

January 12, 1965. 

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an estimated 6.5 x 10 8 μCi/cc. The total gaseous release at the stack was estimated to be .

012 mCi over a two hour period. No personnel or equipment contamination was detected 

(A0035).1

• On August 12, 1965 rainwater from a one-way waste cask was being dumped in a “nonradioactive outside storage area” at the Hot Lab when an “unusually large” volume of water 

spilled out of the cask. The incident report did not provide a detailed description of the location 

of the incident. A survey of the water found it contaminated with 300 μCi of mixed fission 

product in suspension. The runoff was contained and the area secured. An absorbent was 

placed on the liquid and contamination was measured up to 20 mrad/hr. After vacuuming the 

absorbent and scrubbing the area with “Radiac Wash” the contamination was brought down to 

0.2 mrad/hr. Contamination was confined to a 3-foot by 10-foot section of asphalt and no 

personnel contamination resulted from the incident. It was surmised that the cask had been 

improperly tagged and had possible internal contamination (A0441).2

• On October 30, 1967 a radioactive liquid drain system clogged, spreading contaminated water 

throughout controlled areas of Building 4020. The water was pumped into the RMHF 500-

gallon transfer tank, but removable beta contamination levels as high as 1.3x106 dpm/100 cm2 

remained. It was reported that all areas were successfully decontaminated (A0627); however, 

details of the decontamination efforts were not provided. 3

• On May 19, 1971 there was a fire in Decon #4 during the disposal of 100 gallons of liquid 

NaK, which contained 100 μCi of mixed fission products. A hole in a tank fill-line caused the 

release of about 25 gallons of contaminated NaK, which then caught fire. Nearly all 

contamination was contained in Building 4020. Airborne activity and surface radiological 

contamination concentrations inside the building from the event range from 2% to 20% of 

permissible concentration for occupational use and the average concentration released 

through the stack to the outside of the facility was reportedly about 5% of the permitted 

concentration for an unrestricted area (A0052).

• On July 29, 1975 a single irradiated SRE fuel slug ignited and partially burned in Cell #2 

releasing an estimated 0.77 mCi of Kr-85 gas out of the stack and resulting in the momentary 

release of airborne radioactivity to the operating gallery. The fire was extinguished in 20 

minutes with a nitrogen purge. The incident did not result in any overexposure and did not 

constitute a reportable incident. The SRE Core I fuel was being prepared for reprocessing by 

removing all NaK from the fuel slugs.

The planned NaK distillation proved ineffective on badly damaged fuel and a new system was 

1 EPA/HGL, Final HSA 5D, p. 51, citation 1:

1 

Badger, F.H., Atomics International Internal Letter, Re: Incident Report, June 25, 1965. 

2 EPA/HGL, Final HSA 5D, p. 51, citation 2:

2 Badger, F.H., Atomics International Internal Letter, Re: Incident Report, Non-Radioactive Outside Storage 

Area at AIHL Bldg. 020, SS, 8-12-65, August 16, 1965.

3 EPA/HGL, Final HSA 5D, p. 51, citation 3:

3 Alexander, R.E., Atomics International Internal Letter, Re: Radiation Safety Unit Weekly Newsletter for 

Period Ending November 4, 1967, November 10, 1967. 

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developed in Cell #2 to heat the fuel slugs and sweat the NaK out then dissolve it in alcohol. 

The slug in question had been soaked in alcohol for several days and ignited when it was 

heated with a high intensity lamp. The continuous air monitor in the operating gallery 

stabilized at 6,000 cpm. The stack monitor indicated 700 cpm or 6.5 x 10 6 μCi/cc going out 

the stack. Routine analysis of basement air samples indicated 1.87 x 10 8 μCi/cc beta gamma 

for a seven day period following the fire. This was a result of airborne activity released to the 

operating gallery being drawn down the electrical breezeway into up-air dampers in the 

basement (A0054).1

• On October 20, 1981 a “clean” ladder located in an uncontrolled area and scheduled to be 

loaned out was routinely surveyed and found to be contaminated. The feet of the ladder 

indicated the highest activity with 30 mrad/hr beta-gamma and 5,000 dpm/100 cm2 betagamma removable contamination. A general contamination level of 1,000 to 2000 dpm/100 

cm2 was reported. No significant alpha activity was detected. The ladder was bagged and 

placed in a controlled area. All personnel were surveyed and the facility was smeared. No 

other activity was detected, with the exception of 300 dpm/100cm2 beta-gamma in Room 139 

alongside a file cabinet. Further investigation found the ladder was from the radiologicallycontrolled service galley area of Building 4020. It had been used in painting Decon #4 in May 

1980 and then accounting of the ladder after that time became vague (A0092).2

• On December 18, 1981 a survey of the “Rad Pac” loading area of the airlock indicated an 

abnormally high reading from a low-level waste can. A controlled search of the waste revealed 

a “Rad Pac” pin entangled in a Kimwipe. The tantalum-182 source pin measured 16.5 mCi of 

radioactivity. Typically, sources are calibrated individually and stored in numbered holes in a 

lead brick. Then the specific source pins for the customer are grouped for sealing in capsules 

and removed to a storage pig. The remaining source pins are removed and placed in a 

storage vial on a Kimwipe. The vial is closed and placed outside the storage pig. A mechanic 

missed placing one pin in the vial and it became hidden in a fold of the Kimwipe (A0094).3

• On March 22, 1982 zirconium fuel pin fines (fine particles) from grinding of Fermi fuel in Cell 

#1 reacted violently during an oxidation process and nitrogen purge resulting in high airborne 

activity in an uncontrolled area of Building 4020. Fermi fuel pins are clad with zirconium. In 

order to recover the 25% enriched uranium remaining the fuel the zirconium must be 

removed. Center-less grinding of the fuel was the preferred method. The grindings would be 

filtered from the coolant and oxidized in a furnace to render the normally unpredictable 

flammability of the zirconium fines harmless. The wet filter papers with wet fines were stored 

in a sealed tube until loaded into the furnace. The cell atmosphere would be reduced to less 

than 5% oxygen by a nitrogen purge. The furnace and fines were put under a vacuum to 

1 EPA/HGL, Final HSA 5D, p. 52, citation 1:

1 

Badger, F.H., Rockwell International Internal Letter, Re: Fuel Slug Fire – Cell 2, July 29, 1975, September

29, 1975. 

2 EPA/HGL, Final HSA 5D, p. 52, citation 2:

2 Badger, F.H., Rockwell International Internal Letter, Re: Radiological Safety Incident Report, T020, October 

20, 1981, November 19, 1981. 

3 EPA/HGL, Final HSA 5D, p. 52, citation 3:

3 Badger, F.H., Rockwell International Internal Letter, Re: Radiological Safety Incident Report, T020 Airlock, 

12/18/81, December 23, 1981. 

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remove water vapor and then the furnace was turned on to 1,650 degrees Fahrenheit until the 

fines were oxidized. The oxide would then be placed in a storage can for analysis and 

disposition.

During the incident, an “O” ring failed and created a loss of vacuum, allowing the “O” ring to 

burn. The electrical power to the operation was shut down and personnel were evacuated. 

Initial air samples and smears were normal and personnel returned to work. The oxygen 

content in Cell #1 was below 5% so the high volume nitrogen purge was switched to low 

volume and the furnace cool-down air jets were started. A bit later, an air monitor in the 

operating gallery was checked again and showed a sharp increase in beta-gamma activity 

(~1,500 cpm). Additionally, the stack monitor showed a 400 cpm increase in particulate activity 

and a basement air monitor alarmed at ~3,500 cpm. A dilution fan was started and a high 

volume air sample was taken showing 3x10 9 μCi/cc. The nitrogen purge rate was changed 

and the air monitor in the operating gallery stabilized.

Air samples were collected through the building and analyzed. Activities were calculated on 

the basis of a one hour release. The maximum release occurred in the basement, an 

unoccupied area, at 1.3x10 8 μCi/cc. The maximum release in an occupied area was 3.4 x 

10-9 μCi/cc at the face of Cell #2. The activity in the stack sample with a one hour base was 

1.2 x 10-10 μCi/cc. Analysis of the air samples indicated Cs-137 as the dominant isotope and 

possible antimony-125 (Sb-125). No other isotopes were identified. Investigation found 

controllers on the cell purge systems were not working properly and cracks were noted in the 

radioactive exhaust ducts in the basement. The release of airborne activity was found to be 

the result of faulty controllers and activation of the furnace air jet cooling system while on a 

low-volume nitrogen purge. All operations were halted until corrective actions and another 

safety review of the operations could be conducted (A0101).1

• On March 26, 1997 a block with low level contamination was inadvertently not surveyed and 

released as clean. It was transferred to a radiologically uncontrolled area at the Hot Lab. The 

contaminated block (300 cpm) was discovered during a state confirmation survey. The block 

was moved into a radioactive material area and the contamination was reduced to less than 

20 dpm/cm2 alpha, less than 100 dpm/cm2 beta, less than 100 cpm above lowest 

background, and less than 5 μr/hr (units may be incorrectly presented in the source 

document) above ambient background (A0675).2

• On July 2, 1997 a block originally connected to a highly contaminated drain pipe and believed 

to be filled with concrete slurry was removed, leaking contamination onto the soil and asphalt. 

The discovery was made after an employee working in the area found contamination on his 

shoe (800 cpm). The affected area was roped off and highlighted for decontamination. The 

cause of the incident was determined to be incorrect handling procedures for the block, which 

1 EPA/HGL, Final HSA 5D, p. 53, citations 1 & 2:

1 Badger, F.H., Rockwell International Internal Letter, Re: Radiological Safety Incident Report, T020 – Cell 1, 

March 22, 1982, March 29, 1982.

2 

Badger, F.H., Rockwell International Internal Letter, Re: Final Report – Fermi Zirconium Fines Explosion 

Bldg. 20, April 2, 1982. 

2 EPA/HGL, Final HSA 5D, p. 53, citation 3:

3 Harcombe, R., Boeing Internal Letter, Re: Incident Report, Loss of Control of Radioactive Material, A0675, 

T020 Parking Lot, 3/26/97, April 9, 1997. 

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was not originally known to be connected to the contaminated drain pipe (A0677). Additional 

information regarding the decontamination of this area could not be located by the EPA

research team in the documents provided.1

• On October 15, 1997 it was discovered that a roll-off of scrap material from Building 4020 was 

released to an outside contractor without a California Department of Health Services 

confirmatory survey. The materials in the roll-off had been surveyed clean, the survey 

documented, and the roll-off released from radiological control by Rocketdyne Health Physics. 

The state was not given the opportunity to perform a confirmatory survey as required under 

Rocketdyne’s license. A notice of violation was issued. Rocketdyne implemented corrective 

procedural actions to prevent a recurrence (A0689).2

• On March 17, 1998 four unauthorized contract personnel were loading concrete blocks on a 

flatbed truck from the radioactive materials area at Building 4020. The personnel were not 

“radiological worker qualified” and thus not authorized for area access. The “Radioactive 

Materials Area” posting boundary rope had been removed without authorization. A verbal stopwork was issued and personnel were released from the area after being surveyed for 

radioactive contamination. All contract labor received additional training, and it was 

determined that the contract workers were allowed through control gates prior to the site 

supervisor being notified of their arrival (A0682).3 

1 EPA/HGL, Final HSA 5D, p. 54, citation 1:

1 Deschamps, R., Boeing Internal Letter, Re: Incident Report, Contamination Found on Employee’s Shoe, 

A0677, T020, 7/2/97, April 15, 1998. 

2 EPA/HGL, Final HSA 5D, p. 54, citation 2:

2 Guy, E., Energy Technology Engineering Center, Occurrence Report, SAN—ETEC-GENL-1997-0001, 

January 29, 1998. 

3 EPA/HGL, Final HSA 5D, p. 54, citation 3:

3 Schaeppi, W., Boeing Internal Letter, Re: Unauthorized Movement of Radioactive Material, March 23, 

1998. 

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10.0 Fabrication, Use and Storage of Radioactive Sources 

According to the 2006 Site Profile, SSFL operations required many instruments for detecting 

and measuring radioactivity. These instruments must be calibrated periodically using known 

quantities and types of radioactivity called sources, which are sealed containers that contain 

small measured quantities of radioisotopes. Sources are also used for some forms of 

radiography, irradiation testing, and other applications. Sources were manufactured in the SSFL

Hot Laboratory (Building 4020) and used in various facilities at SSFL and elsewhere. 

Approximately 140,000 Ci of radioactive material (primarily Pm-147) were fabricated into 

sources at the Hot Laboratory, stored in secured locations and used under carefully controlled 

conditions. 

The HSA research team located more than one location where radiation sources of this type 

were produced or stored: 

Building 4011 - Not Included in 2006 Site Profile

Building 4029 - Included in the 2006 SSFL Site Profile (insufficiently described)

Building 4055 - Included in 2006 SSFL Profile (insufficiently described) 

A description of operations and processes associated with Building 4011 is provided in Section 

16.0, which contains information on buildings/facilities that were excluded in their entirety from 

the SSFL Site Profile. 

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Section 10.1 

Building Number:

Building Alias: 

4029 

029 / T029 

Building Name: Radiation Measurements Facility (RMF) - 1962

Hazardous Waste Materials Facility (HWMF) - 1974

Building Function: Storage of radioactive source materials

Storage of hazardous waste materials

Storage of sodium waste (un-permitted RCRA violation) 

Radionuclides of Concern: Co-60, Ra-226, Cs-137, Po-Be, Pu-Be 

10.1.1 Description of Operations & Processes 

Building 4029 was constructed between 1962 and 1965 to store radioactive source materials, 

including PoBe, PuBe, Co-60, Ra-226 and Cs-137 for instrument calibration.1 The table below 

shows the calibration sources used at Building 4029, the source strength, and the date the 

source strength was measured. 

Calibration Sources Used at Building 4029 

Source: ETEC Document, 029-AR-0001, “Final D&D Report for Building T029,” March 28, 1996

*A March 1964 incident resulted in the release of Ra-226 from this source capsule at Building 4029.

This incident is discussed in detail below. 

Source Source Strength (mCi) Date

Ra-226* 24.8 1960

Ra-226 132 1960

Ra-226 930 1960

Co-60 Unknown Not Applicable

PoBe Unknown Not Applicable

PuBe Unknown Not Applicable

Cs-137 5310 Sept. 1963

Cs-137 5260 Sept. 1963

Ra-226, Cs-137 and Co-60 were used to calibrate gamma-sensitive instruments while PoBe 

and PuBe were used to calibrate neutron-sensitive equipment. Based on information contained 

in the 1006 D&D report, the Ra-226 sources were replaced with two Cs-137 sources and were 

1 EPA/HGL, Final HSA 5A, p. 161, citation 669:

669 ETEC Document, 029-AR-0001, “Final D&D Report for Building T029,” March 28, 1996. 

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not used in the facility concurrently. The encapsulated Co-60 source was used in the early 

1960’s and was used on a limited basis. It was removed from the facility in about 1965.1 

PoBe and PuBe neutron sources were reportedly fully encapsulated and stored in a small 

concrete pit, which was also used to store gamma sources in lead “pigs” when not in use. The 

neutron storage facility was dismantled in 1964 or 1965. All radioactive source materials were 

reportedly removed by April 1974 and transferred to another facility. Building 4029 was partially 

decommissioned at that time. 

Beginning in 1974, Building 4029 was incorporated into the Hazardous Waste Management 

Facility (HWMF) as a storage area along with Building 4133, which served as a treatment 

facility. The HWMF was for the management of reactive metal waste, including Na, NaK, Li, and 

LiH2, and functioned to convert waste metallic sodium into sodium hydroxide. 

According to a 1988 DOE memorandum, ten cold traps containing sodium (a reactive 

hazardous waste) were stored outside the Building 4029-permitted storage area. While it was 

stated the sodium was contained and would not cause an environmental problem, the storage 

outside of Building 4029 was a violation of RCRA storage regulations. As a result, ETEC was in 

the process of developing sawing techniques to cut the taps to that they could be oxidized at the 

“burn pit.”2 

In 1996, a D&D report documented the work performed to remove residual radioactive 

contamination from a relatively inaccessible area of Building 4029. It was released without 

radiological restrictions in 1997. 

Barrels and drums were stored outside Building 4029 for a short period of time in the early 

1960’s. In 1989, a storage area was reported to be located near Building 4029 that contained a 

permitted long-term storage hazardous waste area. Reactive metals including sodium, 

potassium, sodium potassium, zirconium, and lithium metal were stored in 55-gallon drums. 

There were also apparently 20 “cold traps” containing reactive sodium metal stored outside 

1 EPA/HGL, Final HSA 5A, p. 162, citation 670, 671, 672:

670 ETEC Document, GEN-ZR-0006, “Radiological Survey of the Old Calibration Facility – Building T029,” 

August 19, 1988.

671 ETEC Document, 029-AR-0001, “Final D&D Report for Building T029,” March 28, 1996.

672 Rockwell International, Document N001DWP000024, “Radiological Decontamination of Building 029,”

July 31, 1989. 

2 EPA/HGL, Final HSA 5A, p. 163, citation 677-680:

677 ETEC Document, GEN-ZR-0006, “Radiological Survey of the Old Calibration Facility – Building T029,” 

August 19, 1988.

678 The Boeing Company, Rocketdyne Propulsion and Power DOE Operations Annual Site Environmental 

Report 1997, A4CM-ZR-0012, November 23, 1998. 

679 The Boeing Company, Rocketdyne Propulsion and Power DOE Operations Annual Site Environmental 

Report 1998, RD99-115, September 22, 1999.

680 Scott, Randal, Memorandum Re: Survey Status Report, July 5, 1988. 

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Building 4029. The area where the traps were stored was not within the permitted hazardous 

waste storage area. Surveys of the 1960’s storage area found no detectable activity.1 

The preliminary MARSSIM Classification for Building 4029 is Class 1 based on its previous use 

and the unknown source of the fill material used to backfill excavations during D&D activities. 

10.1.2 Building 4029 Radiological Incident Reports 

There have been several incidents associated with Building 4029 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

10.1.3 Building 4029 Radiological Incident Report Summary - Data Provided by Boeing 

Incident 

File Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0029 11/11/1963 SSFL & DS131 CO-60 Investigation of discrepancies in 

quarterly radioactive byproduct 

material.

A00032 3/24/1964 Source Well Ra-226 Leaking calibration source 

contaminated facility and personnel.

A0367 6/24/1965 Calibration Fac. Ra-226 Alarm failed when source failed to 

return to safe storage.

A0577 11/20/1970 Calibration Fac. Cs-137 Calibration source encapsulation 

failed during use.

• On November 11, 1963 a health and safety supervisor was notified that six radioactive 

sources were reported missing from the Quarterly Radioactive By-Product Material Inventory 

for the third quarter of 1963. The June 30, 1963 inventory was eventually found to be incorrect 

and all the sources were accounted for, with the exception of the 0.1 mC Co-60 source, which 

was believed to have been disposed of from the waste disposal shed (Building 131 at 

Headquarters [DeSoto]) a year earlier. The incident report does not provide any indication that 

a release occurred as a result of the incident (A0029).2

• On March 24, 1964 a leaking calibration source contaminated the building and personnel with 

24.8 mCi Ra-226. According to the incident report, the source was one of three containing 

1 EPA/HGL, Final HSA 5A, p. 168, citations 705-706:

705 ETEC Document, GEN-ZR-0006, “Radiological Survey of the Old Calibration Facility – Building T029,” 

August 19, 1988.

706 Schiffman, Joel, United Stated Department of Energy Environmental Survey Report, June 16, 1989. 

2 EPA/HGL, Final HSA 5A, p. 164, citation 689:

689 Badger, F.H., Internal Letter Re: Investigation of Reported Missing Radioactive Sources, January 8, 

1964. 

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24.8 mCi of Ra-226 as a bromide salt. 
 


 

On March 24, 1964 the source capsule became detached from the nylon string and fell into 

the bottom of the source thimble. The 13-foot fall of the source capsule cracked the outer 

plastic encapsulation surrounding the inner capsule and released some loose Ra-226. The 

contamination was primarily confined to the source storage well and the source thimble. Three 

personnel were contaminated with alpha activity on their hands with 2,500 dpm maximum as 

determined by a portable alpha survey meter. The personnel were evacuated to the Building 

4020 (Hot Lab) “hot change room” for decontamination. The area outside the source holder 

was found to be contaminated with “low-level” removable alpha contamination. 

Decontamination of the facility reduced fixed and removable contaminatin below detectable 

levels with one exception at the source storage well at the floor level, at 6 dpm/100 cm2 alpha.

On March25, 1964 the source was recovered from the source well and was placed in a lead 

shipping container to await disposition (A0032).1

• On June 24, 1965 personnel failed to return a 132 mg Ra-226 source to its well following 

instrument calibration. Following the departure of the personnel, fire Atomics International 

employees spent approximately 2,5 hours in the building installing a glove box and connecting 

it to the building’s absolute filter system. The exposed source was discovered the next day by 

Health and Safety personnel who were present at the building to conduct a smear survey of a 

55-gallon drum containing “a few kg of depleted UC.” The source was returned to the well and 

the personnel who entered the building during the period of exposure were monitored 

(A0367).2

• On November 20, 1970 the encapsulation of a 4.6 Ci Cs-137 calibration source failed during 

use and dropped 10 feet to the bottom of the well, resulting in the source getting stuck in the 

storage well. It was estimated that the external radiation level of the source was 16 R/hr one 

foot away from the source (A0577).3 

During a June 1989 DHS inspection of SSFL, DHS reported Rockwell International had received 

a state permit on December 30, 1983 for a sodium waste storage area at Building 4029. The 

inspection also stated that, “Rockwell failed to submit a written report to the Department (and 

EPA) within 15 days after determination of release of the radioactive materials at the sodium 

storage area (T029).” According to the report, the facility’s Industrial Hygienist “mentioned that 

they had found radioactive contamination in the sodium storage area.” However, no signs were 

1 EPA/HGL, Final HSA 5A, p. 164, citation 690:

690 Busick, D.D., Internal Letter Re: Report of Radioactive Contamination Incident of the Radiation

Measurement Facility – Building 029, April 10, 1964. 

2 EPA/HGL, Final HSA 5A, p. 165, citation 691:

691 Wildanger, A.W., Incident Report Re: Radiation Measurement Facility, A0367, July 9, 1965. 

3 EPA/HGL, Final HSA 5A, p. 165, citation 692:

692 Owen, R.K., Internal Letter Re: Incident Report – Sealed Source Capsule Failure at T029, December 2, 

1970. 

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posted indicating radioactive contamination.1 The HSA research team could not locate 

additional information regarding the contamination referenced by the Industrial Hygienist at 

Building 4029.

1 EPA/HGL, Final HSA 5A, p. 165, citation 693:

693 Motiafard, V., State of California, Department of Heath Services, Addendum Report, Rockwell 

International Corporation Rocketdyne Division, November 30, 1989. 

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11.0 Preparation of Radioactive Material for Disposal 

The 2006 SSFL Site Description indicates that the operation of nuclear reactors and other 

operations at SSFL (including fuel fabrication, reactor and fuel examination, etc.) generates 

radioactive waste and other radioactive material that must be disposed. However, the 2006 

SSFL Site Description states that no radioactive waste was ever incinerated at the RMHF, and 

the full description of operations at Building 4664 (Low-Level Radioactive Waste Incinerator) is 

omitted. 

In addition, the 2006 Site Profile implies that no radiological waste disposal occurred on-site, 

which is incorrect. The disposal of radioactive waste at SSFL is well documented. Radioactive 

waste was incinerated, buried, burned in open earthen “burn pits,” dumped, and submerged into 

ponds. In addition, spills and leaks were directed toward various drainage ditches that were 

directed into the Site Wide Water Reclaim System, which provided Rocketdyne personnel 

throughout Areas I, II, III and IV with water used to hose down work areas, etc. 

The 2006 Site Description provides an incomplete list of the buildings and facilities associated 

with RMHF, and provides an incomplete, insufficient representation of those that are included. In 

addition, only two minor incidents at the RMHF are referenced but the HSA provides robust 

summaries of over 60 incidents involving the RMHF, which are documented in the Incident 

Database. 

The following overview of the RMHF provides additional information, and the excluded locations 

are described in section 12.0. 

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11.1 OVERVIEW: Radioactive Materials Handling Facility (RMDF / RMHF) 

“The principal source of potential radiation dose to the public from SSFL is the RMDF.”1 

— United States Department of Energy 

The Radioactive Materials Disposal Facility (RMDF) was constructed in 1959. It was originally a 

support facility for the Systems for Nuclear Auxiliary Power (SNAP) program, the Sodium 

Reactor Experiment (SRE), and the Hallam Nuclear Power Facility. It was designed to handle 

the storage, volume reduction, packaging and shipping of SNAP and SRE radioactive waste, 

and to treat / process / and dispose of waste and radioactive water generated by all Atomics 

International projects. 

Although Boeing has asserted to EPA and provided data contained in the existing SSFL Site 

Description that asserts no radioactive waste was disposed in the RMHF Incinerator (Building 

4664), drawings of the building depict the low-level radioactive waste processing units and 

equipment. The Incinerator became operational October 5, 1964 and burned low-level 

radioactive waste at a rate of 100 pounds per hour until it was dismantled in the 1980’s, wherein 

the incinerator itself was disposed as radioactive waste.2 

Revelations of Building 4664 Incinerator operations call into question current data applied to 

dose reconstruction for SSFL employees. Emissions, releases, and airborne activity data 

relevant to incinerator operations may have been omitted, resulting in an inaccurate 

characterization of potential worker exposure resulting from operations at Building 4664. 

In 1989, DOE’s Office of Environmental Audit conducted a study of DOE operations at SSFL

and discovered that, “Lack of a meteorological tower [at SSFL] could result in inaccurate dose 

assessment in the event of an unscheduled release. Without a meteorological tower providing 

current wind speed and direction data, it is not possible to accurately predict the area of impact 

of an unscheduled [or scheduled] release. In addition, annual calculations of the air pathway 

dose to the population using the AIRDOS-EPA computer model may be in error, since old and 

inappropriate information from BAP [Burbank Airport] was used. It has been assumed that BAP

information was adequate because upper winds above the site are similar to those at BAP. 

However, data taken from an Area II meteorological tower operated in 1960-1961 showed 

significant differences.”3 

1 DOE Office of Environmental Audit, Environmental Survey Preliminary Report: DOE Activities at SSFL, 1989,

p. 4-31 - 4-33. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 136, citations 5, 6, 7, 8:

5 

Atomics International Drawing, Low Level Radioactive Waste Facility, Site Plan & Details, 303-664-C1, 

Illegible Date.

6 

Atomics International Drawing, Modification for Low Level Radioactive Waste Equipment, Bldg. 664 –

Plans, Sections & Details, 303-664-S4, January 1966.

7 

Atomics International Drawing, Low Level Radioactive Waste Facility 664, Flow Diagram, 303-664-P1, 

Illegible Date.

8 

Atomics International Drawing, Low Level Radioactive Waste Facility, Flow Diagram & Details, 303-664-

M1, Date Illegible. 

3 U.S. DOE Office of Environmental Audit, “Environmental Survey Preliminary Report, U.S. DOE Activities at SSFL,” 

1989. 

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3 

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According to the 1991 Technical Enforcement Report (TER) conducted by EPA, 1 a severe 

incident occurred at the RMHF in the 1960’s. Radioactive wastewater was released to the 

sanitary sewage RMHF Leach Field. The area was excavated and several feet of bedrock were 

removed, sealed with asphalt and backfilled with clean soil. In the late 1980’s when soil samples 

were collected from the area, radioactivity was 200 times background. The 1991 TER indicates 

that the RMHF may have been one of the two primary sources of groundwater contamination at 

SSFL (other than the SNAP reactor facility). 

There were several incidents when the RMHF Leach Field was flooded due to overflow from the 

flocculation system. In addition, according to a 1992 report on tritium release, the RMHF Leach 

Field became contaminated as a result of releasing approximately 5,000 gallons of water from 

the RMHF radioactive water system. It was assumed the major source of radioactive water at 

that time came from cleanup of the Building 4020 hot cells after examination of fuel assemblies 

from the first Sodium Reactor Experiment (SRE) core, or from wash water used at the SRE. 

According to the 1992 report, the RMHF Leach Field is considered a possible source of tritium 

in groundwater.2 

The RMHF was a disposal facility and staging area for waste treatment and final disposition. 

Waste and debris accumulated at the RMHF faster than it could be adequately prepared for 

shipment or disposal. This section provides a detailed account of building uses and processes, 

in addition to an internal Rockwell letter (c. 1975) that describes the RMHF’s physical condition 

and provides context in the day-to-day operations at hazards inherent to the workers of this 

facility. 

The RMHF resides on approximately three acres of land and includes a fenced area paved with 

asphalt, under which contamination reportedly exists due to a radiological incident that occurred 

in the early 1960’s. The paved areas of the RMHF are sloped to carry water away from the 

buildings to the perimeter channel, which drains west to a retention pond.3 

1 Environmental Protection Agency (EPA) Region IX Draft, Visual Site Inspection Report for Rockwell International 

Corporation, Rocketdyne Division, Santa Susana Field Laboratory Technical Enforcement Support at Hazardous 

Waste Sites, February, 1991 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 160, citation 2: 

.	 2 

Tuttle, R.J., Tritium Production and Release to Groundwater at SSFL, RI/RD92-186, Rockwell International,

December 1, 1992, pgs. 4-10–4-11. 

 EPA/HGL, Final HSA 7-3-NBZ, p. 13, citations 1-5:

1 

Interna Correspondence from Horton, P.H. to Nage , W.E., Rockwe Internat ona , Re: Request for Radioactive Maderail

[sic] and Radiation Producing Device User Authorization for RMDF Operations, December 13, 1989.

2 

The Boe ng Company, Radioactive Materials Handling Facility Decontamination and Decommissioning Engineering

Evaluation/Cost Analysis, June 18, 2007, p. A-1.

3 

Kartman, A.S., Aerial Photographic Analysis of Santa Susana Field Laboratory – Area IV, Ventura County , California, 

Volume 1 & 2, U.S. Environmental Protection Agency, Office of Research and Development, Environmental Sciences 

Division, March 2010.

4 

Carroll, J.W. et al., RMDF Leach Field Decontamination Final Report, ESG-DOE-13385, Rockwell International, 

September 15, 1982, p. 7.

5 

Correspondence from Gaylord, G.G., Rockwell International, to Liddle, R., U.S. Department of Energy, Re: Assessment 

Plan for the Radioactive Materials Disposal Facility (RMDF), September 14, 1990. 

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In general, radioactive wastes handled at the RMHF were residues from chemical and 

metallurgical laboratory operations, spent reactor fuel from decladding operations, maintenance 

work on contaminated equipment, and decontamination and decommissioning (D&D) of facilities 

in which nuclear operations were previously conducted. The RMHF received radioactive water 

from the Atomics International Hot Lab (4020), the SRE (4143), and any other DOE facilities that 

generated radioactive water as a part of its operations. 

11.1.1 RMHF Buildings / Facilities: 

4021 

4022 

4034 

4044 

4075 

*4563 

*4621 

*4622 

*4658 

Decontamination / Waste Packaging / Mixed Waste Treatment / Incinerator 

Mixed Waste Storage Vault 

Offices / Contaminated Laundry Storage

Radiation Safety

Low-Level Waste Storage 

Storage Shed

Source / Low-Level Waste Storage 

Radioactive Waste Counting Building 

Guard House 

*4663 

*4664 

Low-Level Waste Storage 

Low-Level Radioactive Waste Incinerator 

4665 

*4668 

Leach Field 

Supplies Storage

Hazardous Materials Storage / Sodium Cleaning Facility

Sanitary Sewer Septic Tank 

*Buildings/facilities and operations excluded from the 2006 Site Profile. More information on 

these locations provided in Section 12.0. 

Visitor Log Entry Codes associated with RMHF: 6017, 6000E, 6001R, RMDF, T022, T034 

11.1.2 Perspective: 1975 “Plan of Action” for the RMDF/RMHF 

The 1975 internal Rockwell document entitled, “Plan of Action for the RMDF”1 is among the 

original facility documents referenced by EPA in the Radiological Incident Reports database. 

The context of the document provides a detailed characterization of the RMHF’s appearance 

and condition, and establishes the precedent of waste handling and storage as documented by 

a Rockwell International employee. 

The document challenges assertions made in the 2006 SSFL Site Description about RMHF 

operations, general working environment, incidents associated with the facility, and the potential 

for worker exposure. It should be noted that although the “Plan of Action” was authored in 1975, 

it describes waste related to Atomics International programs that had ended years before, which 

had remained haphazardly stored at the RMHF in the holdup yard awaiting final disposition. The 

document depicts the work environment at the RMHF over the course of several years. The 

1 Internal Correspondence from J.M. Harris to W.R. McCurnin, Re: Plan of Action for RMDF, November 18, 1975 

(HDMSP00049486.pdf) 

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RHMF was described by DOE as being the greatest source of radioactive constituents at 

SSFL.1 

November 18, 1975

To: W.R. McCurnin From: J.M. Harris 

731-540 T020 731-540 T022 

Subject: Plan of Action for RMDF 

Upon taking responsibility for the RMDF in mid-July, a large number of 

contaminated and non-contaminated items existed at the RMDF complex 

requiring funds for disposition including areas of contaminated
 

asphalt and dirt. This plan documents the existence of these items, 

the extent to which areas were contaminated, the action taken to date, 

items requiring disposition, and the proposed plan of action. 

Status of RMDF Complex in mid-July “75” 

The holdup yard contained 250 drums (30 & 50 gal.), 59 high level 

pigs, 9 concrete conduits, 10 pallets of miscellaneous materials, 4 

transfer casks, 3 pieces of contaminated SRE moderator handling

equipment, 50 storage drums (30 ga.), 20 pallets of concrete blocks, 

and 32 drums for high level waste. There were several areas of 

contaminated asphalt within the holdup yard and 35 contaminated
 

pallets... 

Within the RMDF complex, external to all buildings, and outside the 

holdup yard, was an accumulation of boxes, barrels, conduits, liquid 

tanks and other miscellaneous materials including several large

rectangular metal containers, six trash dumpsters partially filled 

with vermiculite and a retort heater with an AEC property tag. There 

were several areas of contaminated asphalt and dirt at the west end of 

the complex, and a contaminated area extending along the fence west of 

the complex between Building 028 and the company’s permitter fence. 

Contained inside Building 021, within the holdup yard, were 18 small 

casks containing source materials, miscellaneous samples, a plastic

insert for a liquid barrel, remote handling extensions and other items 

wrapped in plastic. There were three contaminated areas on the floor 

and shelves of Bldg. 621. 

Building 075 contained stored SNAP equipment and packaged material 

awaiting shipment for burial. Also stored in this building were
 

pallets of contaminated materials from KEWB, ETB and Building 024 

waiting to be moved to the devon room of Bldg. 021 for
 

decontamination. 

Bldg. 022 contained barrels of sodium in temporary storage and
 

miscellaneous equipment used for handling R/A waste material. 

1 DOE Office of Environmental Audit, Environmental Survey Preliminary Report: DOE Activities at SSFL, 1989,

p. 4-31 - 4-33. 

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Areas of Contaminated Asphalt and Dirt: Some of the items listed above

were contaminated internally. At one time contaminated items were 

wrapped in plastic; however, due to long term storage (three years or 

more) and exposure to the weather, the integrity of the plastic was 

violated exposing the internal surfaces. It is presumed that the 

containers overfilled with rain water carrying contaminated materials 

onto the outer surfaces of the container, onto the pallets and onto 

the asphalt where, after repeated cycles (or perhaps only one good 

rainfall), the contamination spread over the areas shown in the
 

attached map of the RMDF complex. In the path of the spreading

contamination were pallets of non-contaminated materials and, at the 

west end of the holdup yard, there were rows of 30-gallon drums 

stacked on the asphalt. Most of the drums at the bottom of the stack 

and all of the pallets downstream of the source of contamination were 

contaminated. There were at least two separate major spills in the 

holdup yard, each covering 100 square feet or more, and several 

smaller spills on the order of 5 square feet or less. Thee remains an 

area to the south of the yard which has not yet been fully delineated 

due to the need for removing brush. 

At one time, pallets of barrels were stacked at the extreme west end 

of the complex outside the holdup yard and south of where bldg. 075 is 

located. It is presumed that one or more contaminated barrels filled 

with rain water and tipped over, or were full of contaminated water 

and tipped over, or by some other means spread contamination down the 

hill to the south for approximately 100 feet by several feet wide as 

shown on the attached maps. It is also presumed the other spills in 

that general area were caused by another incident. The full extent of 

this last spill has not yet been determined due to the need for 

removing brush along the outside of the perimeter fence and down the 

ditch to the west of the RMDF complex. 

Contamination within Bldg. 621 was confined to three areas. One
 

appeared to be a result of liquid seeping from the barrel liner stored 

on one of the upper shelves. The second appeared a result of a powdery 

substance on the floor and the third was of an unknown origin. 

There are still two other areas of concern, the pad shown as 644 on 

the map and the interim storage yard. Pad 664 contains the old 

evaporator, incinerator and oxidation equipment. Activity from the 

evaporator is sufficient to prevent detection of contamination near 

this area; however, preliminary indications show the ditch outside the 

fence is contaminated to the west of 664. 

The interim storage yard is between the RMDF and the SRE. Preliminary 

surveys of that yard indicate there are areas of contamination on the 

ground. This area must be cleared of brush and pallets of material 

must be moved to further delineate the spills. 

In summary, there are two well defined separate spills responsible for 

major efforts in the cleanup of the RMDF holdup yard. There are two 

separate spills, not fully defined, that require major efforts to 

cleanup the west end of the RMDF complex. There are two areas that 

preliminary R/A surveys show to be contaminated but require brush 

removal to further delineate. 

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Surveys of the asphalt and dirt in the RMDF complex were 35 mRad/hr or 

lower with the exception of limited areas at the start of the two 

major spills in the holdup yard. Details survey results are to be 

issued by HS and RS identifying all areas of detected contamination 

and the level of activity. 

Action Taken to Date: Since mid-July all of the contaminated and nonncontaminated items including the contaminated asphalt and dirt have 

been removed from the holdup yard. Some of the contaminated items were 

decontaminated and released for unrestricted use or disposed of as 

scrap. Contaminated pallets, barrels and other miscellaneous items 

were smashed and put into 19A boxes. Asphalt and dirt were
 

jackhammered out and put into 50 gallon drums. thirteen of the 59 high 

level pigs were contaminated and/or contained activated material so 

they were wrapped in plastic, placed on clean pallets and set aside in 

Building 75 for disposition. The concrete blocks were removed to the 

SRE for use as shielding, if needed, during the D&D program. Other 

non-contaminated items were placed outside the holdup yard within the 

RMDF complex for further disposition. Contaminated areas within Bldg. 

621 were decontaminated to as low as practical. 

The area of contaminated asphalt and dirt outside the holdup yard at 

the west end of the complex was jackhammered out and put into 50 

gallon drums. Brush was cleared along both sides of the fence down the 

hill to the west of the complex of Bldg. 028. A radiological survey of 

the area was made and the result are to be released by HS and RS. 

About 30 cu. ft. of dirt and rocks were removed from the upper part of 

the hill when I was directed by the program office to stop all 

activity and await new direction and funding. While working the upper 

part of the hill and just prior to receiving the above directive, the 

bottom of a buried conduit was exposed. 

The spill extending down the hill from the center of the complex at 

the west end required further removal of brush to fully delineate. 

Twenty-eight hundred cubic feet of R/A waste materials, generated by 

D&D of KEWB and ETB, were shipped to land burial in August removing 

some of the boxes, barrels, tanks and other miscellaneous materials 

from the RMDF complex. Items in Building. 021 that could be
 

decontaminated were returned to service or put into the salvage yard. 

Those items that could not be decontaminated were packaged for
 

disposal. All the sodium temporarily stored in Bldg. 022 and at the 

SRE was permanently stored in Bldg. 665. 

Items for Disposition: The following radioactively contaminated items 

remain at the RMDF complex for disposition. Funds will be available 

toward the end of the D&D program for the disposal of these items; 

however, it is recommended that funds be made available now to dispose 

of these potential sources of contamination. Items are listed in order 

of priority. 

1. Radioactive spills along fence by T028 and other spills outside the 

RMDF complex not yet fully identified. 

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2. Old evaporator system

3. Thirteen (13) high level pigs (contaminated or containing activated 

material
 

4. Conduit buried in the hill by T028

5. Several barrels in T021 contents unknown. 

6. Six (6) dumpsters containing vermiculite and unknown contents.

7. 20 19-A boxes of R/A material for disposal (partially funded)

50 50-gallon barrels of R/A asphalt for disposal

19 50 gallon barrels of KEWB and ETB R/A waste for disposal

6 50-gallon barrels of solidified R/A sludge from evaporator in 

Bldg. 021

8. Barrel of unidentified contents behind Bldg. 075.

9. R/A sources and other materials in Bldg. 621.

10.Heat exchanger from LAD (internally contaminated).

Retort heater AEC No. 165385 (internally contaminated).

Two liquid tanks ~ 200 and 300 gallon (internally contaminated).

11.Old cleaning tans and solar evaporators.

Incinerator system 

The following listed non-contaminated items require funds for
 

disposition. 

12. Small lathe and miscellaneous items on several pallets.

13. Scrap oxidation system by Bldg. 665. 

Plan of Action: If funded, items will be dispositioned as follows: All 

radioactive spills outside the fence of the RMDF complex (Item 1) will 

be picked up as they are identified. Delineation of these spills 

require brush to be removed from the permitter fence. 

The evaporator system (Item 2) will be decontaminated to as low as 

possible, dismantled and disposed of as R/A waste if necessary. The 13 

high level pigs (Item 3) will be moved to the devon room, Bldg. 021, 

and examined for content. Pigs that contain high level material will 

be packaged for burial and those requiring only decontamination will 

be decontaminated accordingly. Pigs that cannot be decontaminated will 

be disposed of as low-level waste. 

Remove the conduit buried in the hill by Bldg. 028 (Item 4).
 

Disposition to be determined by its contents. 

Barrels of unknown content in Bldg. 021 (Item 5) will be examined for 

content and disposed of accordingly. 

The six dumpsters (Item 6) will be surveyed for radioactive content 

and disposed of accordingly. The vermiculite will be use for
 

refurbishing shipping containers and the dumpsters when emptied could 

be used for transporting sealed bags of low-level compressible R/A 

waste. The items listed in 7 will be weighed, monitored and shipped to 

a disposal site for burial. The barrel of unidentified contents (Item 

8) will be examined for content and disposed of accordingly. 

The R/A sources located in Bldg. 621 (Item 9) will be transferred to 

[REDACTED] at DeSoto per his request. The other materials in Bldg. 621 

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will be dispositioned based on their potential use. Contaminated items 

listed as 10 under items for disposition will be packaged and shipped 

for burial. The cleaning tanks, solar evaporator and incinerator
 

systems (Item 11) will be decontaminated, dismantled and dispositioned 

accordingly. All non-contaminated items will be stored in appropriate 

locations. The scrap oxidation plant will be dismantled by a
 

contractor. 

Funding and schedule estimates will be made on request. 

Jim Harris

Remote Technology 

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Section 11.2 

Building Number:

Building Alias: 

Visitor Log Entry: 

4021 

021/T021

6017 / 6000E / 6001R / RMDF / T021 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Decontamination, Packaging, Radioactive Laundry 

Notes: Tritium use and air emissions. 

Radionuclides of Concern: Radioactive materials handled in Building 4021 were primarily in 

the form of items contaminated with mixed fission products and fuels.1 Radionuclides potentially 

present at the RMHF Building 4021 include: uranium fuel materials (U-234, U-235, U-238), 

transuranic isotopes of plutonium (Pu-238, Pu-239, Pu-240, Pu-241, Pu-242) and Am-241, 

fission products Cs-137, Sr-90, Kr-85, and Pm-147, thorium breeder material (Th-228, Th-232), 

and possible neutron activation products such as Co-60, europium isotopes (Eu-152, Eu-154), 

hydrogen-3 (H-3) [Tritium], iron-55 (Fe-55), nickel isotopes (Ni-59, Ni-63), managense-54 

(Mn-54), potassium-40 (K-40), and sodium-22 (Na-22). All radionuclides of concern listed with 

the exception of Fe-55, Kr-85, Mn-54, and Na-22 (due to relatively short half-lives) are included 

in the EPA October 2010 Field Sampling Plan for soil sampling in Area IV. However, for the 

purposes of dose reconstruction these radionuclides should be considered. 

11.2.1 Description of Operations & Processes 

Building 4021 was constructed in 1959. It was originally built to process low-level radioactive 

waste. After 1988, it was used as a processing area for wastes associated with D&D activities at 

the Energy Technology Engineering Center (ETEC). 2 Typical operations at Building 4021 

included the removal and concentration of radioactive particulate matter from liquid waste water 

generated by either the decontamination of equipment in Room 104 or the transfer of 

radioactive water from the other SSFL facilities to Building 4021. A radioactive waste 

evaporation system (RWES) provided capabilities to process liquid radioactive waste.3 Prior to 

1980, the radioactive liquid was stored in a 5,000-gallon aboveground holding tank outside 

Building 4021. Following an overflow of the 5,000-gallon holding tank in 1978 and 1980 (see 

radiological incident report A0070 and A0080), it was replaced with the 8,000-gallon 

underground radioactive liquid storage tank in Vault 2 of Building 4022. 

The natural gas-powered evaporator was taken offline following a 1966 fire and the flocculation 

system was used for a period, until the evaporator could be replaced with an electric-powered 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 52, citation 2:

2 

The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning Engineering

Evaluation/Cost Analysis, June 18, 2007, p. A-1. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 25, citation 4: 

4 The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning Engineering

Evaluation/Cost Analysis, June 18, 2007, p. A-1. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 20 

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evaporator. A 1989 report notes that the flocculation tower was built to pretreat radioactively 

contaminated water to make it easier to filter, but it was made unnecessary by better filters and 

became inactive.1 

Based on drawings and reports, the flocculation system may have existed from late 1964 to 

early 1981, but it appears not to have been used for the entire time period. Atomics International 

site waste management plans dated in the mid-1970’s note the use of the flocculation tower and 

January 1979 incident reports (Incidents A0077 and A0232) note there was an overflow of the 

flocculation tower, resulting in contamination.2 

Building 4021 had its own exhaust and filtration systems to control airborne radiation, arranged 

from areas of lower contamination to areas of high contamination, before being combined in an 

exhaust plenum with the air from Building 4022 and exhausted out the 130-foot high stack. 

According to Rockwell International documents, an emergency blower was not provided at 

Building 4021 because control of emissions during an emergency could be achieved by 

shutdown of operations. 3 A separate blower unit was provided for exhaust air from the 

evaporator and adjacent areas. 

The fenced area between Building 4021 and 4022 contained the blowers, filter banks, and 

compressors for both buildings as well as the shared exhaust stack, and three utility trenches 

used to support the ventilation ducts. Each trench contained a floor drain to convey surface 

water away from the area, and to the RMHF northern slope, bypassing the RMHF catch basin. 

A 1963 operating specification, presumed to be authored by Atomics International for the RMHF 

states that automatic air monitoring systems were installed in Buildings 4021 and 4022 to 

provide continuous air sampling of the decontamination and packaging rooms in Building 4021, 

and the vault areas below the floor of Building 4022. No information was found concerning air 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 16, citation 2:

2 Oldenkamp, R.D. and Mills, J.C., Nuclear Operations at Rockwell’s Santa Susana Field Laboratory – A 

Factual Perspective, N001ER000017, December 20, 1989, p. 29. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 16, citations 4,5,6,7:

4 Atomics International, Site Waste Management Plan, Circa 1972, p. 2.
 

5 Atomics International, Management of AEC-Generated Radioactive Wastes at Atomics International, May 

31, 1974, pgs. 1-2.

6 Gardner, F.W. Rockwell International Internal Correspondence, Re: Release of Radioactivity from RMDF – 

January 1979, February 23, 1979.

7 Gardner, F.W., Rockwell International Internal Letter, Subject: Radiological Safety Incident Report RMDF, 

February 26, 1979. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 17, citations 5,6,1: 

1 Chapman, J.A., Radioactive Materials Disposal Facility Safety Analysis Report, Rockwell International, 

June 16, 1986, p. 63.

5 U.S. Department of Energy Environment, Safety and Health Office of Environmental Audit, Environmental 

Survey Preliminary Report, DOE Activities at Santa Susana Field Laboratories, Ventura California, February

1989, p. 3- 13.

6 Chapman, J.A., Radioactive Materials Disposal Facility Safety Analysis Report, Rockwell International, 

June 16, 1986, p. 56. 

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sampling procedures prior to 1963. The 1963 RMHF operating specification also notes that an 

automatic continuous stack effluent monitoring system was installed to sample and detect any 

radioactive, gaseous, or particulate releases to the atmosphere. Any abnormally high 

radioactive release would cause a buzzer and red light to actuate and remain in the alarm 

position until actions to remedy the cause and reduce the count rate were accomplished or until 

the alarm was manually deactivated.1

According to an Atomics International disposal document, properly identified and monitored 

radioactive dry wastes were collected from generator pickup stations by RMHF personnel and 

transported to RMHF on an established routine schedule, or when requested.2 The EPA HSA 

document explains, in detail, how liquid and dry radioactive wastes were processed. A number 

of internal Rockwell letters were provided to EPA, and depicted annual summaries of waste 

volumes processed and packaged at Building 4021. 

A 1989 EPA report indicated the air emissions from Buildings 4021 and 4022 consisted primarily 

of surface radioactive particles resulting from decontamination processing, packaging activities 

in Building 4021, and from store and handling activities in Building 4022. The particulate matter 

contained uranium, plutonium; cesium-137 (Cs-137), strontium-90 (Sr-90), krypton-85 (Kr-85) 

and promethium-147 (Pm-147) as mixed fission products; and cobalt-60 (Co-60) and 

europium-152 (Eu-152) as activation products. The particulate matter in air was controlled 

through filtration by HEPA filters. The ambient air within Area IV was reportedly monitored daily 

by continuous collection of air particulate samples using a network of eight air samplers. The 

samples were counted for alpha and beta radiation following a 120-hour delay to allow for radon 

and thorium decay. DOE monitored the stacks serving Buildings 4021 and 4022. The DOE 

survey report indicated that emissions from the RMHF Buildings 4021 and 4022 were higher 

than the equivalent ambient air emissions shown in the Rockwell International annual 

monitoring report data for 1981-1987. 

11.2.2 1990 Air Emissions from RMHF

According to Rockwell International 

Radionuclide Quantity (Ci)

Americium-241 (Am-241) 1.0 x 10

Cesium-137 (Cs-137) 3.3 x 10

Cobalt-60 (Co-60) 4.9 x 10

Tritium (H-3) 1.6 x 10

Plutonium-239 (Pu-239) 2.5 x 10

Strontium-90 (Sr-90) 2.9 x 10

1 EPA/HGL, Final HSA 7-3-NBZ, p. 18, citation 7:

7 Operating Specification Radioactive Materials Disposal Unit, November 7, 1963, p. 4. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 22, citation 7: 

7 

Unknown Author, Disposal of Radioactive Materials at Atomics International, Unknown Date, p. 8. 

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According to a 1991 DOE Tiger Team finding, stack emissions had not undergone formal 

evaluation in accordance with established National Emissions Standards for Hazardous Air 

Pollutants (NESHAP) regulations. Additionally, although the samplers at the RMHF had the 1

required continuous radiation monitors to detect sudden increases in radiation during accident 

situations, deficiencies in the radionuclide particulate sampling systems (which were not in use 

until 1970) prevented the samplers from meeting NESHAP requirements. 

Examples of noted stack sampling deficiencies included suitability of sampling locations, 

number of sampling points, and sampler design. Moreover, the HSA indicates that the samplers 

at the RMHF were not originally designed to monitor the large range of particulates, which could 

be present as a result of HEPA filter problems, deposition inside the stack, or corrosion buildup 

in the stack. Due to design flaws, the radioactive particulate emission release rates that were 

supplied for modeling purposes by Rockwell International were determined from air filter 

samples, which were collected in a manner that may not have been representative of actual 

emissions. DOE’s Tiger Team observed that the changing and handling of the filter samples 

revealed that a loss of particulate matter may have occurred, and thus air emission modeling 

may not have been accurate.2 

According to DOE, “The principal source of potential radiation dose to the public from SSFL is 

the RMDF.” A representative of DOE also stated that, “Airborne dose assessment at this facility 

[RMHF] may be imprecise ... because of AIRDOS computer monitoring difficulties.”3 

In 1989, DOE Office of Environmental Audit indicated that air sampling and monitoring practices 

at SSFL were insufficient for the duration of site operations. Given documented onsite burning of 

radioactive waste and the presence of the RMHF incinerator, current estimations of airborne 

activity and emissions as they relate to internal and environmental ambient exposure may be 

inaccurate and even grossly underestimated, and therefore inappropriate for use in dose 

reconstruction. 

EPA’s HSA report provides detailed information on various surveys for radioactivity associated 

with the RMHF. For instance, in 1987 the Monthly Comprehensive Smear Surveys indicated that 

quarterly reviews noted high radiation levels measured by location film badges in the 

decontamination and packaging rooms, although gradual reductions had reportedly occurred in 

past quarterly surveys. 

In 1988, DOE indicated that even though improvement had been made to reduce radiation 

exposure rates, because of changing operations involving radioactive materials handling at the 

RMHF, exposure rates may have exceeded the DOE guideline of 100 millirems per year for 

continuous exposure at the Area IV property boundary north of the RMHF. The guideline was 

intended to prevent members of the public from unknowingly receiving excessive exposure as a 

result of DOE operations. According to the report, long-term exposure to a member of the public 

1 DOE, Tiger Team Assessment, Energy Technology Engineering Center, April, 1991 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 28, citation 1:

1 U.S. Department of Energy, Tiger Team Assessment Energy Technology Engineering Center, DOE/ 

EH-0175, April 1991, pgs. 3-12–3-15. 

3 DOE Office of Environmental Audit, Environmental Survey Preliminary Report: DOE Activities at SSFL, 1989,

p. 4-31 - 4-33. 

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was unlikely due to the rugged terrain along the north boundary and daily security patrols. 

However, no reference was made to concerns about potential worker exposure.1 

In 2007, Boeing performed radiological surveys on Buildings within the RMHF complex, 

including Building 4022. The original survey for Building 4022 called for multiple survey units but 

because of relatively high levels of radiation (up to 140 micro roentgens per hour) primarily on 

the west side of the building and likely due to radiation levels from Building 4021, Boeing 

decided that the building could not be satisfactorily surveyed as originally proposed. 

The preliminary MARSSIM classification for the Building 4021 is Class 1, due to its former use, 

radiological incidents, and previous radiological investigations. As of 2011, Building 4021 was 

still standing but no longer in use. According to Boeing and the EPA HSA, the inside floor and 

walls were contaminated and under fixative paint, the upstream HEPA ventilation ducting in 

Building 4021 and between Building 4021 and 4022 was highly contaminated, and areas under 

the floor drain lines were highly contaminated.2 

11.2.3 Building 4021 Radiological Incident Reports 

There have been several incidents associated with Building 4021 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

11.2.4 Building 4021 Radiological Incident Report Summary, Data Provided by Boeing 

Incident 

File Name

Date of Incident Location of 

Incident

Isotopes Description of Incident

A0348 3/16/60 Decon RM Mixed Fission A Boiling Decon Solution Foamed Over and Spread Contamination in 

Decon Rm

A0336 12/21/60 Decon RM Glove Box and Be Contaminated Cutoff Wheel Separation Caused 

Floor Contamination

A0017 6/12/62 RMDF Decon 

Rm

Mixed Fission SRE Sodium Pipe for Decon Exhibited High Dose Rates and Airborne 

Activity

A0019 10/30/62 Decon Rm T021 Mixed Fission Employee Failed to Wear Prescribed Respiratory Equipment

A0408 6/10/64 Bdg 21 Decon Mixed Fission RMDF Decon RM Moderator Can Cleaning Resulted in Sodium Fire 

and Explosion

A0448 12/22/64 North of T021 Mixed Fission Heat Exchanger Bell Leaked Water onto Ground While Being Moved

A0297 11/11/66 T021 Evaporator Mixed Fission Filters Plugged and Collapsed When “Water” Evaporator Pan Caught 

Fire

A0387 6/26/68 Decon Rm Fire Started During Primary Sodium Draining from a Hallam Throttle 

Valve

A0388 10/31/78 Evaporator RM RMDF Evaporator Fire in Building 4021

1 EPA/HGL, Final HSA 7-3-NBZ, p. 46, citation 2:

2 Weiner, L.A. and Barisas, S., Status Report of the DOE Activities at the Santa Susana Field Laboratories 

Site Environmental Survey, U.S. Department of Energy, June 30, 1988, p. 10. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 43 

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A0070 5/22/78 R/A Tank Area Mixed Fission Sump Pump Failure Allowed R/A Liquid to Overflow into Yard

A0558 10/13/78 R/A Decon Rm Employee Contaminated Previous Cut

A0559 10/10/79 RMDF Decon 

Rm

Mixed Fission Splashed R/A Water Contaminated Employee and Subsequently His 

Film Badge

A0080 1/9/80 RMDF & 

Drainage

Mixed Fission Water Hose Broke Overfilling R/A Waste Tank and Drained to 28 Pond

A0243 2/29/80 T021 Sump 

Pump

Employee Became Contaminated Working on R/A Sump Pump

A0566 12/4/81 RMDF Decon 

Rm

Mixed Fission Bucket Overfilled with R/A Liquid Spilling into Drum

A0180 11/9/1987 Decon RM 021 Mixed Fission Waste Packaging Generated High Airborne Activity When tem Was 

Dropped

A0179 11/18/87 Decon Rm 021 Mixed Fission Escorted Worker Entered Contaminated Area Before He Could Be 

Stopped

A0208 8/9/90 Decon RM 021 Mixed Fission Contaminated Power Tool Cut Finger During Size Reducing R/A Waste

A02011 11/15/90 Decon Rm 021 Ce 144 Exit Survey Revealed Contamination on Employee Shorts

A0266 2/14/91 Exhaust Fan R/A Exhaust Blower Failure Caused Alarm

A0580 8/6/93 RMDF Laundry 

RM

Employee Worked in Controlled Area Without Permission

A0655 10/12/94 RMDF T021 PK 

RM

Cs 137 Cs 137 & Co 60 Contamination of Personnel Shoes

A0680 10/3/97 Lower Park Lot State Found 4 Blocks Over the Limit 100 120 Cpm Fixed Contamination

A0671 1/10/97 DeCon Room Fixed R/A Contamination on ndividual s Personal Shoe

Incident 

File Name

Date of Incident Location of 

Incident

Isotopes Description of Incident

• On June 10, 1964 personnel were attempting to disassemble a moderator can for disposal. 

There was a reaction of the sodium that caused an explosion and fire. Contaminated smoke 

resulted in an average beta gamma contamination level of 4.5 x 106 dpm / 100 cm2 . The 

incident report indicated that sodium sparks were blown approximately 25 feet, searing plastic 

on the floor and red-line clothing worn by personnel. The fire was extinguished with calcium 

carbonate. Personnel received contamination in the nose (60 dpm beta gamma) as a result of 

opening the door of the decontamination room immediately after the explosion and inhaling 

the smoke. After the smoke had cleared from the room, personnel drove a fork lift into the 

room to push the moderator can back into the storage can. The incident report indicated that 

the moderator can covered with sodium had a radiation level on a section of the can of 50 rad/

hr, including mixed fission products. The incident report did not provide additional information 

to indicate whether there were any releases of contaminated air to the atmosphere, whether 

the surfaces in the decontamination room became contaminated and required 

decontamination, or how the moderator can was handled after being placed in the storage can 

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(A0408).1

• On July 21, 1964 while washing out an empty moderator storage can, a sodium reaction 

caused an explosion and fire. The explosion and fire were reported to have been confined to 

the inside of the storage can and were smothered with calcium carbonate. Continuous air 

monitoring reportedly indicated no increase in airborne contamination during the incident. The 

incident report stated that an air sample taken 75 minutes following the incident indicated 

airborne concentrations of 3.2 x 10 11 μC/cc alpha and 1.2 x 10-10 μC/cc beta gamma. Nose 

swipes of the personnel involved indicated no detectable contamination (A0413).2

• On December 22, 1964 a component from the SRE was taken to Building 4021 for 

decontamination. The incident report and the Boeing Incidents Database identified the 

component as a “heat exchanger bell.” According to the incident report, while transferring the 

component from a storage area to the decontamination room, rain water contained under 

plastic covering spilled on the pad and asphalt in back of the decontamination room resulting 

in the contamination of the pad, asphalt, forklift, and a pallet. The maximum contamination 

measured was 7,000 dpm / 100 cm2 with a radiation level of 6 mrad/hr from the asphalt. A

2005 historical site assessment indicated a maximum contamination level of 2000 dpm / 100 

cm2 for this incident, and it is unclear from the source document whether the level of 

contamination was 2,000 dpm / 100 cm2 or 7,000 dpm / 100 cm2. At the time of the incident 

report, the asphalt and the pad were being cleaned “without significant progress.” The incident 

report did not elaborate on how the area was being cleaned or if the area had been 

successfully cleaned. The forklift was cleaned to 30 dpm / 100 cm2 and the pallet was 

discarded as contaminated waste, although the incident report did not provide information to 

indicate how or to where the pallet was discarded.

Investigation of the incident had determined that the plastic covering for the SRE component 

was contaminated to a level of 2,000 dpm. In addition, the single plastic covering had a 

number of rips and tears from which the contamination was spread by the entrance of rain 

water. The incident report noted that it was unknown whether or not the item was in the 

reported condition upon receipt to the RMHF or if it was damaged at the RMHF. It could not be 

determined from the incident report where the item was stored prior to decontamination as the 

incident report identifies the area only as a “storage area.” Given that rain water had infiltrated 

the plastic covering of the SRE component, it is assumed this storage area was located 

exterior to Building 4021 (A0448).3

• On November 11, 1966 a water evaporator pan caught fire. Upon arrival to the building, 

smoke was observed to be emanating from Building 4021 through all the openings, including 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 37, citation 1:

1 Young, L.N., Atomics International Internal Correspondence, Re: Incident Report Building 21-

Decontamination Room, June 6, 1964, June 17, 1964. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 37, citation 2: 

2 Young, L.N., Atomics International Internal Correspondence, Re: Incident Report, Building 021,

Decontamination Room, July 21, 1964, July 22, 1964. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 39, citation 1:

1 Young, L.N., Atomics International Internal Correspondence, Re: Incident Report, North Side – Building 

021, December 22, 1964, January 18, 1965. 

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doors, ventilators, and stack. A portable air sampler was placed in the smoke just outside 

building 4021, and the Rocketdyne weather station was contacted to obtain wind direction 

information. This information was not available, however, because the Rocketdyne 

anemometer was inoperative at the time. Visual inspection observed the smoke to be moving 

in a southeasterly direction and it was recommended that ventilation be turned off in Buildings 

4036 and 4037. [There is no mention in the incident report that employees working outside at 

neighboring areas were warned of the risks associated with this incident]. Rocketdyne 

personnel monitored the smoke, office areas and building walls with a GM portable survey 

instrument and detected no radiation levels above “normal background,” as reported by 

Rocketdyne.

The fire was contained approximately 10 minutes following the initial alarm. A fireman entered 

the attic of Building 4021 to inspect ductwork of the building and reported no unusual 

conditions except for residual smoke. An employee reported to the scene to perform radiation 

and contamination surveys of all equipment and trucks in the area.

Investigation of the incident determined that the fire occurred in the east evaporator in the 

packaging room of Building 4021. The fire was caused by the “apparent ignition by radiant 

heaters of wax floating on the liquid undergoing evaporation.” All firemen were checked for 

contamination and it was determined that contamination was limited to shoes at levels of up to 

1 mrad/hr. These were decontaminated and returned. Some personnel had hand 

contamination of up to .54 mrad/hr, which was successfully decontaminated to less than 30 

dpm / 100 cm2 betta gamma. Bioassay samples indicated no evidence of inhalation of 

radioactive material. Environmental sampling, including air sampling showed concentrations of 

6x10 10 μCi/cc betta gamma. Fire trucks, “MSA equipment,” fire bottles, and office areas were 

also surveyed and had measurements ranging from less than 50 dpm / 100 cm2 to less than 

500 dpm / 100 cm2 beta gamma.

The incident report concluded that all areas, vehicles, and equipment were successfully 

decontaminated to less than 30 dpm / 100 cm2 beta gamma. The report also summarized that 

no internal or external personnel exposure resulted and “no significant release of radioactive 

materials occurred” (A0297).1

• On June 26, 1968 a sodium fire occurred at the RMHF. According to the incident report, a 

Hallam reactor primary system throttle sodium valve (Valve #303) was being em ptied of 

sodium in front of the south roll-up door of Room 105 of Building 4021, when the sodium 

caught fire. The heating and draining process consistent of wrapping the valve with strip 

heaters, heating the valve under controlled conditions, and allowing the liquid sodium to drain 

into a pan containing calcium carbonate.

During the course of draining Valve #303, small fires occurred as the sodium drained into the 

pan. Two technicians extinguished the fires with calcium carbonate; however, the small fires 

merged into one large fire in the pan and could not be controlled with calcium carbonate. The 

volume of the sodium that had been drained into the pan was estimated to be approximately 

15 gallons. Control of the fire was achieved with an Ansul Net-L-X extinguisher and calcium 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 39, citation 2:

2 Clow, H.E., Atomics International Internal Correspondence, Subject: Fire Alarm, Bldg. 021, November 15, 

1966. 

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carbonate. Upon receiving “HSRS” approval, staff removed the pan containing the sodium and 

fire-fighting chemicals to the “dangerous chemicals disposal pit for burning.”1

The incident report indicated that two high-volume air samples were taken during the fire and 

included a five-minute sample and a 30-minute sample. Both samples were placed in the path 

of the smoke cloud. According to Atomics International, each sample showed a concentration 

of approximately 2.6x10 10 μCi/cc. The samples were recounted after 48 hours and were 

reported to have a count rate of less than background. The regulatory limit for occupational 

exposure in 1968 was reported to be 9x10 9 μCi/cc.2

Additionally, the ventilation systems in Buildings 4036/4037 and 4027 were shut down since 

the smoke cloud was moving in the direction of those buildings. Nasal smears and clothing 

surveys were made of all personnel involved in controlling the fire. In addition, smear and 

radiation surveys were made of the areas located in the path of the smoke cloud and of the 

general fire area. The incident report reported the measurements to be at background. The 

regulatory limit for insoluable sodium-22 was reported to be 3x10 10 μCi/cc at the time of the 

incident (A0387).3

• A fire occurred in the evaporator area of Building 4021 on October 31, 1974. The evaporator 

system in the building consisted of an angle iron frame, a stainless steel tank, a tank cover, 

and two independently controlled gas burners. Limited information is available regarding the 

actual fire; however, a summary of the damage indicated that the fire damaged the high 

voltage ignition wire on each burner and the flame detectors. In addition, the aluminum 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 40, citations 2-4:

2 Mooers, A.R., Atomics International Internal Correspondence, Subject: Sodium Fire at the RMDF, July 10,

1968. 

3 Tschaeche, A.N., Atomics International Internal Correspondence, Subject: Sodium Fire at SS-021 on June 

26, 1968, July 2, 1968.

4 Corning, F.E., Memorandum, Sodium Fire, Building #022, Santa Susana, June 26, 1968. The building 

number is incorrectly identified on this June 26, 1968 memorandum subject line. The content of the 

memorandum indicates the fire occurred on the south side of Building 4021. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 40, citations 5-7:

5 Mooers, A.R., Atomics International Internal Correspondence, Subject: Sodium Fire at the RMDF, July 10,

1968.

6 Tschaeche, A.N., Atomics International Internal Correspondence, Subject: Sodium Fire at SS-021 on June

26, 1968, July 2, 1968.

7 Corning, F.E., Memorandum, Sodium Fire, Building #022, Santa Susana, June 26, 1968. The building 

number is incorrectly identified on this June 26, 1968 memorandum subject line. The content of the 

memorandum indicates the fire occurred on the south side of Building 4021. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 40, citations 8-10:

8 Mooers, A.R., Atomics International Internal Correspondence, Subject: Sodium Fire at the RMDF, July 10,

1968.

9 Tschaeche, A.N., Atomics International Internal Correspondence, Subject: Sodium Fire at SS-021 on June

26, 1968, July 2, 1968.

10 Corning, F.E., Memorandum, Sodium Fire, Building #022, Santa Susana, June 26, 1968. The building 

number is incorrectly identified on this June 26, 1968 memorandum subject line. The content of the 

memorandum indicates the fire occurred on the south side of Building 4021. 

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removable section of the tank cover melted and the ignition transformer on one burner was 

damaged by the water that was used to hose down the area following the fire. Additional 

information regarding the type of cause of the fire could not be located in available documents 

reviewed by the EPA research team (A0388). 1

• On May 22, 1978 a sump pump at Building 4021 stopped working and contaminated liquid 

flowed out of a 5,000-gallon radioactive liquid holdup tank. Approximately 40 to 50 gallons of 

water contaminated with mixed fission products ran through a break in the asphalt dike 

contaminating asphalt and eight tires from a trailer parked on the asphalt. The contaminated 

water ran about 40 yards west and south of the dike and ranged from 2 feet to 10 feet in 

width. The wet area was covered with Pell-A-Cell absorbent, which was later picked up with 

shovels and put into 55-gallon drums. Foam was also applied and vacuumed up for further 

cleanup.

Smear samples were taken of the foamed-and-vacuumed area the following day. Five smears 

showed contamination ranging from 108 dpm/100 cm2 to 356 dpm/100 cm2. These smears 

were all within an area of 15 feet from the holdup tank enclosure fence. This area was foamed 

and vacuumed again. Fifteen additional smears were taken of this area and the results were 

all less than 100 dpm / 100 cm2.

As indicated above, the tires of a trailer parked in the contaminated area also became 

contaminated. Contamination ranged from 59 dpm/100 cm2 to 601 dpm/100 cm2. The tires 

were reported to have been cleaned to less than 50 dpm/100 cm2 and the trailer was removed 

from the area. The trailer was reported to have belonged to an “outside contractor.”

An additional survey was performed using a PUG with a shielded Eberline HP-210 probe, and 

direct readings observed by the instrument ranged from 20,000 cpm to 45,000 cpm.2 

Background readings in this area were approximately 4,000 cpm. As a result of the high 

readings, on May 30 1978 [over a week after the original incident] the contaminated areas 

were foamed with “Ammoniated Turco Meteor All Purpose Water Emulsion Cleaner” and 

scrubbed with a stiff bristle brush and vacuumed. Reading with the PUG and HP-210 probe 

showed no significant change, and on June 1, 1978 the contaminated area was covered with 

“Midwest’s Z-40 Floor and Road Seal Emulsion” to fix any possible loose contamination. The 

incident report did not indicate any additional surveys and did not provide any figures to show 

the extent of contamination and the location of the trailer. Apparently, the incident report did 

not mention the potential personnel contamination of those involved in the incident of 

subsequent cleanup efforts (A0070). 3

1 EPA/HGL, Final HSA 7-3-NBZ, p. 41, citation 1:

1 Gutierrez, E.P., Rockwell International Internal Letter, Subject: Evaporator Fire Incident, Bldg. 021, 

November 8, 1974. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 41, citation 2:

2 A PUG is a radiation survey instrument made by Technical Associates, which has been developing 

radiation detection equipment since 1946. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 41, citation 3:

3 Bradbury, S.M, Rockwell International Internal Correspondence, Re: Sump Overflow in the 5,000-Gal 

Holdup Tank Enclosure at the RMDF, June 7, 1978. 

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A follow up to the incident noted that during the “Great Rain of ’78” there was great difficulty in 

handling the amounts of contaminated water generated at several facilities. The 5,000-gallon 

radioactive liquid storage tank was surrounded by a security fence and low berm, but was 

unprotected from the weather. Rain poured off the building 4021 roof and fell directly into the 

holdup sump and was automatically pumped to the holdup tank.

To prevent rainwater from filling the holdup tank, the berm was broken to allow the water to 

drain out. However, because the berm and sump pit were the only protection against release 

of water from overflow and leakage, a work order to repair the berm was issued soon after the 

end of the rains. The work order was not completed in time to prevent the incident. The berm 

and sump pit, when empty, provided retention capacity for about 2,000 to 3,000 gallons of 

water. The 5,000-gallon holdup tank usually contained 3,000 to 5,000 gallons of water. Thus, 

the area did not provide enough protection against total release of the holdup contents. The 

radiation exposure rate at the fence around the tank averaged 50 mR/hr, but because of the 

large size of the tank, radiation exposure extends “significantly” beyond the enclosure. The 

incident followup letter notees that these problems and potential hazards could be eliminated 

by construction of a concrete block wall enclosing the tank with a weather-tight roof. It also 

states that there have been ‘many bad experiences with contaminated water at Santa 

Susana,” and the facility must be upgraded to prevent future incidents. [At the time this letter 

was generated, SSFL had operated for 30 years; the RMHF Building 4021 had operated for 

nearly 20 years].1

• On January 9, 1980 a water hose broke and caused the 5,000-gallon radioactive liquid 

storage tank to overflow at Building 4021 and drain to the RMHF 4614 Holdup Pond. The leak 

was discovered by RMHF personnel located outside the RMHF perimter fence who noticed 

liquid running out the top of the storage tank. The source was found to be coming from the 

decontamination room of Building 4021 and the water was shut off. According to the incident 

report, the excess water was a result of an employee neglecting to turn off the water hose at 

the valve.

The hose was used with a spray nozzle for rinsing during decontamination operations in the 

decontamination room of Building 4021. The incident report stated that although the nozzle 

was shut off, the excess pressure caused the hose to break, causing the tank to fill and 

overflow. This incident resulted in the release of approximately 100 gallons of liquid containing 

.01 μCi of mixed fission products and caused the contamination of approximately 2,500 

square feet of asphalt paving and an accumulation of about 3,000 gallons of contaminated 

liquid with 100 gallons being released. According to the incident report, the loss occurred 

because the asphalt berm that should have contained the contaminated spill leaked and 

permitted the contaminated liquid to spread over a sizable area.

Heavy rain prior to the spill and during the cleanup operations resulted in a significant 

increase in the decontamination effort. Runoff from the rainfall contributed to a major increase 

in the amount of contaminated water that was collected first in the RMHF “radioactive pond” 

and then transferred to holdup tanks in the RMHF. The incident report indicated that the tank 

that overflowed and released the contamination liquid was taken out of service and that all 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 42, citation 1:

1 Internal Correspondence from Tuttle, R.J. to Breese, L.S. and Walter, J.H., Rockwell International, Re: 

RMDF Liquid R/A Waste Holdup Tank, July 27, 1978. 

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future radioactive water was being stored in an underground tank with complete retention of 

any overflow or spill. The incident report did not provide information to indicate how the 

affected areas were cleaned or to what levels they were cleaned (A0080).1 

• In 1986, during the third week of the third quarter electric power to the HEPA filtered exhaust 

blower for the decontamination and packaging rooms was inadvertently left off after a test and 

the exhaust system was inoperative for an unknown period of time. Air sampling data showed 

that radioactive contamination migrated across the decontamination room and into the 

packaging room.2

• On October 3, 1997 four concrete blocks in the lower RMHF parking lot were discovered to 

have small areas, less than 1 foot square, of fixed contamination ranging from 100 to 800 

cpm. The concrete blocks were decontaminated, resurveyed, and released without 

radiological restrictions (A0680).3

As of July 27, 2011 Boeing notes that there is no storage of radioactive waste in Building 4021. 

However, the inside floor and walls are contaminated and under fixative paint, the upstream 

HEPA ventilation ducting in Building 4021 and between Building 4021 and Building 4022 is 

highly contaminated, and areas under the floor drain lines are highly contaminated. 

The EPA provides a detailed chronology of radiological investigations that are associated with 

each building and facility. However, given the nature and severity of RMHF contamination, 

NIOSH and ABRWH are encouraged to review the EPA HSA 7-1 NBZ document pertaining to 

the Area IV Study in its entirety, with particular attention on the RMHF given consistently high 

radiation measurements throughout the buildings and facilities associated with the RMHF, 

Rockwell International’s definitions of “acceptable” readings, and clear indications of worker 

exposure risk. 

EPA found it to be significant that a 1993 internal letter between Rockwell staff notes that there 

is suspected significant contamination beneath the asphalt-covered RMHF area. This letter 

states that such contamination was not included in a site characterization plan because the 

complex will be remediated after the RMHF ceases operation. However, as noted, in 2011 that 

had yet to occur. 4

1 EPA/HGL, Final HSA 7-3-NBZ, p. 42, citation 2:

2 Bradbury, S.M., Rockwell International Internal Correspondence, Re: Radiological Safety Incident Report,

RMDF, January 9, 1980, April 9, 1980. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 43, citation 1:

1 Internal Correspondence from Moore, J.D. to Remley, M.E., Rockwell International, Re: Quarterly Review 

of the RMDF for Radiation Safety, Third Calendar Quarter, 1986, December 15, 1986. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 43, citation 2:

2 Deschamps, R., The Boeing Company Internal Correspondence, Re: Incident Report No. A0680, October 20, 1997. 

4 EPA/HGL, Final HSA 7-3-NBZ, p. 47, citation 5:

5 Internal Correspondence from Author, D.W., to Hoffman, N.J., Re: SSFL Area IV Site Characterization 

Plan, June 17, 1993. 

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Section 11.3 

Building Number:

Building Alias: 

Visitor Log Entry: 

4022 

022/T022

6017 / 6000E / 6001R / RMDF / T022 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Vault Storage / Waste Compactor 

Note: Tritium use and air emissions. 

Radionuclides of Concern: Radioactive materials handled in Building 4022 were primarily in 

the form of fuels. Radionuclides potentially present at the RMHF include: uranium fuel materials 

(U-234, U-235, U-238), transuranic isotopes of plutonium (Pu-238, Pu-239, Pu-240, Pu-241, 

Pu-242) and americium-241 (Am-241), fission products Cs-137 and Sr-90, thorium breeder 

material (Th-228, Th-232), and possible neutron activation products such as cobalt-60 (Co-60), 

europium isotopes (Eu-152, Eu-154), hydrogen-3 (H-3), iron-55 (Fe-55), nickel isotopes (Ni-59, 

Ni-63), managense-54 (Mn-54), potassium-40 (K-40), and sodium-22 (Na-22). All radionuclides 

of concern listed with the exception of Fe-55, Mn-54, and Na-22 (due to relatively short halflives) are included in the EPA 2010 Field Sampling Plan for soil sampling in Area IV.1 However, 

for the purposes of dose reconstruction, these radionuclides should be considered. 

11.3.1 Description of Operations & Processes 

Building 4022 was constructed in 1959 as part of the RMHF complex, which was a support 

facility to SNAP, the SRE, and the Hallam Nuclear Power Facility (storage, volume reduction, 

packaging, shipping of SNAP / SRE radioactive waste). Building 4022 was used for the dry 

storage of used and unused nuclear fuel. During ETEC operations, the building vaults were 

used to store containerized wastes from decontamination and decommissioning activities. 

There were seven below-grade storage vaults constructed with reinforced concrete and aircooling systems, which ranged in size and were used for the storage of mixed waste and lowlevel radioactive waste.2 By 1980, an 8,000-gallon radioactive liquid storage tank was located in 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 76, citation 3:

3. The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning

Engineering Evaluation/Cost Analysis, June 18, 2007, p. A-2. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 56, citations 9, 10, 11, 12: 

9 The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning 

Engineering Evaluation/Cost Analysis, June 18, 2007, p. A-2.

10 Internal Correspondence from Seward, F.A. to Nagel, W.E., Rockwell International, Re: Request for

Radioactive Material and Radiation Producing Device User Authorization for RMDF Operation, December 

10, 1985. 

11 

Haley and Aldrich, Inc., Closure Plan for the Radioactive Materials Handling Facility (RMHF), October 2006, p.

3-2.
 12 

Daniel, Mann, Johnson & Mendenhall, Calculations for R/A Waste & Fuel Storage Facility, Santa Susana, 

California, AEC Contract AT(11-1)-632, August 1958.
 

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Vault 2 of Building 4022, replacing the aboveground tank outside Building 4021 following 

overflow of that tank (please see Building 4021 Radiological Incident Report). 

Storage of the SRE Core I at Building 4022 began in 1959 and SRE Core II was stored 

beginning in 1964. Both cores were stored in 3.5-inch diameter by 166-inch long canisters with 

threaded closures. Core I contained 2.8 percent uranium-235 (U-235). Core II had different 

assemblies with approximately 92-93 percent U-235 enrichment. 1 The EPA-HSA contains a 

comprehensive chronology of fuel stored in the RMHF vault.2 

In addition to the 8,000-gallon radioactive liquid storage tank in Vault 2, two 5,000-gallon 

storage tanks were located in the vault. According to EPA’s discussions with Boeing personnel, 

these two tanks were secondary tanks brought from the SRE (Building 4143) to collect water, 

primarily rain water, during the RMHF Leach Field cleanup.3 

Building 4022 had its own exhaust and filtration systems to control airborne radiation. The 

building had no windows. During fuel transfer operations, all of the doors were closed. The 

ventilation for the above-grade working area of Building 4022 was brought about by natural 

convection from the work floor to the vents on the roof. In the event of a blower failure, an 

emergency diesel generator provided power for the emergency backup blower, with a filter 

bank. Each filter bank had a set of 10 pre filters and 10 HEPA filters. After filtering, air from 

Building 4022 was combined in an exhaust plenum with air from Building 4021, and exhausted 

out the 3-foot-in-diameter 130-foot high stack. 

A 1989 EPA report indicated the air emissions from Buildings 4021 and 4022 consisted primarily 

of surface radioactive particles resulting from decontamination processing, packaging activities 

in Building 4021, and from store and handling activities in Building 4022. The particulate matter 

contained uranium, plutonium; cesium-137 (Cs-137), strontium-90 (Sr-90), krypton-85 (Kr-85) 

and promethium-147 (Pm-147) as mixed fission products; and cobalt-60 (Co-60) and 

europium-152 (Eu-152) as activation products. The particulate matter in air was controlled 

through filtration by HEPA filters. The ambient air within Area IV was reportedly monitored daily 

by continuous collection of air particulate samples using a network of eight air samplers. The 

samples were counted for alpha and beta radiation following a 120-hour delay to allow for radon 

and thorium decay. 

DOE monitored the stacks serving Buildings 4021 and 4022. The DOE survey report indicated 

that emissions from the RMHF Buildings 4021 and 4022 were higher than the equivalent 

ambient air emissions shown in the Rockwell International annual monitoring report data for 

1981-1987. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 61, citation 5:

Correspondence from Remley, M.E., Atomics International Division of Rockwell International, to Page, R.G., 

U.S. Atomic Energy Commission, Re: Physical Protection of Special Nuclear Materials Docket 70-25, August 

16, 1974. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 62 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 57, citation 3: 

3 Correspondence from Chell, M., MWH, to Waite, P., The Boeing Company, Re: Questions for Paul Waite 

Regarding RHMF [sic] Site History (Discussion took place via email during March, 2009), March 16, 2009. 

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EPA’s HSA references a 1991 DOE Tiger Team finding that stack emissions had not undergone 

formal evaluation in accordance with established National Emissions Standards for Hazardous 

Air Pollutants (NESHAP) regulations. Additionally, the HSA indicates that although the samplers 1

at the RMHF had the required continuous radiation monitors to detect sudden increases in 

radiation during accident situations, deficiencies in the radionuclide particulate sampling 

systems (which were not in use until 1970) prevented the samplers from meeting NESHAP

requirements. 

Examples of noted stack sampling deficiencies included suitability of sampling locations, 

number of sampling points, and sampler design. Moreover, the HSA indicates that the samplers 

at the RMHF were not originally designed to monitor the large range of particulates, which could 

be present as a result of HEPA filter problems, deposition inside the stack, or corrosion buildup 

in the stack. Due to design flaws, the radioactive particulate emission release rates that were 

supplied for modeling purposes by Rockwell International were determined from air filter 

samples, which were collected in a manner that may not have been representative of actual 

emissions. DOE’s Tiger Team observed that the changing and handling of the filter samples 

revealed that a loss of particulate matter may have occurred, and thus air emission modeling 

may not have been accurate.2 

According to DOE, “The principal source of potential radiation dose to the public from SSFL is 

the RMDF.” A representative of DOE also stated that, “Airborne dose assessment at this facility 

[RMHF] may be imprecise ... because of AIRDOS computer monitoring difficulties.”3 

In 1989, DOE Office of Environmental Audit indicated that air sampling and monitoring practices 

at SSFL were insufficient for the duration of site operations. Given documented onsite burning of 

radioactive waste and the presence of the RMHF incinerator, current estimations of airborne 

activity and emissions as they relate to internal and environmental ambient exposure may be 

inaccurate and even grossly underestimated, and therefore inappropriate for use in dose 

reconstruction. 

EPA’s HSA report provides detailed information on various surveys for radioactivity associated 

with the RMHF. For instance, in 1987 the Monthly Comprehensive Smear Surveys indicated that 

quarterly reviews noted high radiation levels measured by location film badges in the 

decontamination and packaging rooms, although gradual reductions had reportedly occurred in 

past quarterly surveys. 

In 1988, DOE indicated that even though improvement had been made to reduce radiation 

exposure rates, because of changing operations involving radioactive materials handling at the 

RMHF, exposure rates may have exceeded the DOE guideline of 100 millirems per year for 

continuous exposure at the Area IV property boundary north of the RMHF. The guideline was 

intended to prevent members of the public from unknowingly receiving excessive exposure as a 

1 DOE, Tiger Team Assessment, Energy Technology Engineering Center, April, 1991 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 28, citation 1:

1 U.S. Department of Energy, Tiger Team Assessment Energy Technology Engineering Center, DOE/ 

EH-0175, April 1991, pgs. 3-12–3-15. 

3 DOE Office of Environmental Audit, Environmental Survey Preliminary Report: DOE Activities at SSFL, 1989,

p. 4-31 - 4-33. 

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result of DOE operations. According to the report, long-term exposure to a member of the public 

was unlikely due to the rugged terrain along the north boundary and daily security patrols. 

However, no reference was made to concerns about potential worker exposure.1 

As of 2011, Building 4022 was still standing. Boeing noted that there was no storage of 

radioactive waste in Building 4022. However, the underground vaults of Building 4022 still had 

low levels of contamination and the HEPA ventilation ducting between Building 4021 and 

Building 4022 was highly contaminated. 

According to the HSA, the Radiological Contamination Potential based on the preliminary 

MARSSIM classification for the Building 4022 area was Class 1, due to the facility’s former use 

as a radioactive materials storage facility, documented radiological incidents, and previous 

investigations. 

11.3.2 Building 4022 Radiological Incident Reports 

There have been several incidents associated with Building 4022 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

11.3.3 Building 4022 Radiological Incident Report Summary, Data Provided by Boeing 

Incident 

File Name

Date of Incident Location of 

Incident

Isotopes Description of Incident

A0588 12/29/65 RMDF Yard Drums of Contaminated Sodium Exploded in Rain Storm and Burned

A0631 5/9/69 RMDF Vault Pit Employee Fell 12 Feet to Bottom of Sump Pit

A0589 5/19/77 RMDF High Bay Mixed Fission SRE Piping Caught Fire in the Sodium Melt Vessel at RMDF

A0078 4/5/79 High Bay Sr/Cs Recovery of Jammed Fuel Element Can in Hallam Transfer Cask

A0314 8/14/79 Gate to T022 Waste Truck Load Leaked R/A Liquid

A0239 10/29/79 Substation 783 R/A Exhaust Lost When Raccoon Shorted Out Substation and 

Generator Failed

A0086 5/28/81 High Bay MFP High Level R/A Liquid Ran Out of NAC Cask onto Floor Blocks

A0567 7/23/81 RMDF High Bay HP Contaminated Finger While Smearing “Clean” Tools

A0259 11/24/1981 Vault 7 Waste Mixed Fission Absorbed High Level NAC Waste Open to Atmosphere

A0326 6/30/82 High Bay RAS Alarm Response nappropriately Cancelled Because of Known 

Cause

1 EPA/HGL, Final HSA 7-3-NBZ, p. 46, citation 2:

2 Weiner, L.A. and Barisas, S., Status Report of the DOE Activities at the Santa Susana Field Laboratories 

Site Environmental Survey, U.S. Department of Energy, June 30, 1988, p. 10. 

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A0115 3/23/83 High Bay RAS Alarm Caused by High Radiation During Fuel Transfer

A0116 3/30/83 High Bay Response to RAS Alarm

A0158 6/11/86 High Bay Mixed Fission Off Scale Dosimeter During BM Cask Work

A0159 6/13/86 High Bay Mixed Fission Employee Placed Unprotected Hand on Contaminated Cask

A0161 6/25/1986 High Bay Mixed Fission Employee Contaminated Shoe During Decon of NAC Cask

A0166/

A0167

12/22/86 High Bay Cs/Co R/A Water Sprayed Employee During Fuel Cask Unloading / R/A Water 

Drained from Transfer Cask Contaminating Employee and Floor

A0168 1/6/87 High Bay Employees Shoes Found to be Contaminated

A0171 1/7/87 High Bay Mixed Fission Employee Contaminated During Routine Survey of Tools and 

Equipment

A0170 1/15/87 All RMDF Review of R/A ncident at RMDF

A0172 10/2/90 DeCon Rm 021 Mixed Fission Employees Became Contaminated While Unloading R/A Cable for 

Decon

A0209 2/21/97 High Bay Door 

Tent

Fixed R/A Contamination on Left Wrist and Personal Pants

A0674 3/25/97 High Bay Discreet Particle on Employee s Right Shoe

Incident 

File Name

Date of Incident Location of 

Incident

Isotopes Description of Incident

• A health physics logbook entry for the RMHF dated December 29, 1965 describes several 

loud explosions thought to be sonic booms by personnel in the RMHF Office building 4034. 

The explosions turned out to be coming from a sodium drum in a shipment of radioactive 

waste awaiting pickup. An emergency team of three firemen responded first followed by the 

health physics emergency team. Smoke from the fire was reported to blow away from the 

RMHF to an unpopulated area.

The logbook entry notes that weather conditions at the time (high winds and heavy rain) made 

normal operating procedures, including air sampling and smearing, difficult. an exclusion zone 

was established and equipment and personnel were monitored prior to leaving the area. One 

fireman’s uniform was found to have contamination at 90 dpm. All other uniforms were at 

background, below .1 mrad/hr and 30 dpm. A water sample taken from rain runoff indicated 

1x10-5 μCi/cc. A handwritten note obtained from Boeing indicates that on December 29, 1965, 

drums of contaminated sodium exploded in a rain storm and “burned.” The note indicated 

there was a fire, liquid spill, outside contamination, and airborne release to the environment. A

formal incident report for this incident could not be located in available documents (A0588).1

• On May 19, 1977 the “Sodium Melt Vessel lid” was removed in preparation for removing the 

SRE piping from the vessel. Approximately 20 minutes following the removal of the lid, it was 

reported that smoke was detected coming from the vessel. An employee took the dry 

chemicals fire extinguisher and unloaded it into the vessel and the employees present 

evacuated the building. Two employees wearing self-contained breathing apparatus 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 71, citation 2:

2 

Unknown Author, Handwritten Note: 588, 12-29-65, RMDF Yard, Unknown Date. 

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equipment entered the building to put the lid back on the melt vessel to extinguish the fire.

Air samples were reported to have resulted in an immediate count of 2.3x10-10 μCi/cc and a 

1-hour delay count of 2.0 x 10-10 μCi/cc. Nasal smears of the employees present during the 

incident were reported to have been negative. Following the incident, the vault area was damp 

mopped and allowed to dry. The incident report indicated that the cause of the fire was either 

excess alcohol or deficient argon gas purge. It was also indicated that the RMDF air stack 

sampler was removed and analyzed and found to be “negative.” On May 20, 1977 130 smears 

taken in the vault area indicated for smears with a maximum count of 75 dpm. These areas 

were reported to have been cleaned and the building was released. The incident report did not 

indicate the background levels of the building (A0589).1

• On August 14, 1979 an outside vendor, Heavy Transport, Inc., truck-trailer was loaded with 

approximately 44,000 pounds of low-level radioactive waste destined for Beatty, Nevada for 

burial. The shipment contained six 8-food long boxes with the remainder of the waste in 55-

gallon drums. The truck-trailer was moved outside the immediate RMHF facility area to a lowbackground area for a final U.S. Department of Transportation radiation survey. The trailer 

was parked on the left side of the facility access road and was tilted slightly toward the outside 

edge of the roadway. Soon after moving the vehicle, liquid was observed dripping from the 

trailer to asphalt apron. A shipping box leaked approximately 1 pint of radioactive liquid 

containing cesium-137 (Cs-137) and strontium-90 (Sr-90) on the asphalt outside the RMHF 

and on the trailer. An immediate survey found readings of about 85 mrad/hr on the asphalt and 

125 mrad/hr on the trailer. According to the incident report, “protective measures were taken to 

contain and control the area and liquid escaping from that location.” This included the 

placement of a bucket to catch what was suspected to be alcohol and a rope barrier was 

established to control access to the spill area.

Alcohol splattered onto a rear tire and a mud flange, which were covered by personnel using 

green tape. The truck and trailer were moved back into the controlled area to unload the 

leaking shipping box. The incident report indicated that plastic was used to catch additional 

leakage and to cover the road to protect uncontaminated tires. One employee contaminated a 

white lab coat to about 700 cpm, which resulted in the lab coat being placed in a radioactive 

waste receptacle for disposal.

Upon removal of the shipping containers, it was determined that alcohol dripped through the 

wooden floor and contaminated part of the undercarriage of the trailer. Personnel were 

assigned to clean this radioactive material, which took approximately three days to reduce all 

detectable activity to 200 cpm or less. This included the installation of new wooden slats and 

the painting of all bare metal surfaces as necessary. The incident report did not indicate how 

the asphalt apron along the RMHF access road was decontamination (A0314).2

1 EPA/HGL, Final HSA 7-3-NBZ, p. 71, citation 3:

3 Bradbury, S.M., Rockwell International Internal Letter, Subject: Fire in the Sodium Melt Vessel in Building 

022, May 23, 1977. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 72, citation 1:

1 Owens, D.E., Rockwell International Internal Correspondence, Re: Radiological Safety Incident

Report, RMDF, August 14, 1979, A0314, August 20, 1979. 

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• On May 28, 1981 in the Building 4022 high bay, personnel were rotating “the N.A.C. cask” 180 

degrees on the longitudinal axis from the normal transport position. The cask was then 

lowered to the horizontal position. This was to allow the opening of a drain plug that normally 

faced down in transport. The incident report indicated that as the cask approached horizontal, 

liquid ran out the top vent line hole and the cover port located “135 degrees below top dead 

center.” The liquid was reported to have run down the cask support member to the belly pan 

and through holes in the corrugated metal pan to the uncovered vault floor blocks. Personnel 

spread absorbent towels over the spill and placed a bucket under the belly pan to catch the 

liquid draining from the belly pan. The cask was elevated to stop the flow of liquid.

The contents of the tank were reported to have been approximately 1-year old mixed fission 

products from a failed power reactor fuel element that was suspended in liquid. The dose rate 

from the major accumulation of liquid in the bucket used to catch the liquid was 25 rad/hr beta 

gamma at 4 inches above the bucket. One spot on the floor indicated 25 rad/h were 17 R/hr at 

2 inches. It was estimated that 50 gallons of liquid were located in the cask. The incident 

report did not indicate what volume spilled

The available incident report indicated that additional summaries were provided on the back; 

however the research team only received one page of the incident report. As a result, 

information regarding the contamination and cleanup of the incident could not be summarized 

by the EPA research team (A086). 1

• A 1985 second quarter radiation safety review for RMHF Building 4022 indicated that fixedlocation dosimeters were placed at eight locations around the RMHF perimeter fence. Results 

from the initial monitoring period showed a maximum exposure of 810 mrem at the northwest 

perimeter fence. This result is above the annual limit of 500 mrem/yr for radiation exposure in 

uncontrolled areas.2

• A 1986 first quarter radiation safety review for the RMHF again showed a high exposure 

(1,140 mrem) at the northwest perimeter fence. It was noted that the possibility of exposure to 

the general public was unlikely, due to the terrain north of the RMHF fence line and the 

continuous remote surveillance of the area. Apparently, the possibility of exposure to 

employees in the area was not referenced in the report.3

1 EPA/HGL, Final HSA 7-3-NBZ, p. 72, citation 2:

2 Badger, F.H., Rockwell International Internal Letter, Subject: Radiological Safety Incident Report, T022 – 

High Bay, May 28, 1981, June 9, 1981. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 73, citation 1:

1 Internal Correspondence from Moore, J.D. to Remley, M.E., Rockwell International, Re: Quarterly Review 

of the RMDF (T022) for Radiation Safety, Second Calendar Quarter, 1985, September 18, 1985. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 73, citation 2:

2 Internal Correspondence from Moore, J.D. to Remley, M.E., Rockwell International, Re: Quarterly Review 

of the RMDF for Radiation Safety, First Calendar Quarter, 1986, June 23, 1986. 

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Section 11.4 

Building Number:

Building Alias: 

Visitor Log Entry: 

4034 

034/T034

6017 / 6000E / 6001R / RMDF / T034 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Shipping / Receiving 

Note: Designated “Non-Radiological.”

Workers may not have been adequately monitored.

Main entry/exit point at RMHF; “Authorized Use Location”

Radioactive Materials / Laundry / Receiving / Processing 

Notes: Several incidents at 4021 & 4022, detailed in the Incident Reports Database, describe 

incidents that resulted in the ventilation systems of Building 4034 being turned off to avoid 

personnel exposure and possible contaminant migration from RMHF 4034. In addition, tritium is 

listed among the radionuclides of concern associated with this location.

Radionuclides of Concern: EPA’s research team found no evidence that nuclear or radioactive 

materials were known to have been stored or handled in Building 4034. However, because 

Building 4034 is located southeast of the Building 4022 RMHF storage vault, contamination 

could have migrated from Building 4022 to Building 4044. Radionuclides associated with 

potential migration from Building 4022 include: uranium fuel materials (U-234, U-235, U-238), 

transuranic isotopes of plutonium (Pu-238, Pu-239, Pu-240, Pu-241, Pu-242), and americium241 (Am-241), fission products cesium-137 (Cs-137) and strontium-90 (Sr-90), thorium breeder 

material (Th-228, Th-232), and possible neutron activation products such as cobalt-60 (Co-60), 

europium isotopes (Eu-152, Eu-154), hydrogen-3 (H-3), iron-55 (Fe-55), nickel isotopes (Ni-59, 

Ni-63), managense-54 (Mn-54), potassium-40 (K-40), and sodium-22 (Na-22). 

11.4.1 Description of Operations and Processes: 

According to the EPA HSA, Building 4034 was constructed in 1961 for intended use as an Office 

Building. Building 4034 served as the main office and point of entry for RMHF, and was among 

those buildings listed as “authorized use locations” for the receiving, processing, packaging and 

shipping of radioactive wastes and materials generated by Atomics International activities: 

receiving processing, packaging and shipping of radioactive laundry; receiving and processing 

liquid radioactive wastes; storage of radioactive materials and maintaining facilities equipment 

and material. Buildings 4021, 4022, 4044, 4075, and 4621 were also listed as authorized use 

locations with the authorized users being based out of Building 4034.1 

Because use authorization applications list the location of the principle user, most use 

authorizations identified for the RMHF complex indicate that the principle user was located at 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 82, citation 1: 

.	 1 Kneff, D.W. and R.D. Meyer, Radioactive Materials Handling Facility (RMHF) Demolition Program

Management Plan, PMP-00008, The Boeing Company, January 30, 2004, pgs. 10-11, 19. 

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the RMHF Building 4034. However, because Building 4034 may be considered an 

administrative building, it may be referenced in Boeing Employment Summaries and Document 

Acquisitions Requests (DARs) as a “non-radiological” work location, where an assigned

employee may not have been monitored for radiation exposure or assumed to have been at 

risk. However, given the documented facility use and operations, such an assertion could result 

in an underestimation of the health risks associated with Building 4034. 

The preliminary MARSSIM classification for the Building 4034 area is Class I, due to its 

proximity to RMHF Building 4022 and the potential for contaminant migration. 

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Section 11.5 

Building Number:

Building Alias: 

Visitor Log Entry: 

4044 

044/T044

6017 / T022 / T034 / 6000E / 6001R 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Clean Shop

Radiation Safety Office

Health Physics Instrumentation

Technician Support Area

Break Room 

Storage 

Notes: Possible Tritium Contamination. 

Possible “Non-Radiological” designation.

Workers may not have been adequately monitored for radiation exposure. 

Radionuclides of Concern: As the health physics office, Building 4044 was used to count 

removable contamination measurements and store calibration sources. A 1989 quarterly 

radioisotope inventory for Building 4044 found sources containing the following radionuclides: 

Bi-210, Pb-210, Pu-239, Sr-90, Tc-99, and Th-230. An historical photo (c. 1990) shows two 

small containers labeled “Radioactive Material” and “Radioactive Material - Contaminated 

Waste” under a desk in Building 4044.1 The radiological incident associated with this building 

(described below) notes that mixed fission products were handled in Building 4044. 

In addition, because Building 4044 is located just east of the Building 4022 RMHF storage vault, 

contamination could have migrated from Building 4022 to Building 4044. Radionuclides 

associated with potential contaminant migration from RMHF Building 4022, according to EPA, 

include: U-234, U-235, U-238, Pu-238, Pu-239, Pu- 240, Pu-241, Pu-242, Am-241, Cs-137, 

Sr-90, Th-228, Th-232, Co-60, Eu-152, Eu-154, H-3, Fe-55, Ni-59, Ni-63, Mn-54, K-40, and 

Na-22. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 89, citation 6:

6 

Rockwell International, Photograph of SSFL Building 44 RMDF Clean Shop, circa 1990, HDMSP00048945.
 

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11.5.1 Description of Operations & Processes 

Building 4044 was constructed in the mid-1960’s as the RMDF Clean Shop.1 The building 

contained a radiation safety office with health physics instrumentation and a technician support 

area / break room. Building 4044 contained radiation counters for radioactive surveillance of the 

RMHF. The health physics office had been used as a counting area for removable 

contamination measurements, storage, and use of calibration sources. Radioactive waste was 

generated as part of operations at Building 4044.2 

As of July 27, 2011, according to Boeing, there were 10 assorted micro-curie level check 

sources in use at the Health Physics Laboratory in Building 4044, containing cesium-137 

(Cs-137), strontium-90 (Sr-90), technetium-99 (Tc-99), thorium-230 (Th-230), and uranium-238 

3 (U-238). As of 2012, Building 4044 was still active as a break room and health physics office for 

the RMHF. 

The preliminary MARSSIM classification for Building 4044 is Class I, due to its use as a health 

physics counting office and proximity to Building 4022, the RMHF storage vault. 

11.5.2 Building 4044 Radiological Incident Reports 

There has been one incident associated with Building 4044 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 86, citations 1, 2, 3: 

1 The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning

Engineering Evaluation/Cost Analysis, June 18, 2007, p. A-4.

2 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992.

3 Kartman, A.S., Aerial Photographic Analysis of Santa Susana Field Laboratory – Area IV, Ventura County , 

California, Volume 1 & 2, U.S. Environmental Protection Agency, Office of Research and Development, 

Environmental Sciences Division, March 2010. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 86, citations 1-3:

1 The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning

Engineering Evaluation/Cost Analysis, June 18, 2007, p. A-4.

2 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992.

3 Kartman, A.S., Aerial Photographic Analysis of Santa Susana Field Laboratory – Area IV, Ventura County , 

California, Volume 1 & 2, U.S. Environmental Protection Agency, Office of Research and Development, 

Environmental Sciences Division, March 2010. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 87, citations 3-4:

3 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. I-13.

4 

MWH, RCRA Facility Investigation Work Plan Addendum Amendment Radioactive Materials Handling 

Facility RFI Site (SWMU 7.6 and Area IV AOC) Santa Susana Field Laboratory, Ventura County, California, 

March 2008. 

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11.5.3

Building 4044 Radiological Incident Report Summary Based on Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0082 7/24/80 RMDF Mixed Fission Radioactive Waste Container Emptied 

Into Clean Waste Gondola.

• On July 24, 1980 it was discovered that a 5-gallon radioactive waste container from the 

counting room of Building 4044 was inadvertently emptied into a regular trash gondola on July 

23, 1980. Upon discovering the error, the trash gondola was returned to the RMHF Building 

4022 and had the entire contents placed into a proper radioactive waste disposal container. 

The incident report did not indicate the location of the trash gondola prior to being transported 

back to the RMHF Building 4022 (A0082).1

The EPA HSA provides a chronology of radiological investigations associated with this facility. 

Use Authorization 106 pertained to radioactive waste and contaminated materials generated at 

other authorized facilities. Use Authorization 106M, issued February 12, 1992 noted that 

operation of the RMHF included receiving, processing, packaging, and shipping radioactive 

wastes and materials generated by Rocketdyne activities; receiving, processing, packaging and 

shipping radioactive laundry; receiving and processing liquid radioactive wastes; storing 

radioactive materials; and maintaining facilities equipment and material. Buildings 4021, 4022, 

4044, 4075 and 4621 were listed as authorized use locations with the authorized users based 

out of Building 4034. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 86, citations 1-3: 

2 Owen, R.K., Rockwell International Internal Correspondence, Re: Radiological Safety Incident Report,

RMDF, July 24, 1980, August 20, 1980. 

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Section 11.6 

Building Number:

Building Alias: 

Visitor Log Entry: 

4075 

075/T075

6017 / T022 / T034 / 6000E / 6001R 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Contaminated Equipment Storage Building

Staging Area for Radioactive Waste Shipments 

Radionuclides of Concern: Possible radionuclides of concern include uranium, plutonium, 

thorium, and mixed fission products. The incident database described isotopes of cobalt and 

europium as being part of the radioactive waste that was released.1 

11.6.1 Description of Operations & Processes 

Building 4075 was constructed in 1971 as the Contaminated Equipment Storage Building and 

stored low specific activity packaged waste and was primarily used for staging of waste 

shipments. 

There were no exhaust filter systems or additional contamination control safeguards 

implemented at this building. Building 4075 had no air conditioning, water supply or natural gas 

connection. EPA’s 2009 site visit noted that Building 4075 was “designed so that the floor drain 

was located in the topographic low point of the building, so any spilled chemicals would flow to 

the drain.” The drain was controlled by a valve and drained via subsurface pipeline into the 

storm water culvert, which eventually directed drainage to the Site-Wide Water Reclaim System 

that transported water throughout all areas of SSFL, for use by personnel. Building 4075 was 

considered a Resource Conservation and Recovery Act (RCRA) permitted building for the 

support and storage of radioactive wastes (low level, mixed, and transuranic wastes) pending 

shipment. 

According to the HSA, an Atomics International site waste management plan dated circa 1972 

and an updated plan dated May 1974 provided information regarding the waste disposal 

activities at SSFL. According to the waste management plan, which was prepared for the 

management of AEC contract wastes, provided the following summary: 

“Temporary storage of low-level radioactive waste is performed in Buildings 621, 663 and 075. 

Low-level waste and contaminated equipment are temporarily stored in Buildings 075, 621, and 

663, and in the fenced storage area surrounding Building 621. These buildings are steel 

buildings, which are designed to provide protection from the elements and security for the 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 100 

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stored items. Contamination control is provided by stringent packaging requirements, so 

exhaust filter systems and additional contamination control safeguards are unnecessary.” 1 

The HSA references a November 1975 Rockwell letter, which notes that at the time Building 

4075 contained stored Systems for Nuclear Auxiliary Power (SNAP) equipment and packaged 

material awaiting shipment for burial. The SNAP program ended in 1973. It is unknown exactly 

when the SNAP materials were transported to the RMHF for storage, or how long they were 

present before the 1975 Rockwell letter was issued. The 1975 Rockwell letter states that while 

the SNAP waste was inside the building awaiting transport to a DOE disposal site, steel billet 

stored at the RMHF complex could be used to form a bay inside Building 4075 and lead 

shielding currently stored at the RMHF could be hung on the walls to provide more adequate 

shielding. The additional shielding was requested expeditiously to bring radiation levels into 

acceptable DOE limits at the time.2 

The generation of the 1975 Rockwell request regarding additional shielding to store SNAP

materials suggests that similar improvisation may have been required or requested on occasion, 

depending on materials stored at this location. This presents concerns over assumptions of safe 

working conditions and data that may be applied during dose reconstructions for SSFL

employees, which may reflect underestimated dose. 

Although the building was designated as a “low-level” storage facility, the HSA references an 

August 15, 1988 site visit report noting solid radioactive materials were stored inside Building 

4075 in drums or large metal containers, or were “temporarily” stored in the RMHF yard during 

staging procedures. On the day of the site visit, the yard contained portions of approximately 

300 drums of cobalt-60 (Co-60) contaminated sand from the Building 4059 (SNAP) area. Drums 

3 stored in the yard were destined for Hanford, Washington for disposal. Aerial photos taken in 

1976 depict a disturbed area on the hillside north of Building 4075, suggesting a radiological 

burial or disposal location.4 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 93, citations 1&2:

1 

Atomics International, Site Waste Management Plan, Circa 1972, p. 2.

2 

Atomics International, Management of AEC-Generated Radioactive Wastes at Atomics International, May

31, 1974, pgs. 1-2. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 94, citation 3:

3 Internal Correspondence from Harris, J.M. to McCurnin, W.R., Rockwell International, Re: Plan of Action 

for RMDF, November 18, 1975. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 94, citation 5: 

5 Lavagnino, G., ETEC Meeting and Site Visit, U.S. Department of Energy, August 17, 

1988. 

4 EPA/HGL, Final HSA 7-3-NBZ, p. 99, citations 2&3: 

2 

MWH, RCRA Facility Investigation Work Plan Addendum Amendment, Radioactive Materials Handling 

Facility RFI Site (SWMU 7.6 and Area IV AOC), Santa Susana Field Laboratory, Ventura County, California, 

March 2008, p. 2-4.

3 

MWH, RCRA Facility Investigation Work Plan Addendum Second Amendment, Radioactive Materials 

Handling Facility RFI Site (SWMU 7.6 and Area IV AOC), Santa Susana Field Laboratory, Ventura County, 

California, October 2008, p. HDMSP00015875. 

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According to the HSA, the preliminary MARSSIM classification for the Building 4075 area is 

Class 1, due to its former use and a documented release of radiological material. 

11.6.2 Building 4075 Radiological Incident Reports 

There has been one documented incident associated with Building 4075 that could have 

resulted in a release to the environment and worker exposure. The following table provides 

information presented in an incidents database provided by Boeing. 

11.6.3 

Building 4075 Radiological Incident Report Summary Based on Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0188 8/15/1988 Radioactive 

Storage 075

Co/Eu Fork Truck Operator Punctured 55 Gal 

Drum of R/A Waste.

• On August 15, 1988 a forklift driver punctured two drums containing radioactive sand that 

originated from Building 4059. approximately 40 grams of sand spilled onto a 200 square 

centimeter area of the drum pallet and floor. One-half gram of the sand had a beta gamma 

reading of 24,000 dpm. The spilt sand was vacuumed up into a radioactive vacuum cleaner, and 

the punctures were sealed with duct tape. The drums were transported to Building 4021 

(decontamination room) and the remaining sand was transferred to new drums. Post cleanup 

surveys found no detectable contamination (A0188).1

1 EPA/HGL, Final HSA 7-3-NBZ, p. 96, citations 2&3:

2 McGinnis, E.R., Rockwell International Internal Correspondence, Re: Radiological Safety Incident Report,

RMDF- T075, August 15, 1988, August 23, 1988.

3 Moore, J.D., Rockwell International Internal Correspondence, Re: Quarterly Review of the Radioactive

Materials Disposal Facility (RMDF) for Radiation Safety, Third Calendar Quarter, 1988, December 22, 1988. 

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Section 11.7 

Building Number: 4654 

Building Alias: 654/T654

Visitor Log Entry: 6017 / T022 / T034 / 6000E / 6001R 

Building Name:	 Radioactive Materials Handling Facility (RMHF) 

Building Function:	 Interim Radioactive Waste Storage Facility (ISF)

Sodium Reactor Experiment (SRE) Support & Storage

Organic Moderated Reactor Experiment (OMRE) Waste Storage

Systems for Nuclear Auxiliary Power (SNAP) Waste Storage 

Radionuclides of Concern: The primary radionuclides of concern for Building 4654 are 

uranium and mixed fission and activation products.1 

11.7.1 Description of Operations & Processes 

Building 4654 was constructed in 1958 as the Interim Storage Facility (ISF) to support the 

Sodium Reactor Experiment (SRE). The ISF was designed to store dummy and spent fuel 

elements, shipping and storage casks, and hot waste generated at the SRE. After the SRE 

stopped operating, Building 4654 was used to store waste from the Organic Moderated Reactor 

Experiment (OMRE) and the Systems for Nuclear Auxiliary Power (SNAP) program. While the 

EPA HSA indicates that Building 4654 was taken out of service between 1964-1984, it is 

noteworthy that the SNAP program’s years of research and development were 1960’s through 

the late 1970’s. Low-level radiation was released from the storage of containers at this facility.2 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 193, citation 2:

2 

Vitkus, T.J., Verification Survey of the Interim Storage Facility (T654), Santa Susana Field Laboratory, 

Rockwell International, Ventura County, California, ORISE 97-1900, Oak Ridge Institute for Science and 

Education, November 1997, p. 7. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 184, citations 1-8: 

1 Johnson, R.P. and Speed, D.L., Interim Storage Facility Decommissioning Final Report, ESG-DOE-13507, 

Rockwell International, Rocketdyne Division, March 15, 1985, pgs. 1, 3.

2 Shah, Satish, N., Final Report Decontamination & Decommissioning of Interim Storage Facility 4654,

EID-04364, The Boeing Company, May 20, 1999, p. 5.

3 Dahl, Farley, C., Building T654 Supplemental Final Radiological Survey Report, SSWS-AR-0011, The 

Boeing Company, January 20, 1999, p. 3.

4 

MWH, Group 7 – Northern Portion of Area IV RCRA Facility Investigation Report, Santa Susana Field 

Laboratory, Ventura County, California, Volume II – RFI Site Reports, June 2009, pg. B.2-7.

5 Shah, Satish, N., Final Report Decontamination & Decommissioning of Interim Storage Facility 4654,

EID-04364, The Boeing Company, May 20, 1999, p. 5.

6 

MWH, Group 7 – Northern Portion of Area IV RCRA Facility Investigation Report, Santa Susana Field 

Laboratory, Ventura County, California, Volume II – RFI Site Reports, June 2009, pg. B.2-7.

7 Johnson, R.P. and Speed, D.L., Interim Storage Facility Decommissioning Final Report, ESG-DOE-13507, 

Rockwell International, Rocketdyne Division, March 15, 1985, p. 1.

8 Shah, Satish, N., Final Report Decontamination & Decommissioning of Interim Storage Facility 4654,

EID-04364, The Boeing Company, May 20, 1999, p. 12. 

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Building 4654 was open to the environment, which provided the potential for surface water to 

seep into the storage tubes and to possibly corrode the steel pipes. This building design 

increased the potential for contaminated water to leak into the subsurface below the storage 

tubes. 

A 1964 inspection tour of SSFL fuel storage areas notes that one of the SRE Core I fuel 

elements removed from the Building 4654 storage pits contained in an unleaded tube, which 

separated just as it was being fitted into the RMHF storage thimble. It was recommended that 

the tube be weld-sealed for contamination control. The inspection report notes that the Building 

4654 storage area was also temporarily being used to store classified scrap materials.1 

11.7.2 1995 ORISE Verification Survey 

In 1995, the Oak Ridge Institute for Science and Education (ORISE) performed independent 

verification of Rocketdyne’s final status survey of Building 4654 following 1984-1985 D&D 

efforts. ORISE determined that exposure rates at Building 4654 ranged from 10 to 15 μR/h.

According to ORISE, background rates ranged from 12 to 16 μR/h with an average of 14 μR/h.

The NRC exposure rate limit is 5 μR/h above background. 

ORISE conducted a document review and independent measurement and sampling. ORISE 

found deficiencies in Rocketdyne’s final status documentation. Deficiencies included inadequate 

final status survey methods, no discussion of specific contaminants, inconsistent specification of 

all applicable guidelines and presentation of data that may be compared to the guidelines, 

absence of quantitative laboratory data, and inconsistent presentation of adequate figures 

documenting remediated areas and measurement and sampling locations.2 

11.7.3 Building 4654 Radiological Incident Reports 

There have been a couple of incidents associated with Building 4654 that could have resulted in 

a release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 185, citation 1:

1 Internal Correspondence from Matten, K.L. to Fuel Committee Members, Atomics International, Re: 

Inspection Tour of Santa Susana Fuel Storage Areas, April 23, 1964. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 185, citations 5,6,1: 

5 

Vitkus, T.J. and Bright, T.L., Verification Survey of the Interim Storage Facility; Buildings T030, T641, and 

T013; an Area Northwest of Buildings T019, T013, T012, and T059; and a Storage Yard West of Buildings 

T626 and T038, Santa Susana Field Laboratory, Rockwell International, Ventura County, California, ORISE 

96/C-4, Oak Ridge Institute for Science and Education, February 1996, pgs. 5-15.

6 Shah, Satish, N., Final Report Decontamination & Decommissioning of Interim Storage Facility 4654,

EID-04364, The Boeing Company, May 20, 1999, pgs. 11, 14.

1 

MWH, Group 7 – Northern Portion of Area IV RCRA Facility Investigation Report, Santa Susana Field 

Laboratory, Ventura County, California, June 2009, p. 3-12. 

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The preliminary MARSSIM classification for the Building 4654 area is Class 1, due to the 

building design, its former use, radiological incidents, and previous investigations. 

11.7.4 Building 4654 Radiological Incident Report Summary, Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0014 1/23/1962 T654 R/A STORAG Mixed Fission Loss of Containment of R/A Equipment 

Stored Outside

A0079 7/5/79 654 R/A YARD Mixed Fission Contamination Found on Pavement and 

Outside Storage Yard Fence

• On January 23, 1962 a radiation survey of Building 4654 was conducted and revealed 

contamination had spread from the inside of the fenced area to the asphalt outside the fence. 

At that time, the contamination level outside the fenced area of Building 4654 was between 2 

and 3 mrad/hr. The contamination on exposed equipment surfaces inside the fence was as 

high as 17 rad/hr. This information was written in the SRE operations logbook. The incident 

report indicated that no action was taken to decontaminate the area or prevent further spread 

of contamination. On February 13, 1962 a similar survey was conducted and indicated an 

increased spread of contamination due to rain. The contamination on the pavement was 

between 2 and 17 mrad/hr outside the enclosure. Exposed equipment surfaces inside the 

enclosure showed loss or displacement of 5 rad/hr of contamination as the maximum 

detectable level was 12/rad hr. Four soil samples were taken along the drainage pathways. 

Soil contamination ranged from 20-30 μCi/g. Additional information regarding any cleanup of 

the area, or whether contamination continued to spread, was not provided in the incident 

summary (A0014).1

• During a preliminary survey of the storage yard on July 5, 1979 contaminated shipping casks 

stored in the area were found to be emitting high levels of radiation. These high levels were 

noted in the northeast corner of the fenced area. An area of asphalt measured greater than 

50,000 dpm. A smear of a 3-gallon shipping cask found radioactive contamination of 12,000 

dpm and an instrument survey indicated 35 mrad/hr at the cask. Some of the other casks 

surveyed indicated radioactive intensities of 40 mrad/hr. A preliminary survey of the soil 

outside the fenced area indicated possible soil contamination along the northeast fence line 

along the east side ranging from 200 to 400 cpm. The incident report did not include 

information to indicate any clean up operations of the area (A0079).2

• Seals and packing on some casks and equipment stored at the ISF deteriorated from 

exposure to the elements to such an extent that low-level contamination was released. This 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 186, citation 1:

1 Badger, F.H., Atomics International Internal Letter, Re: Spread of Contamination in Area # 654, February 

18, 1962. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 186, citation 3:

3 Owen, R.K., Handwritten Note Re: SRE Interim Storage Yard, July 5, 1979. 

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release contaminated the asphalt surface near the casks and soil just outsisde the ISF fence. 

The casks and other sources of potential contamination were sent to an offsite disposal 

facility. Radioactive core components placed in the storage tubes contaminated the internal 

storage cells. The exact location where the casks and equipment were stored inside the ISF 

fenced storage area is unknown.1

• During D&D excavation activities the hydraulic hammer mounted on the end of the backhoe 

punctured storage tube 7. The area was surveyed for contamination but none was found.2 

11.7.5 Radiological Investigations 

The EPA HSA contains a detailed chronology of radiological investigations and decontamination 

procedures associated with Building 4654. Of note is that 1995 ORISE Verification Survey 

following Rocketdyne’s 1984-1985 D&D efforts. ORISE found deficiencies in Rocketdyne’s final 

status documentation, which included inadequate final status survey methods, no discussion of 

specific contaminants, inconsistent specification of all applicable guidelines and prsentation of 

data that may be compared to the guidelines, absence of quantitative laboratory data, and 

inconsistent presentation of adequate figures documented remediated areas and measurement 

and sampling locations. 

In addition, ORISE determined that Rocketdyne’s surface scans for alpha, beta, and gamma 

activity did not identify any locations of residual contamination. Exposure rates for the ISF 

ranged from 10-15 μR/h. According to ORISE, background rates ranged from 12 to 16 μR/h with 

an average of 14 μR/h. The NRC exposure rate limit is 5 μR/h above background. 

ORISE was unable to verify the radiological status of the ISF because of document deficiencies 

and recommended additional surveys. A supplemental survey plan was developed by 

Rocketdyne for a cooperative soil sampling effort with ORISE.3 Several subsequent 

investigations of the location were performed. On Febuary 1, 2005 DOE released Building 4664 

for unrestricted use.

1 EPA/HGL, Final HSA 7-3-NBZ, p. 186, citation 4:

4 Johnson, R.P. and Speed, D.L., Interim Storage Facility Decommissioning Final Report, ESG-DOE-13507, 

Rockwell International, Rocketdyne Division, March 15, 1985, p. 1. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 186, citation 5:

5 Johnson, R.P. and Speed, D.L., Interim Storage Facility Decommissioning Final Report, ESG-DOE-13507, 

Rockwell International, Rocketdyne Division, March 15, 1985, pgs. 9, 13. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 190, citations 5,6,1:

5 Vitkus, T.J. and Bright, T.L., Verification Survey of the Interim Storage Facility; Buildings T030, T641, and 

T013; an Area Northwest of Buildings T019, T013, T012, and T059; and a Storage Yard West of Buildings 

T626 and T038, Santa Susana Field Laboratory, Rockwell International, Ventura County, California, ORISE

96/C-4, Oak Ridge Institute for Science and Education, February 1996, pgs. 5-15.

6 Shah, Satish, N., Final Report Decontamination & Decommissioning of Interim Storage Facility 4654,

EID-04364, The Boeing Company, May 20, 1999, pgs. 11, 14

1 

MWH, Group 7 – Northern Portion of Area IV RCRA Facility Investigation Report, Santa Susana Field 

Laboratory, Ventura County, California, June 2009, p. 3-12.

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Section 11.8 

Building Number:

Building Alias: 

Visitor Log Entry: 

4665 

665/T665

6017 / T022 / T034 / 6000E / 6001R 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Oxidation Facility

Equipment Storage

Longterm Sodium Storage (SRE)

General & Hazardous Materials Storage 

Radionuclides of Concern: Radioactive waste and material may have been stored or handled 

at this facility. The most probable radionuclides of concern are isotopes of uranium, plutonium, 

thorium, and mixed fission products. 

11.8.1 Description of Operations & Releases: 

Building 4665 was constructed in 1964 as the Oxidation Facility, and was connected to Building 

4664 via a water line. Building 4665 was used as an oxidation facility and equipment storage 

area for the RMHF. It was used for long-term storage of sodium that had previously been 

temporarily stored in Building 4022 and at the Sodium Reactor Experiment (SRE, Building 

4143). An undated fire preplan report notes that Building 4665 was also used for the storage of 1

paper, plastics, paints, filters, and lighting.2 

A November 1996 building reconnaissance report states that the building had a minor crack in 

its floor and was used for the storage of hazardous materials. Equipment and supplies were 

noted throughout the building. The report states that less than 300 gallons of hazardous 

material was in the building storage racks at the time.3 

The preliminary MARSSUM classification for the Building 4665 area is Class 1, due to its 

potential for contamination by storing items previously stored in Building 4022, the RMHF 

storage vault, as well as its proximity to Building 4022. According to Boeing, there are no 

accidental releases or incident reports associated with Building 4665. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 146, citations 1, 2, 12:

1 Carroll, J.W. et al., RMDF Leach Field Decontamination Final Report, ESG-DOE-13385, Rockwell 

International, September 15, 1982, p. 11.

2 

Atomics International Drawing, Radioactive Scrap Oxidation Facility, Building 665, Santa Susana Field 

12The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning 

Engineering Evaluation/Cost Analysis, June 18, 2007, p. A-5. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 147, citation 3: 

3 

Unknown Author, RMDF Fire Preplans – Structures, Unknown date. 

3 3 EPA/HGL, Final HSA 7-3-NBZ, p. 147, citation 4:

4 

Unknown Author, Building Reconnaissance Report, Building 665, November 8, 1996. 

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Section 11.9 

Building Number: Leach Field Area 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Notes: Possible source of Tritium in Groundwater. 

Radionuclides of Concern: Radionuclides of concern detected in the RMHF Leach Field 

include Cs-137, Sr-90, and Y-90. 1 Other radionuclides of concern likely associated with the 

RMHF Leach Field are those associated with Building 4021, primarily mixed fission products 

and fuels.2 uranium fuel materials (U- 234, U-235, U-238), transuranic isotopes of plutonium 

(Pu-238, Pu-239, Pu-240, Pu-241, Pu- 242) and americium-241 (Am-241), fission products 

Cs-137, Sr-90, krypton-85 (Kr-85), and promethium-147 (Pm-147), thorium breeder material 

(Th-228, Th-232), and possible neutron activation products such as cobalt-60 (Co-60), europium 

isotopes (Eu-152, Eu-154), hydrogen-3 (H-3), iron-55 (Fe-55), nickel isotopes (Ni-59, Ni-63), 

managense-54 (Mn-54), potassium-40 (K- 40), sodium-22 (Na-22), Fe-55, Kr-85, and Mn-54. 

11.9.1 Description of Operations & Releases: 

The RMHF Leach Field area comprises the RMHF Leach Field and the land surrounding it north 

of the RMHF fence line and near the northern edge of the Area IV boundary. The RMHF Leach 

Field was constructed in early 1959 as an irregular-shaped excavation that was backfilled with 4 

feet of gravel, covered with a paper barrier, and then covered with native soil to the original 

grade. The Leach Field was the gravity-fed disposal point for a 1,500-gallon septic tank sanitary 

waste system that accepted waste from the Building 4021 lavatories, shower, and toilets. A

second connection to the RMHF Leach Field, bypassing the septic tank, was made from the 

radioactive water processing system at the waste holdup tank, located on the west side of 

Building 4021. Liquid waste from the radioactive holdup tank was reportedly discharged to the 

leach field only after sample analysis indicated radioactive contamination was within “acceptable 

3 limits.” The term, “acceptable limits” is an unknown context, and likely does not apply to current 

standards or with the December 2010 Administrative Order on Consent (AOC) regarding 

environmental cleanup at SSFL. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 167, citation 3 & 4: 

3 

Rockwell International, Radioactive Materials Disposal Facility Leach Field Environmental Evaluation

Report, DOE-SF-3, February 23, 1982, p. 8.

4 Collins, J., Final Radiological Inspection of the RMDF Leach Field for Release for Unrestricted Use,

N704TP990010, Rockwell International, August 13, 1980, p.11. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 167, citation 5:

5 The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning 

Engineering Evaluation/Cost Analysis, June 18, 2007, p. A-1. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 157, citations 3,4,5:

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Th RMHF Leach Field was used from early 1959 through late 1961 for the disposal of sanitary 

waste associated with Building 4021. Sanitary waste was subsequently accepted at the Area III 

sewage treatment plant. The RMHF septic system was abandoned in place. 

11.9.2 Former Radiological Burial or Disposal Locations 

The June 2009 Resource Conservation and Recovery Act Facility Investigation (RCRA RFI) 

report depicts a large potential debris area north and east of the RMHF Leach Field, and calls 

this area the Western Drainage Debris area. The report associates it with Building 4133; 

however the debris area appears to be closer to the RMHF Leach Field than Building 4133. The 

debris noted as concrete, scrap, metal, and empty drums may be associated with Building 4133, 

but the location of the debris area depicted is nearer the RMHF Leach Field.1 

The preliminary MARSSIM classification for the RMHF Leach Field area is Class 1, due to its 

former use and the radiological contamination found in the RMHF Leach Field. 

11.9.3 Leach Field Radiological Incident Report Summary, Data Provided by Boeing 

There have been a couple of incidents associated with Building 4654 that could have resulted in 

a release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0056 / 

A0064

11/3/1976 RMDF Leach Field Mixed Fission Contaminated leach field discovered 

during RMDF North Slope Recovery

A0079 2/14/1978 RMDF Leach Field Mixed Fission Leach Field during decontamination 

flooded with rain water.

• According to the 1982 RMDF Leach Field decontamination final report, prior to the connection 

of Building 4021 to the central sanitary sewer in late 1961, a valve to the emergency overflow 

of the radioactive water processing system was left partially open and allowed an unknown 

amount of contaminated water to enter the leach field system. Further investigation indicated 

that the distribution box flooded and leaked near its upper surface when the radioactive 

materials were introduced. The radioactive materials percolated northeasterly as well as down 

the north leach line and into the gravel bed.

Historical information regarding this release of contaminated water to the leach field could not 

be located by the EPA research team. According to a 1992 report on tritium release, the 

RMHF Leach Field became contaminated as a result of releasing approximately 5,000 gallons 

of water from the RMHF radioactive water system.

1 EPA/HGL, Final HSA 7-3-NBZ, p. 167, citation 2: 

2 Kartman, A.S., Aerial Photographic Analysis of Santa Susana Field Laboratory – Area IV, Ventura County , 

California, Volume 1 & 2, U.S. Environmental Protection Agency, Office of Research and Development, 

Environmental Sciences Division, March 2010. 

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It was assumed the major source of radioactive water at that time came from cleanup of the 

building 4020 (Hot Lab) hot cells after examination of fuel assemblies from the first Sodium 

Reactor Experiment (SRE) core or from wash water used at the SRE. The 1992 Rockwell 

report author estimates 0.1 curies (Ci) of Cs0137 and .017 Ci of tritium were released to the 

leach field, although the values are arbitrary as they try to account for losses to evaporation. 

According to the 1992 report, the RMHF Leach Field is considered a possible source of tritium 

in groundwater. 1

• On November 3, 1976 RMHF Leach Field connected to Building 4021 was found to be 

contaminated at levels up to 200 mrad/hr. To date the EPA research team has not located the 

incident report documenting this incident.

• On February 14, 1978 the RMHF Leach Field was found to have contaminated water draining 

from it as a result of heavy rains in February. These rains caused mud slides blocking some of 

the drainage paths around the leach field resulting in flooding of the field. This was 

compounded by flow of surface water directly into the gravel bed of the leach field through 

exploratory wells that had been drilled during 1976. The drainage water was reported to be 

contaminated with strontium-90 (Sr-90) at a level of 4 x 10 5 microcuries per milliliter (μCi/mL)

gross beta activity although other contaminants were possible. The point of drainage was 

dammed by use of sandbags and a sump was dug to retain additional drainage. Soil was 

removed from the surface of drainage channels downstream of the leach field to remove the 

contamination deposited by the runoff. A “large number” of water and soil samples were taken 

and analyzed from the point of drainage downstream to the “site boundary and beyond.” The 

results showed that surface water runoff samples ranged from a maximum of 4.6 x 10 5 μCi/

mL beta in the drainage from the leach field to 1.0 x 10 7 μCi/mL beta approximately 0.25 

miles past the Area IV boundary line. Soil radioactivity ranged from a maximum of 1,880 

picocuries per gram (pCi/g) beta at the leach field to normal background of approximately 24 

pCi/g beta near the Area IV boundary line. 

Based on estimates, the U.S. Environmental Protection Agency (EPA) finds it significant that 

the incident report indicated that the release from the leach field was estimated to not exceed 

3.2 millicuries (mCi). It was recommended the leach field be decontaminated in 1978. The 

incident report did not provide a figure showing the extent of contamination (A0056, A0064).2 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 160, citation 2: 

2 

Tuttle, R.J., Tritium Production and Release to Groundwater at SSFL, RI/RD92-186, Rockwell 

International, December 1, 1992, pgs. 4-10–4-11. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 161, citation 1:

1 Tuttle, R.J., Rockwell International Internal Correspondence, Re: Contaminated Runoff from RMDF Leach 

Field, March 2, 1978. 

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12.0 RMHF: Additional Locations / Operations 

Several important buildings and facilities associated with RMHF were left out of the 2006 Site 

Description and Site Profile. As a result, all associated data relevant to potential environmental 

releases and worker exposures associated with these locations has been similarly excluded. 

The RMHF was a major contributor to onsite ambient radiation exposure and airborne 

emissions. One of the most important facilities excluded from the 2006 Site Profile is the 

Building 4664 Low-Level Radioactive Waste Incinerator. According to Boeing, this facility was 

never used to incinerate radioactive waste. However, the EPA HSA research team located 

historical documents showing Building 4664’s construction for the purpose of burning low-level 

radioactive waste; that it functioned in such a capacity for nearly 20 years; and that upon its 

decommissioning it was disposed as radioactive waste. 

While the exclusion of many important buildings, operations, releases and incidents from a Site 

Profile results in a vastly diminished perception of site operations and worker exposure, as well 

as compromises the environmental dose estimations applied to workers on site, the exclusion of 

the Building 4664 Low-Level Radioactive Waste Incinerator alone has likely had a substantive 

impact on the validity of all environmental data submitted for SSFL. 

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Section 12.1 

Building Number:

Building Alias: 

Visitor Log Entry: 

4622 

622/T622

6017 / T022 / T034 / 6000E / 6001R 

Building Name: Radioactive Materials Handling Facility (RMHF)

Entry Point

Guard Shack 

Building Function: Health Physics Counting Area / RA Waste Counting 

Radionuclides of Concern: Given numerous references to Building 4622 as a building used 

for radioactive counting and the EPA research team’s inability to locate any more specific 

information on radionuclides of concern, it is presumed that waste handled at other RMHF 

complex buildings had the potential to be sampled and counted at Building 4622. Thus, this 

building area presents the same concerns for radionuclides that are reflected in the rest of the 

RMHF complex. 

12.1.1 Description of Operations & Processes 

Building 4622 appears to have been constructed in or before 1959, and has been identified as 

the Radioactive Waste Counting Building, and an Entry Building for the RMHF complex.1 A July 

3, 1962 Atomics International letter identifies Building 4622 as a “guard shack.” According to the 

letter, when the RMHF was activated, Building 4622 was equipped with a portal monitor and a 

hand-and-foot counter to check all personnel in and out of the RMHF area. However, 

background activity levels rendered the equipment useless and it was replaced with portable 

equipment better suited for monitoring personnel.2 Historical documents indicate Building 4622 

was considered an entry building, and also used as a storage location for visitor film badges and 

the control badge. It was reported on April 19th, 1965 that Building 4022 had been used for the 

recent open-air transfer of 20 Sodium Reactor Experiment (SRE) moderator cans in the vault of 

that building. As a result of the proximity of Building 4622 to Building 4022, background radiation 

in Building 4622 was raised from 005 “mr/hr” to approximately .2 “mr/hr” over a period of 

approximately 120 hours, resulting in elevated film badge readings for the month of April.3 

According to a 2005 HSA and a Resource Conservation and Recovery Act (RCRA) facility 

investigation work plan, Building 4622 was used as a health physics counting area. Samples of 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 122, citation 2:

2 Shoemaker, Operating Specification Radioactive Materials Disposal Unit, February 24, 1964. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 122, citation 5:

5 Schlapp, F.W., Atomics International Internal Letter, Subject: Radiation Monitoring Equipment in Dormant

Storage, July 3, 1962. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 122, citation 7:

7 Young, L.N., Atomics International Internal Letter, Subject: Overexposure of April Film Badges at Building 

622, April 19, 1965. 

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waste contained at the RMHF were taken to Building 4622 for radioactive counting. EPA’s 

research team could not locate historical documents to confirm these building operations. 

However, it should be noted that industrial planning maps for SSFL identify Building 4622 as a 

Radioactive Waste Counting Building from March 1962 through March 1975. The building is no 

longer present on industrial maps, beginning in July 1977. The 2005 RCRA HSA indicates that 

Building 4622 was demolished in approximately 1976. 

The preliminary MARSSIM classification for the Building 4622 area is Class 1, due to its 

possible use as a radioactive waste counting building and the limited information and 

investigation material available for this building. 

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Section 12.2 

Building Number:

Building Alias: 

Visitor Log Entry: 

4563 

563/T563

6017 / T022 / T034 / 6000E / 6001R 

Building Name: Radioactive Materials Handling Facility (RMHF)

Storage Yard 

Building Function: Waste Storage 

Radionuclides of Concern: The most probable radionuclides of concern are uranium, 

plutonium, thorium, and mixed fission products.1 

12.2.1 Description of Operations & Processes: 

Building 4563 was constructed in 1958, but at that time it was an uncovered, asphalt and 

concrete paved storage yard. It served as a storage area for containerized radioactive waste 

pending shipment to a disposal facility. According to the HSA, on average, fewer than 40 waste 

containers were stored in this yard pending shipment. No radiological incident reports were 

found.2 Use Authorizations dated 1970 indicate this location was used for the storage of 

irradiated stainless steel from the Systems Nuclear Auxiliary Power 8 Development Reactor 

(SNAP8-DR) shield assembly, at the the storage yard. 

The HSA indicates that the preliminary MARSSIM classification for the Building 4563 area is 

Class 1, due to its former use as a storage yard and the lack of previous investigations. 

The EPA HSA research team did not locate any radiological incident reports associated with this 

location. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 106 citation 2:

2 Cabrera Services, Final Combined Summary Report: Radioactive Materials Handling Facility Building

Surveys, Santa Susana Field Laboratory, Ventura County, California, October 2007, p. 4 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 103 

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Section 12.3 

Building Number: 4621 

Building Alias: 621/T621

Visitor Log Entry: 6017 / T022 / T034 / 6000E / 6001R 

Building Name Radioactive Materials Handling Facility (RMHF) 

Building Function: Radioactive Accountable Waste Storage Building 

Radionuclides of Concern: Building 4621 stored radioactive materials primarily in the form of 

mixed fission products from various waste sites.1 A radioactive waste packaging document 

indicated that a container stored at Building 4621 included radioisotopes Co-60, Cs-137, 

strontium-90 (Sr-90), and thorium-232 (Th-232). Internal Rockwell letters regarding use 

authorizations indicate the following sources have been stored in Building 4621: Am-241, Co60, Cs-137, iridium-192 (Ir-192), plutonium-beryllium (PuBe), and “Th-170” [sic, possibly Tm170].” A 1992 semi-annual leak test and a 1995 radioactive materials license renewal provided 

the following inventory of sources at Building 4621: Am-241, bismuth-210 (Bi-210), Cf-252, 

Cs-137, chromium-51 (Cr-51), cobalt isotopes (Co-56, Co-57, Co-58, and Co-60), Ir- 192, 

lead-210 (Pb-210), Kr-85, manganese isotopes (Mn-52, and Mn-54), Ra-226, Sr-90, and 

thulium-170 (Tm-170).5,6, Cf-252, Co- 56, Co-57, Co-58, Cr-51, Ir-192, Kr-85, Mn-52, Mn-54, 

and Tm-170. 

12.3.1 Description of Operations & Processes 

Building 4621 was constructed in the mid-1960’s as the Radioactive Accountable Waste Storage 

Building, designated for the storage of containerized low-to-intermediate level radioactive 

waste.2 Dry and liquid mixed waste materials were stored in 55-gallon drums. According to an 

Atomics International site waste management plan dated circa 1972, exhaust filter systems and 

additional contamination control safeguards were considered unnecessary for this facility, based 

on the assumption that stringent packaging requirements were sufficient.3 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 118, citation 6: 

. 6 The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning

Engineering Evaluation/Cost Analysis, June 18, 2007, p. A-2. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 109, citations 1, 2, 3: 

1 The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning

Engineering Evaluation/Cost Analysis, June 18, 2007, pgs. A-2 – A-3.

2 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992.

3 Correspondence from Seward, F.A. to Nagel, W.E., Re: Request for Radioactive Material and Radiation

Producing Devise User Authorization for RMDF Operation, December 10, 1985. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 93, citations 1&2:

1 

Atomics International, Site Waste Management Plan, Circa 1972, p. 2.

2 

Atomics International, Management of AEC-Generated Radioactive Wastes at Atomics International, May

31, 1974, pgs. 1-2. 

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According to the HSA, as of 2011 Building 4621 was still standing and the mixed-waste storage 

yard was still active.1 

The preliminary MARSSIM classification for the Building 4621 area is Class 1, due to its former 

use as a mixed waste storage area and the numerous spills of radiological material in the area. 

12.3.2 Building 4621 Radiological Incident Reports 

There have been several incidents associated with Building 4621 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

12.3.3 Building 4621 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0553 9/4/1975 621 Yard Ra226 During Tour of RMDF A Lost Source was 

Discovered

A0055 6/2/75 RMDF & Adjacent Mixed Fission Nine Spills Have Been Identified @ 

RMDF Complex

A0570 8/18/85 RMDF T-621 A False High Radiation Alarm Occurred 

And Was Responded To.

A0341 4/7/89 T621 RAS Response to False RAS Alarm

• On September 4, 1975, a 132 millicurie (mCi) radium-226 (Ra-226) source was discovered 

lying on the ground between Building 4621 and Building 4075. The source was not in a 

shielded container and was not labeled or tagged and had two feet of string attached. 

Following its discovery, the source was properly marked and stored in a secure condition. 

According to the incident report, personnel assigned to the RMHF had been instructed to 

clean out building 4621. Personnel removed all sources stored within the building to the open 

yard. While moving items, personnel dropped the spherical container and spilled the 

encapsulated source on the ground. The incident report did not indicate how long the source 

remained unnoticed. The incident report indicated that no significant exposure or spread of 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 112, citations 2 & 3: 

2 Group 7 – Northern Portion of Area IV RCRA Facility Investigation Report, Santa Susana Field Laboratory, 

Ventura County, California, Volume I – Text, Tables, and Figures, MWH, June 2009, p. 3-9.

3 

MWH, RCRA Facility Investigation Work Plan Addendum Amendment Radioactive Materials Handling 

Facility RFI Site (SWMU 7.6 and Area IV AOC) Santa Susana Field Laboratory, Ventura County, California, 

March 2008. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 111, citations 1 & 2: 

1 

Badger, F.H., Rockwell International Internal Letter, Subject: Lost Source Discovery, September 9, 1975.

2 Cabrera Services, Final Combined Summary Report: Radioactive Materials Handling Facility Building

Surveys, Santa Susana Field Laboratory, Ventura County, California, October 2007, p. 5. 

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contamination resulted from the incident (A0053).1

• An October 17, 1975, Rockwell letter discusses nine significant radioactive spill areas and 

numerous spot areas at the RMHF complex. Figure 2.7e presents the map of these areas, 

many of which were located in the western storage yard areas of the RMHF complex. The 

activity associated with the spills ranged from 0.5 to 3,500 millirads per hour. The spills were 

assumed to have occurred between 1972 and 1975 and to have followed the water course. 

One spill was a result of an empty high-level waste cask containing 15 rad per hour on the 

inside of the cask in the storage yard. The cask filled with rain water and flooded 

contaminated water out. Six similar casks were found. Additionally, three barrels of radioactive 

materials stored in “the backyard” tipped over and leaked or rotted out and leaked causing 

three other areas of contamination.As of October 17, 1975, all spills in the storage yard and 

two spills outside the storage yard were decontaminated or removed. Cleanup on another spill 

had begun and two other areas were identified for cleanup, but had not yet been delineated 

(A0055).2

• A November 18, 1975 Rockwell letter follows up on the October 17, 1975 letter discussed 

above and describes the current status of the RMHF complex. The letter notes that there were 

three contaminated areas on the floor and shelves in Building 4621. One area appeared to be 

a result of liquid seeping from a barrel liner stored on one of the upper shelves. The second 

area appeared to be the result of a powdery substance on the floor. A third area of 

contamination was of an unknown origin.

The November 18, 1975 letter also describes the holdup yard as containing significant 

radiological sources including approximately 250 drums (30 and 50 gallons), 59 high level 

pigs, 50 liquid storage drums (30 gallon), 32 drums of high level waste, 20 pallets of concrete 

block, 9 concrete conduits, 10 pallets of miscellaneous materials, 4 transfer casks, and 3 

pieces of contaminated Sodium Reactor Experiment (SRE) moderator handling equipment. 

Several areas of contaminated asphalt and approximately 35 contaminated pallets were noted 

in the holdup yard. The U.S. Environmental Protection Agency (EPA) finds it significant that 

the November 1975 letter notes that there were at least two separate major spills in the 

Building 4621 holdup yard, each covering approximately 100 square feet or more and several 

smaller spills of about 5 square feet or less.

Additionally, external to all buildings and the holdup yard proper, an accumulation of boxes, 

barrels, conduits, liquid tanks, large rectangular metal containers, trash dumpsters, and other 

miscellaneous materials was noted.

Several areas of contaminated asphalt and dirt were identified at the west end of the RMHF 

perimeter. The areas of contamination described in this letter appear to be the same areas 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 111, citations 4: 

4 Internal Correspondence from Badger, F.H. to McCurnin, Jr., W.R., Rockwell International, Re: Radioactive 

Spills in RMDF, October 17, 1975. 

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described above in the October 17, 1975 letter. Contamination of dirt and asphalt was thought 

to be the result of long-term, outdoor storage. Contaminated items were generally wrapped in 

plastic and stored in the storage yard, but after three or more years of storage and exposure 

to weather, the integrity of the plastic wrapping was compromised exposing contaminated 

surfaces. Rain water could then carry contamination onto outer surfaces of containers and 

along drainage pathways of the RMHF complex. Rows of 30-gallon drums stacked on the 

asphalt at the west end of the RMHF perimeter and pallets of non-contaminated material 

presumably became contaminated in this manner. According to the November 1975 letter, 

since mid-July all of the contaminated and non- contaminated items were removed from 

Building 4621 holdup yard. Items that could be decontaminated were and items that could not 

be decontaminated were packaged for disposal. Asphalt and dirt were jack hammered out and 

put into 50-gallon drums. Concrete blocks were moved to the SRE for use as shielding during 

decontamination and decommissioning operations, if necessary. Contaminated items within 

Building 4621were “decontaminated to as low as practical.” The letter recommends that funds 

be made available soon to dispose of all the contaminated items found at the RMHF 

complex.1 2 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 112, citations 1-4: 

1 Internal Correspondence from Harris, J.M. to McCurnin, W.R., Rockwell International, Re: Plan of Action 

for RMDF, November 18, 1975.

2 Internal Correspondence from Harris, J.M. to McCurnin, W.R., Rockwell International, Re: Plan of Action f 

for RMDF, November 18, 1975.

3 Internal Correspondence from Harris, J.M. to McCurnin, W.R., Rockwell International, Re: Plan of Action 

for RMDF, November 18, 1975.

4 Internal Correspondence from Harris, J.M. to McCurnin, W.R., Rockwell International, Re: Plan of Action 

for RMDF, November 18, 1975. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 113, citation 1: 

1 Internal Correspondence from Harris, J.M. to McCurnin, W.R., Rockwell International, Re: Plan of Action 

for RMDF, November 18, 1975. 

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Section 12.4 

Building Number:

Building Alias: 

Visitor Log Entry: 

4658 

658/T658

6017 / T022 / T034 / 6000E / 6001R 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Guard Shack / Entrance / Exit 

Building 4658 was originally constructed in 1959 for use as the Guard Shack. In 1981 or 1982, 

the entrance to the RMHF complex was reconfigured. New fencing, a new entrance gate, and a 

new guard shack were installed. 

12.4.1 Building 4658 Radionuclides of Concern 

Although no radionuclides of concern could be identified specific to Building 4658, the building 

was located at the entry point for all incoming and outgoing fuel and waste shipments. 

Therefore, EPA determined that radionuclides at other RMHF facilities should also be sampled 

at this location. EPA determined that sampling beneath the concrete foundation of Building 4658 

should be sampled because the area served as a storage location for radioactive materials, but 

had only received limited investigation. EPA determined the aboveground pipeline near Building 

4658 should be investigated because the contents of the pipeline were unknown, and pooling 

locations had been observed around pipeline supports. Surface drainage areas were also 

sampled, and investigated for residual contamination. 

According to the preliminary MARSSIM classification, Building 4658 area is a Class 1 due to its 

proximity to other buildings at the RMHF complex that handled radioactive material. 

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Section 12.5 

Building Number:

Building Alias: 

Visitor Log Entry: 

4663 

663/T663

6017 / T022 / T034 / 6000E / 6001R 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Storage Area 

Radionuclides of Concern: The most probable contaminants of concern associated with any 

possible radioactive material storage at Building 4663 were isotopes of uranium, plutonium, and 

thorium, and mixed fission products.1 

12.5.1 Description of Operations & Processes 

Building 4663 was constructed in the late 1950’s or early 1960’s as a Storage Area that may 

have been used for the storage of radioactive materials. A 1964 inspection tour of SSFL fuel 2

storage areas notes that “the accumulation of ‘loaded’ filters in storage building 663 is becoming 

a problem. These filters are being ‘loaded’ at a rate of up to 20/week. Each filter may contain a 

meaningful quantity of uranium when one is machining fuels containing fully enriched uranium.”3 

The Atomics International Reactor Fuels Committee discussed this accumulation of “loaded” 

ventilation system filters in Building 4663. According to the committee, these filters were being 

‘loaded’ at a rate of about 48 absolute filters per month (containing approximately 107 grams of 

uranium per absolute filter) in the “FEM and FED” machining operations. The Committee 

thought that the uranium content of these filters could present problems, both with regard to 

potential criticality (following certain types of accidents, e.g., fires), and overall accountability 

economics. However, according to the Committee, at the time, the RMHF had no way of 

recovering these fuel values from the filter media.4 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 134, citation 1: 

1 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. I-39. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 130, citations 4 & 5:

4 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. I-39.

5 

MWH, RCRA Facility Investigation Work Plan Addendum Amendment Radioactive Materials Handling 

Facility RFI Site (SWMU 7.6 and Area IV AOC) Santa Susana Field Laboratory, Ventura County, California, 

March 2008. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 130, citation 6:

6 Internal Correspondence from Matten, K.L. to Fuel Committee Members, Atomics International, Re: 

Inspection Tour of Santa Susana Fuel Storage Areas, April 23, 1964.

4 EPA/HGL, Final HSA 7-3-NBZ, p. 130, citation 7: 

7 Internal Correspondence from Reactor Fuels Committee to Balent, R. et al., Atomics International, Re: Tour 

and Inspection of AI Fuel Storage Facility – Minutes of Meeting of May 4, 1964, May 14, 1964. 

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According to the HSA, an Atomics International site waste management plan dated circa 1972 

and an updated plan dated May 1974 provided information regarding the waste disposal 

activities at SSFL. According to the waste management plan, which was prepared for the 

management of AEC contract wastes, provided the following summary: 

“Temporary storage of low-level radioactive waste is performed in Buildings 621, 663 and 075. 

Low-level waste and contaminated equipment are temporarily stored in Buildings 075, 621, and 

663, and in the fenced storage area surrounding Building 621. These buildings are steel 

buildings, which are designed to provide protection from the elements and security for the 

stored items. Contamination control is provided by stringent packaging requirements, so 

exhaust filter systems and additional contamination control safeguards are unnecessary.” 1 

The preliminary MARSSIM classification for the Building 4663 area is Class 1, due to its former 

use as a radioactive storage facility, a radiological incident and nearby spills, and its proximity to 

other facilities handling and storing radioactive waste. 

12.5.2 Building 4663 Radiological Incident Reports 

There has been one incident associated with Building 4663 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing. Summaries of all available incident reports are 

provided in Attachment A. 

12.5.3 Building 4663 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0616 5/21/67 RMDF Storage 

Yard

UC A Drum of U Metal Vented By Bullet 

Holes Caught Fire Upon Opening

• On May 21, 1967, a 55-gallon drum containing uranium metal under calcium carbonate 

powder was found burning in the RMHF storage yard. Violent venting through a hole in the 

drum indicated a likely explosion. Three rifle shots were fired into the drum to relieve the 

pressure, and the drum was then moved by forklift to a steel rack. The fire “extinguished itself” 

and the drum was left outside to cool overnight. The drum was then moved to Building 4021 

the following morning and was opened. Three of the 1-gallon cans within the drum began 

burning and the fire was extinguished with G-1 powder. The drum was left in Building 4021 

until final disposition. According to the report, air samples, nasal swabs and contamination 

surveyed indicated that no significant release of radioactive materials or contamination of 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 93, citations 1&2:

1 

Atomics International, Site Waste Management Plan, Circa 1972, p. 2.

2 

Atomics International, Management of AEC-Generated Radioactive Wastes at Atomics International, May

31, 1974, pgs. 1-2. 

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personnel occurred as a result of the fire (A0616).1

• An October 17, 1975, internal letter discusses nine significant radioactive spill areas and 

numerous spot areas identified at the RMDF complex. Figure 2.10b presents the map of these 

areas, most of which were located in the western storage yard areas of the RMDF complex. 

The activity associated with the spills ranged from 0.5 to 3,500 millirads per hour. The spills 

were assumed to have occurred between 1972 and 1975 and to have followed the water 

course. One spill was a result of a supposedly empty high level waste cask in the storage yard 

that filled with rain water and flooded contaminated water out. Six similar casks were found. 

Three barrels of radioactive materials stored in the backyard tipped over and leaked or rotted 

out and leaked causing three other areas. As of October 17, 1975, all spills in the storage yard 

and two spills outside the storage yard were decontaminated or removed. Cleanup on another 

spill had begun and two other areas were identified for cleanup, but had not yet been 

delineated.2 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 132, citation 1:

1 Unknown Author to Remley, M.E., Atomics International Internal Correspondence, Re: Radiation Safety

Unit Weekly Newsletter for Period Ending May 27, 1967, June 1, 1967. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 132, citation 2:

2 Internal Correspondence from Badger, F.H. to McCurnin, Jr., W.R., Rockwell International, Re: Radioactive 

Spills in RMDF, October 17, 1975. 

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Section 12.6 

Building Number: 4664 

Building Alias: 664/T664

Visitor Log Entry: 6017 / T022 / T034 / 6000E / 6001R 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Low-Level Radioactive Waste Incinerator 

Flocculation 

Evaporator

Low-Level Radioactive Waste Processing 

Radionuclides of Concern: The most probable contaminants of concern are isotopes of 

uranium, plutonium, and thorium, and mixed fission products. 

12.6.1 Description of Operations & Processes 

Building 4664 was constructed in 1964. Boeing indicated to EPA and in the existing SSFL Site 

Profile that this location was not used to incinerate radioactive waste. According to Boeing, the 

incinerator was a “test installation” that proved unsuccessful in the incineration of nonradioactive wastes, and it was dismantled. 

Atomics International’s “Site Plans and Details” documents generated in 1964 include drawings 

of the “Low-Level Radioactive Waste Facility” at Building 4664 and depict the low-level 

radioactive waste processing units and equipment on an uncovered equipment pad, which 

included the clarification unit, evaporator unit, incinerator unit, flocculation unit, exhaust stack, 

bag filter, cooling tower, and associated tanks, pumps, trench, and piping. 1

A September 1964 internal letter regarding the segregation of radioactive wastes indicates that 

the RMHF acquired incineration equipment to burn low-level and toxic contaminated dry 

wastes that were generated by programs at Atomics International. The letter also indicates 

that the equipment was to be operational after October 5, 1964, to burn contaminated wood, 

plastics, paper, rags, leather and other burnables containing less than 10% moisture, at a rate of 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 136, citations 5, 6, 7, 8:

5 

Atomics International Drawing, Low Level Radioactive Waste Facility, Site Plan & Details, 303-664-C1, 

Illegible Date. 
 6 

Atomics International Drawing, Modification for Low Level Radioactive Waste Equipment, Bldg. 664 –

Plans, Sections & Details, 303-664-S4, January 1966.

7 

Atomics International Drawing, Low Level Radioactive Waste Facility 664, Flow Diagram, 303-664-P1, 

Illegible Date.

8 

Atomics International Drawing, Low Level Radioactive Waste Facility, Flow Diagram & Details, 303-664-

M1, Date Illegible. 

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100 pounds per hour. 1

A 1967 test report for the Building 4664 incinerator notes that an acceptance test was 

successful in that cellulose waste was safely incinerated at the designed rate of 100 pounds per 

hour. Metallic objects were not reduced as successfully as paper products. No information 

regarding management of the incinerator ash was located. The test report also notes that the 

concrete trench, designed to receive any contaminated spill, is not waterproofed and seeped 

water at a slow rate. 

In 1982, Building 4664 incinerator and support equipment were removed. The incinerator was 

disposed as radioactive waste. The concrete pad that previously supported the 

flocculation tower was scabbled to reduce the contamination to less than 500 counts per 

minute. About 800 square feet of contaminated asphalt and soil was removed and disposed as 

radioactive waste. A September 24, 1982 letter noted that work continued to clean the concrete 

pad and the area between the pad and the fence.2 

Given the evidence of low-level radioactive waste incineration at the RMHF and its exclusion 

from the Site Profile, it is likely all data associated with airborne particulate emissions or other 

releases at this location have also been excluded. As a result, estimations of worker dose due to 

ambient and environmental exposures are called into question. 

Building 4664 was used as a processing facility for low-level radioactive waste at the RMHF, 

and as the former flocculation and incinerator area. The natural gas-powered evaporator was 

taken offline following a 1966 fire and the flocculation system was used for a period, until the 

evaporator could be replaced with an electric-powered evaporator. A 1989 report notes that the 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 137, citations 1-4: 

1 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. I-41.

2 

MWH, RCRA Facility Investigation Work Plan Addendum Amendment, Radioactive Materials Handling 

Facility RFI Site (SWMU 7.6 and Area IV AOC) Santa Susana Field Laboratory, Ventura County, California, 

March 2008. 

3 

Rockwell International, Nuclear Operations at Rockwell’s Santa Susana Field Laboratory – A Factual 

Perspective, N001ER000017, December 1989, Rev. C May 30, 1991, p. 29.

4 Internal Correspondence from Schlapp, F.W. to All P.E.s, R.E., and Supervision, Atomics International,

Re: Segregation of Radioactive (R/A) Wastes, September 18, 1964.

5 Ibid. 122119551 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 138, citations 5, 6, 7:

5 Internal Correspondence from Page, J.P. to Roberts, W.J., Rockwell International, Re: Engineering

Development & Text Highlites – Week Ending 9/24/82, September, 27, 1982. 

6 

MWH, Group 7 – Northern Portion of Area IV RCRA Facility Investigation Report, Santa Susana Field 

Laboratory, Ventura County, California, Volume II – RFI Site Reports, Appendix C, Radioactive Materials 

Handling Facility RFI Site (SWMU 7.6 and Area IV AOC), June 2009, p. C.2-12. 7 Internal Correspondence from McCurnin, W.R. to Walter, J.H., Re: Highlights – Week Ending 09-24-82, 

September 24, 1982. 

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flocculation tower was built to pretreat radioactively contaminated water to make it easier to 

filter, but it was made unnecessary by better filters and became inactive.1 

Based on drawings and reports, the flocculation system may have existed from late 1964 to 

early 1981, but it appears not to have been used for the entire time period. Atomics International 

site waste management plans dated in the mid-1970’s note the use of the flocculation tower and 

January 1979 incident reports (Incidents A0077 and A0232) note there was an overflow of the 

flocculation tower, resulting in contamination.2 

The preliminary MARSSIM classification for the Building 4664 area is Class 1, due to its former 

use as a low-level radioactive waste processing facility, radiological incidents, and lack of 

previous investigation. 

12.6.2 Building 4664 Radiological Incident Reports 

There have been three incidents associated with Building 4664 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. 

12.6.3 Building 4664 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0362 2/10/65 T664 Flocculation 

Tower

Mixed Fission R/A Liquid Processing System Tygon 

Hose Separated Spilling onto Pavement

A0489 5/13/65 Bldg. 22 Area Mixed Fission Flocculating Tower Overflowed Spilling 

R/A Water on Equipment, Pad and Soil

A0615 5/21/67 RMDF Truck UC A Drum of UC Sludge Exploded on a 

Truck Outside

A0077/

a0232

1/17/1979 Floc Tower Area Mixed Fission R/A Water from Flocculation Tower 

Contaminated Drainage Ditch and Pond.

• On February 10, 1965, the evaporator system backed up. While attempting to remove a 

stoppage by reverse flow provided by the circulating pump of the flocculation system, 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 16, citation 2:

2 Oldenkamp, R.D. and Mills, J.C., Nuclear Operations at Rockwell’s Santa Susana Field Laboratory – A 

Factual Perspective, N001ER000017, December 20, 1989, p. 29. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 16, citations 4,5,6,7:

4 Atomics International, Site Waste Management Plan, Circa 1972, p. 2.
 

5 Atomics International, Management of AEC-Generated Radioactive Wastes at Atomics International, May 

31, 1974, pgs. 1-2.

6 Gardner, F.W. Rockwell International Internal Correspondence, Re: Release of Radioactivity from RMDF – 

January 1979, February 23, 1979.

7 Gardner, F.W., Rockwell International Internal Letter, Subject: Radiological Safety Incident Report RMDF, 

February 26, 1979. 

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excessive pressure buildup occurred and blew the flexible hose from its connection, releasing 

approximately 5 gallons of radioactive contaminated liquid onto the asphalt. The asphalt was 

contaminated to a maximum level of 2,000 dpm/100cm2. Other components of the system 

were also contaminated to a maximum level of 2,000 dpm/100cm2. Decontamination of the 

area was performed with hydrochloric acid and a scrub brush (A0362).

• On May 13, 1965 while filling the flocculation tower with treated liquid for separating the 

sludge from liquid “suitable for processing in the evaporator system,” the operator’s attention 

was diverted and he failed to see the level indicator light on. The flocculation tower 

overflowed, spilling approximately 10 gallons of “chemically treated contaminated liquid” onto 

the tank and the tank equipment, the cement pad, and the surrounding soil. The incident 

report stated that smears indicated no contamination to the employee; however, the incident 

report did not indicate whether the surrounding area had been contaminated or whether the 

area was cleaned up (A0489).1

• On May 21, 1967, one of eighteen 55-gallon drums of uranium carbide sludge and other 

unspecified sludges exploded on a truck outside the RMHF incinerator, contaminating asphalt 

and evaporator equipment. The drum contained three 1-gallon cans of uranium carbide sludge 

that had been removed from the scrubber at Building 4005. Vermiculite and uranium 

contamination was blown into the evaporator and incinerator area, areas of asphalt, and over 

the north boundary fence. Contamination levels ranged from 300 to 5,000 dpm/100 cm2. Initial 

cleanup of the area was performed with a vacuum cleaner and the evaporator and incinerator 

equipment was scrubbed down. The area was then “hosed down.” The truck was 

decontaminated in the decontamination room and the drums were placed in the packaging 

room where the six drums containing sludge material were kept until final disposition. The 

other twelve drums containing uranium carbide were decontaminated and shipped to Building 

4064 for storage. The incident report did not include additional information regarding the 

decontamination procedures or the level to which the area was decontaminated. While the 

report indicated that area was “hosed down,” additional information was not provided to 

indicate whether the area was still ontaminated or whether the runoff was collected or allowed 

to enter the surrounding drainage channels (A0615).2

• A November 18, 1975 Rockwell letter describes an “area of concern” at the Building 4664 pad. 

The letter nots the equipment pad contained the “old evaporator, incinerator, and oxidation 

equipment.” Although activity from the evaporator was sufficient to prevent the detection of 

contamination near the area, the letter states that preliminary indications show the ditch 

outside the fence is contaminated to the west of Building 4664. No other details are provided 

in the letter to indicate what may have caused the contamination, how or when it occurred, or 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 139, citation 2:

2 Young, L.M., Atomics International Internal Correspondence, Re: Incident Report Building 022, May 13,

1965, May 18, 1965. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 140, citation 1:

1 Unknown Author to Remley, M.E., Atomics International Internal Correspondence, Re: Radiation Safety

Unit Weekly Newsletter for Period Ending May 27, 1967, June 1, 1967. 

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what further steps were taken to characterize and remediate the contaminated area.1

• On January 17, 1979, radioactive water from the Building 4664 flocculation tower was leaked 

onto paving and rinsed into the RMHF 4614 Holdup Pond contaminating the drainage ditch 

and pond. It was estimated that approximately 2 gallons of contaminated water containing 

approximately 5 x 10 2 microcuries per milliliter (μCi/mL) of mixed fission products leaked from 

the tank. The incident report indicated that initially the flocculation tower pad and runoff ditch 

contained contamination levels of 25,000 counts per minute (cpm) with hot spots of 35,000 

cpm. There was no information about any cleanup or removal actions. However, an asphalt 

sealant was applied on January 31, 1979, and levels reduced to 20 to 50 percent of the 

original values with an additional reduction of approximately 20 percent following the “runoff of 

the sealant.” (A0077, A0232). 2

• On October 23, 1979, approximately 1 gallon of radioactive liquid representing waste from 

both the Sodium Reactor Experiment and Hot Lab was released into a catch basin under the 

“Building 4021” flocculation tower. The catch basin was “cleaned to about the same 

contamination level as usual,” which was less than 1,000 dpm/100 cm2 of beta- gamma 

activity. The incident report noted it was not practical to remove more activity at the time due 

to the deteriorated condition of the metal floor.3 

The EPA HSA provides a detailed chronology of radiological investigations and decontamination 

and cleanup of releases associated with Building 4664. No Radiological Use Authorizations 

were found.

1 EPA/HGL, Final HSA 7-3-NBZ, p. 140, citation 2:

2 Internal Correspondence from Harris, J.M. to McCurnin, W.R., Rockwell International, Re: Plan of Action 

for RMDF, November 18, 1975. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 140, citations 3 & 4:

3 Gardner, F.W. Rockwell International Internal Correspondence, Re: Release of Radioactivity from RMDF –

January 1979, February 23, 1979.

4 Gardner, F.W., Rockwell International Internal Letter, Subject: Radiological Safety Incident Report RMDF, 

February 26, 1979. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 140, citations 5:

5 Internal Correspondence from Owens, D.E. to Radiation & Nuclear Safety Energy Systems Group,

Rockwell International, Re: Radiological Safety Incident Report, RMDF, October 23, 1979, dated October 29, 

1979. 

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Section 12.7 

Building Number: 4668 

Building Alias: 668/T668

Visitor Log Entry: 6017 / T022 / T034 / 6000E / 6001R 

Building Name: Radioactive Materials Handling Facility (RMHF) 

Building Function: Storage & Sodium Cleaning 

Radionuclides of Concern: Although designated a “non-radioactive” chemical storage facility, 

radioactive material was apparently stored under the Building 4688 structure. The most 

probable radionuclides of concern are isotopes of uranium, plutonium, and thorium, and mixed 

fission products.1 

12.7.1 Description of Operations & Processes 

Building 4688 was originally constructed northeast of Building 4003 (located in the Sodium 

Reactor Experiment [SRE] complex area) in approximately 1962. It was moved to the RMHF in 

approximately 1967 to serve as a storage structure.2 

Building 4688 was used for storage and likely supported sodium cleaning activities. The building 

was used primarily as protection against the rain and sun. A 1993 report states that Building 

4688 was a hazardous materials storage shed that stored chemicals in small containers of one 

gallon or less, including acetone, alcohols, solvents, laboratory reagents, paints, oils and 

gasoline. A large storage container with unknown contents is visible in the Building 4688 area in 

historical photographs from 1961-1989. 

A November 18, 1975 Rockwell letter describes an “interim storage yard” between the RMHF 

complex and the SRE, and reflects areas of contamination on the ground. The letter notes the 

area must be cleared of brush and pallets of material must be moved to further delineate spill 

areas.3 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 154, citation 2:

2 Kartman, A.S., Aerial Photographic Analysis of Santa Susana Field Laboratory – Area IV, Ventura County , 

California, Volume 1 & 2, U.S. Environmental Protection Agency, Office of Research and Development, 

Environmental Sciences Division, March 2010. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 152, citations 1 & 2:

1 The Boeing Company, Radioactive Materials Handling Facility Decontamination and Decommissioning

Engineering Evaluation/Cost Analysis, June 18, 2007, p. A-6.

2 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962– 

November 1992. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 153, citations 2 & 3:

2 Internal Correspondence from Harris, J.M. to McCurnin, W.R., Rockwell International, Re: Plan of Action 

for RMDF, November 18, 1975. 

3 Internal Correspondence from Badger, F.H. to McCurnin, Jr., W.R., Rockwell International, Re: Radioactive 

Spills in RMDF, October 17, 1975. 

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In 2007, Boeing performed a radiological survey on Building 4688 and other buildings within the 

RMHF complex. Some survey data was collected from Building 4688 roof surfaces, and spot 

checks of the building revealed significant levels of beta-emitting residual radioactivity on 

several of the support columns. Boeing directed the survey activities be discontinued. Limited 

investigation of the building roof concluded that additional survey data would be needed to 

determine suitability for release.1 

Indications that waste had been dumped near and around building 4688 were identified in 

February 2008.2 In 2009, EPA’s site visit noted that the Building 4688 area was then used for 

equipment, hazardous waste, and chemical storage. Three locked storage cabinets located 

under the Building 4688 awning stored hazardous waste and chemicals, and were inspected 

weekly. 

The preliminary MARSSIM classification for the Building 4688 area is Class 1, due to the drum 

storage in the area and Building 4688’s proximity to RMHF Building 4022, which stored 

radioactive materials, and limited previous investigations. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 153, citation 4:

4 Cabrera Services, Final Combined Summary Report: Radioactive Materials Handling Facility Building

Surveys, October 2007, pgs. 19, 40. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 154, citation 1:

1 

MWH, RCRA Facility Investigation Work Plan Addendum Amendment, Radioactive Materials Handling 

Facility RFI Site (SWMU 7.6 and Area IV AOC), Santa Susana Field Laboratory, Ventura County, California, 

March 2008, p. 2-4. 

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13.0 Research on Reprocessing Used Reactor Fuel 

The 2006 Site Description excluded operations and processes associated with facilities used to 

research and reprocess used reactor fuel. 

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Section 13.1 

Building Number:

Building Alias: 

4003 

003/T003 

Building Name: Engineering Test Building (ETB) “Hot Cave” 

Building Function: Fuel Element Assembly (SRE)

Molten Uranium Research

Fuel Decladding

Irradiation Experiments

SNAP Fuel Analysis & Corrosion Tests

Other Related Buildings: Substation Building 4693

Support Building 4825 

Notes: Tritium 

Radionuclides of Concern: Uranium, thorium, transuranic elements, mixed fission products, 

and activation products. Possible radionuclides include natural and enriched uranium (U-238, 

U-234, U-235, U-236), isotopes of plutonium (Pu-239, Pu-240, Pu-241, Pu-242), americium-241 

(Am-241), thorium-228 (Th-228), Th-232, Th-234, activation products (tritium (H-3), carbon -14 

(C-14), sodium-22 (Na-22), Na-24, chromium-51 (Cr-51), manganese-54 (Mn-54), nickel-59 

(Ni-59), Ni-63, iron-59 (Fe-59), cobalt-60 (Co-60)) and fission products (krypton-85 (Kr-85), 

strontium-89 (Sr-89), Sr-90, antimony-125 (Sb-125), iodine-129 (I-129), I-131, cesium-134 

(Cs-134), Cs-137, cerium-144 (Ce-144), barium (lanthanum)-140 (Ba (La)-140), niobium-95 

(Nb-95), ruthenium-103 (Ru-103), Ru-106, xenon-133 (Xn-133), Xe-135, promethium-147 

(Pm-147), samarium-151 (Sm-151), europium-152 (Eu-152), radium-226 (Ra-226),

actinium-228 (Ac- 228). 

13.1.1 Description of Operations & Processes 

Building 4003 was constructed in 1957.1 It was known as the Engineering Test Building (ETB) 

and the “Hot Cave.” Between 1957-1964 it was used to assemble fuel elements for the SRE. In 

this operation, uranium and thorium metal slugs were loaded into metal tubes, the remaining 

tube space was filled with sodium and the tubes were sealed. The Hot Cave portion of the 

building was designed and constructed to investigate the chemistry of molten uranium, and to 

study the separation of fission products and plutonium from uranium systems. 

The HSA provides aerial photographs and a detailed description of building features and 

specifications. Building 4003 contained offices, a change room, several laboratories, a Freon 

1 EPA/HGL, Final HSA 6, p.12, citation 1: 

1

ETEC website at: www.etec.energy.gov/History/Major-Operations/SRE.html 

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compressor, a concrete pit, a tank pit, a sump, an expansion tank, a boiler, and a cooling tower.1 

It should be noted that because Building 4003 contained offices, employees that may have been 

assigned to the offices may have been considered “non-radiological” employees. However, 

given the operations at the facility, such a designation may grossly misrepresent their exposure 

risks. 

Building 4003 also contained an exhaust system that included nine fume hoods, valves, ducts, 

blowers, filters, filter plenum chambers, and a stack. In addition, Building 4003 contained two 

radioactive waste sinks connected to 5-gallon bottles, and a highly shielded area designed for 

remote manipulation of radioactive materials, known as the “Hot Cave.” The Hot Cave contained 

identical east and west “hot cells,” a pair of mechanical manipulators for each of the two cells, a 

transfer tunnel, liquid waste lines, east and west cell “small” radioactive liquid waste holdup 

tanks located below the Hot Cave, and a hot cell exhaust system with HEPA filters and two 

blowers located on the roof of Building 4003. 

The west hot cell of the Hot Cave was used primarily for kilogram-scale processing experiments 

involving chemical reactions of irradiated reactor fuels at temperatures up to 1,800 ̊ C. The east 

hot cell was used primarily for mechanical operations, such as decladding, component 

disassembly, and inspection. The transfer tunnel was used for sample handling, waste removal, 

inter-cave transfer, and the storage of kilocurie sources.2 According to DOE, later activities 

involved bench-scale research into reprocessing of used (or irradiated) fuel assemblies. The 

research involved the removal of fission products from used fuel.3

Building 4003 was also used for the analysis of Systems for Nuclear Auxiliary Power (SNAP) 

fuel burn-up samples and the evaluation of irradiation experiments. Use of the Hot Cave ended 

when the SNAP program was terminated in 1973. 

The 2006 Site Profile indicates that “some traces of radioactivity were later found in the drain 

line, which was removed.” However, according to Atomics International (1973), the inner 

surfaces of the Hot Cave were found to be “grossly” contaminated with mixed fission products. 

According to the HSA, containment of this high-level contamination required continuous 

operation of the radioactive exhaust system as well as filtering and sampling activities. 

Radioactive contamination was found to be on the internal surfaces of the hot cells, in the liquid 

waste lines, the liquid waste holdup tanks, the radioactive exhaust system ducting and on filter 

1 EPA/HGL, Final HSA 6, p.12, citations 3 & 4: 

3 Montgomery Watson Harza, Table 1, List of Historical Document Map Features at SRE, August 1, 2003.

4 Rockwell internal letter from C. D. Bingham to W. F. Heine, Re: Application to Perform Radiochemical 

Operations in Building 003A (ETB Annex), December 10, 1970. 

2 EPA/HGL, Final HSA 6, p.13, citation 10: 

10 Strausberg, S., Gardner, W. J., Guon, J., Luebben, T. E., and Mills, T. H., Modified Hot Cave Facility for

Reprocessing Experiments, Atomics International Report No. NAA-SR-2687, June 30, 1958, p.3. 

3 EPA/HGL, Final HSA 6, p.14, citation 1: 

1 ETEC website at: www.etec.energy.gov/History/Major-Operations/SRE.html. 

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plenums. In 1974, the building 4003 facilities were declared “excess” and a facilities dismantling 

plan was prepared.1 

Atomics International informed the AEC that the alpha-emitting radionuclides present were from 

enriched uranium; some transuranics may also have been present. The beta-gamma emitting 

radionuclides resulted from fission products and stainless steel activation products.2 Atomics 

International commenced D&D in January 1975 and ended in June 1975. Dismantling of the Hot 

Cave was completed in April 1975. 

After 1961, Building 4003 was connected to a site-wide (Areas I, II, III, IV) sewage treatment 

system. In 1982, the sewer lines were found to be radioactively contaminated.3 The sewage 

treatment plant began in the 1950’s to collect sanitary waste from Areas II, III, and IV. In addition 

to sanitary waste, the treatment plant received cooling tower discharges from non-chromated 

cooling tower systems and treated groundwater from various site groundwater recovery 

systems. The treated discharge was allowed to flow into the R-2A Discharge Pond in Area II. As 

part of the Site Wide Water Reclaim System, the Area II R-2A Pond serviced the Coca Rocket 

Engine Test Stands and Saturn V personnel. In July, 1982 Rockwell reported enriched uranium 

was identified in the sewer sump outside Building 4003 in the range of 23-32 μg/g. Rockwell 

concluded that ANL would object to the unrestricted use of Building 4003 but asserted that 

cleanup was impractical and almost impossible.4 

The HSA provides a detailed chronology of radiological investigations and D&D activities 

associated with the Building 4003 area. The preliminary MARSSIM classification for the Building 

4003 area is Class 1, due to its former use as an SRE support building and its proximity to SRE 

Building 4143. 

13.1.2 Building 4003 Radiological Incident Reports 

There have been several incidents associated with Building 4003 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

1 EPA/HGL, Final HSA 6, p.14, citation 2: 

2 Graves, A. W., Facilities Dismantling Plan for Building (D+D) T003 Hot Cave, Atomics International Report 

No. FDP-704-990-001, October 10, 1974, p. 2. 

2 EPA/HGL, Final HSA 6, p.14, citation 3: 

3 Atomics International letter from W. F. Heine to R. L. Westby, U.S. Atomic Energy Commission, re: 

Contamination Limits for Release of KEWB and Building 003 for Unrestricted Use, November 21, 1974. 

3 EPA/HGL, Final HSA 6, p.22, citation 1: 

1 Letter from B. D. Sujata, The Boeing Company, to J. Evans, County of Ventura, re: Information Regarding

Permit – Septic Tank and Leach Field, October 23, 2001. 

4 EPA/HGL, Final HSA 6, p.14, citation 4: 

4 Rockwell International Telephone Conversation Record from W. Smith to B. F. Ureda, Re: Building 003

Contamination, July 2, 1982. 

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13.1.3 Building 4003 Radiological Incident Report Summary - Data Provided by Boeing 

*Isotopes are written as they are presented in the incident database. The research team believes that MFP is an 

acronym for mixed fission products. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0530 10/7/1959 ETB Maybe Employee contaminated his hands when 

he failed to wear prescribed protective 

clothing.

A0423 12/22/1959 ETB MFP Contractor removed R/A exhaust stack 

without health physics cognizance.

A0549 7/22/1960 ETB Rm 160 Employee cut hand in controlled area.

A0295 2/1/1965 ETB & D004 Varied work location & late return of film 

badge resulted in exceeding guidelines.

Employee exposure: 2,000 mrem gamma 

and 640 mrem beta

A0438 9/15/1969 ETB Hot Cell MFP Multiple projects resulted in employee 

exposure above guidelines.

Employee exposure: 6,065 mrem

A0048 9/2/1969 ETB Chem Labs ETB Lab equipment and floor found 

contaminated.

• On October 7, 1959, an employee performed maintenance on a radioactive contaminated 

sodium line with his bare hands. The employee was aware the line was contaminated. Both of 

his hands became contaminated to a level of 1.5 mrad/h. After 35 minutes of cleaning, the 

employee’s hands were successfully decontaminated (A0530). 1

• On December 22, 1959 contract personnel were working on the exhaust stack of the 

radioactive vent systems at the ETB without the knowledge of the health physics department, 

without the proper permit, and without film badges. The contract crew was unaware the 

exhaust system was radioactive. No personnel contamination was discovered. Work was 

immediately stopped until the proper permit was filled out and film badges assigned (A0423).2

• On July 22, 1960 an employee cut his hand during a drilling operation. No activity was found 

above background on his hand (A0549).3

1 EPA/HGL, Final HSA 6, p.16, citation 2: 

2 Bell, C. E., Internal Letter, re: Violation of Health Phys. Practices, October 8, 1959. 

2 EPA/HGL, Final HSA 6, p.16, citation 3: 

3 Lang, J. C., Atomics International Internal Letter, re: Notice of Rule Infraction, December 31, 1959. 

3 EPA/HGL, Final HSA 6, p.17, citation 1: 

1 Illegible Author, Atomics International Internal Letter, re: Radiological Safety Incident Report, Room 160 

ETB Annex, July 22, 1960, July 27, 1960. 

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• In October 1965, a missing film badge belonging to a chemist was found. The badge had been 

missing since January 1965 and when processed, indicated an exposure 2,000 mrem gamma 

and 640 mrem beta. When added to the chemist’s February and March exposures, the 

quarterly total exposure was 2,600 mrem gamma and 1,870 mrem beta. Review of the 

chemist’s duties in January suggested that he did in fact receive the exposure indicated on the 

film badge, despite the possibility of other factors affecting the reading. The chemist’s lifetime 

accumulated exposure through September 1965 was 28.4 rem, below the permissible 

exposure of 65 rem at that time (A0295).1

• On September 15, 1969 film badges worn by an employee over the course of a calendar 

quarter were evaluated and the combined exposure from the badges was 6,065 mrem. This 

overexposure was found to be the result of multiple different operations at the ETB (A0438). 2

• On September 2, 1969 a smear survey was requested for the upstairs laboratory at the ETB 

because contamination had been found on the floor. The contamination was first noted on an 

employee’s shoe (200 cpm). The employee smeared the stairs and found up areas of 

contamination up to 1,000 dpm/100 cm2. This prompted him to smear the floor where he was 

setting up equipment. Repetitive smears of the floor brought forth “hotter” and “hotter” results. 

The employee eventually roped off the area and requested further surveying of the lab. 

Additional smear surveys found gross contamination on the bench top and covered sink. A

couple of spots read up to 1.5 rad/h. This material was easily removable. Parts of two ring 

stands on the work bench indicated 0.5 rad/h and were bagged up. Another “hot” area of the 

lab was located on an adjacent bench top where a survey detected 30,000 cpm. Smear 

surveys taken on September 4 and 5 1969 found contamination greater than 150,000 dpm/

100 cm2 (the counter detectibility limit) on the outside of a furnace, a balance, and a hot plate 

in the hood. The hot plate was measured to be 330 mrad/h. Also found in the hood, behind two 

lead bricks, was a bottle of liquid waste from earlier equipment, which read 5 rad/h. According 

to the incident report, the “episode was somewhat mysterious.” It appears that there may have 

been two sources of contamination because gamma activity was missing from one set of 

smears. It was unknown whether a chemical or physical process had removed gamma 

emitters, such as cesium, from this area. As of September 8, 1969, the lab had been fairly well 

decontaminated and most items outside the hood were less than 100 dpm / 100 cm2 betagamma (A0048).3

As part of the dismantling operations, Atomics International opened the west cell and conducted 

a radioactivity survey; the cell was found to contain significantly higher levels of radioactivity 

than the east cell. The west cell contained test analysis equipment and experimental residue. 

Miscellaneous waste was removed including three trays (7 rad/h) and two 1-gallon paint cans 

1 EPA/HGL, Final HSA 6, p.16, citation 2: 

2 Correspondence from Remley, M. E., Atomics International, to Levy, J., U.S. Atomic Energy Commission, 

Re: Apparent Type B Radiation Exposure, November 5, 1965. 

2 EPA/HGL, Final HSA 6, p.16, citation 3: 

3 Bresson, J. F., North American Aviation Internal Letter, re: Film Badge Exposure – Contamination at the 

ETC, SS003, Bresson to Heine, 9/13/69, October 15, 1969. 

3 EPA/HGL, Final HSA 6, p.16, citation 4: 

4 Bresson, J. F., North American Rockwell Corporation Internal Letter, re: Contamination at the ETC, S003, 

September 15, 1969. 

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(approximately 25 rad/h). These items were sent to the RMHF (Radioactive Materials Handling 

Facility) for disposal. Five prefilters (250 mrad/h) from the lower section of the cell and a 30-

gallon bag of solid waste (2 rad/h) were then removed and the cell was vacuum cleaned again 

to remove most loose contamination. The roof shield blocks and associated manipulators were 

removed while the openings in the roof were covered with plastic sheeting to contain loose 

contamination. The remaining items in the cell were then removed. These included a Lucite 

enclosure, a shelf, a table, and 90 gallons of solid radioactive waste. Dry uranyl salt solution (25 

rad/h at 10 cm) were removed from the floor. A container of SNAP burnup samples was located 

in the transfer tunnel between two cells. Using tongs, this container was transferred to a 5-

gallon can and then into a lead cask, which was then transferred to RMHF.1 

The HSA provides a detailed chronology of radiological investigations and decontamination / 

cleanup procedures, including Atomics International’s excavation of the Hot Cave. 

1 EPA/HGL, Final HSA 6, p.17, citation 3: 

3 Ureda, B. F., Building 003Decontamination and Disposition Final Report, AI-ERDA-13158, February 25, 

1976, pp. 13-15. 

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14.0 Operation of Particle Accelerators 

Two particle accelerators were used at SSFL; one in Building 4030 (described in this section) 

and one in Building 4009 (excluded from the 2006 Site Description. Please see Section 7.1 

pertaining to Building 4009 and the OMR/SGR Reactors). 

There are other ways to generate artificial radioactivity besides nuclear fission. One way is to 

bombard a target material with atomic particles accelerated to high speeds by means of a 

particle accelerator. A common form of particle accelerator is a Van de Graaf generator, which 

uses a high-voltage electrostatic field to accelerate atomic particles to high speeds (high energy 

levels). Collisions of these particles with a target material (such as aluminum or tritium) can 

generate small amounts of radioactivity. Atomics International operated a Van de Graaf 

generator in Building 4030, bombarding tritium targets with deuterons to produce neutrons. 

A second Van de Graaf generator was operated in Building 4009 (OMR/SGR). It should be 

noted that, while the 2006 SSFL Site Description references a second particle accelerator at 

SSFL, it briefly associates it with the Sodium Reactor Experiment (SRE, Building 4143) and the 

general time period of the SRE’s decontamination / decommissioning, rather than its actual 

location at the OMR/SGR in Building 4009, and time period of operation. 

Additionally, the 2006 Site Description does not mention the known tritium contamination 

associated with the Van de Graaf accelerator (Buildings 4009 or 4030), or the conversion of 

Building 4030 to offices before the Van de Graaf was removed and the tritium contamination 

discovered. As a result of being converted to administrative offices, Building 4030 may be 

considered a “non-radiological” location during the years after the Van de Graaf’s operations 

and job processes associated with this location may be considered low-risk. Workers assigned 

to this location may not have been adequately monitored for radiation exposure, given the 

facility’s operations and potential for radioactive contamination.

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Section 14.1 

Building Number:

Building Alias: 

4030 

030/T030 

Building Name: Van de Graaf Particle Accelerator

a.k.a. AE-6 Counting Room

a.k.a. AE-6 Work Shop

a.k.a. AE-6 Office Annex 

Building Function: SNAP Program Support

Radioactivity Generation

Outside Storage of Drums Containing Mixed Fission Product

Site Purchasing Office, Traffic, Warehousing 

Other Buildings Involved: 4035 (035/T035) 

Notes: DOE-NASA

Tritium Use. 

Radionuclides of Concern: Regulated radiological materials were managed at Building 4030. 

The Van de Graaf accelerator produced neutrons by H3(p,n)He3 nuclear reaction. The 

potential contaminant of concern is tritium. Activation of building materials was negligible 

because drums of borated water were used around the target to thermalize and capture 

neutrons. 

14.1.1 Description of Operations & Processes: 

Building 4030 was constructed in 1958. It included two connected sections. The front section, 

was originally known as Building 4035 before the rear section was added. Combined, the two 

buildings formed Building 4030.1 

A Van de Graff accelerator was moved into the facility in 1960 and operated through 1964 in 

support of the SNAP program. The accelerator provided a proton beam of up to tens of 

microamperes in current, with continously adjustable energies from a few hundred KeV up to a 

maximum of about 1 MeV. The Van de Graaf accelerator provided an adjustable energy proton 

beam to bombard a tritium target to produce neutrons. The building had an associated leach 

field likely used until 1961 or 1962, when the building was connected to the newly-built Area III 

site-wide sewage system. The Van de Graff accelerator was reportedly mothballed and 

remained in the building until 1966. A 1966 smear survey of the accelerator detected significant 

tritium contamination. To date, the disposition of the accelerator has not been determined. 

1 EPA/HGL, Final HSA 5-A, p. 116, citation 498: 

498 Rocketdyne Report, 030-AR-0002, “Decontamination and Decommissioning (D&D) of Building T030,” 

November 13, 1997. Ref. 122119552 

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In 1965 (a year before the tritium contamination was discovered), Building 4030 was converted 

to an office. In 1972, it was used as a purchasing office for the site, and for traffic and 

warehousing. This location was used as an office until 1995.1 

On March 29, 1966, personnel conducted a tritium smear on Building 4030 and its associated 

equipment. The tritium contamination was severe. Two of the samples contained sufficient 

contamination to saturate the gas proportional counter utilized in measuring tritium 

contamination. According to the report, the activity of the samples was “estimated” by covering 

each sample with a clean piece of sample paper with a 1/8 inch diameter hole at its center. As a 

result, the tritium beta particles from the sample area outside the hole were shielded and only a 

1/8 inch diameter area of the sample was counting. This produced a count rate of approximately 

75,000 dpm. As recorded, assuming an equal distribution of the contamination over the total 

area of the sample, a ratio of count rate to the area was established and the total sample count 

rate was calculated by extrapolating to the total area of the 1-inch diameter sample. The 

following are the result of five sample locations with the most severe estimated contamination. 

1966 Building 4030 Tritium Contamination Levels 

LOCATION ACTIVITY (dpm/100 cm

Plate below target holder ~1x105

Tritium target storage can ~1x105

Lower cooling hose ~1x105

Target holder (sides and front) ~3.6x10

Target holder (rear) ~3.6x10

The EPA provides a detailed chronology of radiological investigations for this location. There are 

discrepancies between the 1988 findings by Rocketdyne and the 1995 findings by ORISE. In 

1995, ORISE found that one sample of total tritium activity exceeded the average guideline for 

beta-gamma emitters (6,600 dpm/100 cm2 2 ), and recommended additional sampling. In a March 

1 EPA/HGL, Final HSA 5-A, p. 117, citation 507-509: 

507 ETEC Document, GEN-ZR-0007, “Radiological Survey of Shipping /Receiving and Old Accelerator AreaBuildings T641 and T030,” August 19, 1988.

508 Rocketdyne Report, 030-AR-0002, “Decontamination and Decommissioning (D&D) of Building T030,” 

November 13, 1997.

509 Mooers, A.R., Atomics International Internal Letter Re: Tritium Smear Survey, Building 030 Van de Graaff 

Accelerator, March 29, 1966. 

2 EPA/HGL, Final HSA 5-A, p. 120, citation 519: 

519 ORISE Document 96/C-4, “Verification Survey of the Interim Storage Facility; Buildings T030, T641, and 

T013; an Area Northwest of Buildings T019, T013, T012, and T059; and a Storage Yard West of Buildings 

T626 and T038, SSFL, Rockwell International, Ventura County, California,” Vitkus, T. J., and T. L. Bright, 

February 1996. 

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1996 letter, the release of the building was put on hold as a result of DOE’s recommendation 

that Rocketdyne resurvey the building.1 

The preliminary MARSSIM Classification for Building 4030 is Class 1 based on previous site 

activities.

1 EPA/HGL, Final HSA 5-A, p. 120, citation 521: 

521 Montes, Michael, Letter Re: Buildings 030, 019, and 654, March 29, 1996. 

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15.0 Research Using Radioisotopes 

Buildings 4064 / 4023 

The 2006 Site Description described research at SSFL that required the use of special 

radioisotopes, and specified that “only small quantities of specially-prepared radioisotopes were 

brought to the SSFL for use in laboratories under carefully controlled conditions.” The 2006 Site 

Description provides the requirements of special radioisotopes used for the TRUMP-S program 

as an example: 5 grams of plutonium, 4 grams of neptunium, and 4 grams of americium (which 

were apparently shipped to Missouri). 

The 2006 Site Profile neglected to describe Building 4064 as a 4,418 square foot structure, 

equipped with a roll-up door, 11” thick walls, surrounded by an 8’ fence topped with a barbed 

tape (double later) and an intrusion alarm, which was constructed specifically to store special 

nuclear materials, fissionable materials, in addition to radioactive waste. 

According to the 2006 Site Description, Building 4023 operated between 1976-1983 in the 

performance of small sodium loop tests. The building’s operations between 1962-1976 are not 

described, nor are operations that included TRUMP-S support (which involved transuranic 

products), nor are building operations that began in 1987 (Rocky Flats Plutonium Recovery 

Project). 

It is apparent that the 2006 Site Profile provides an insufficient description of Buildings 4064 and 

4023, used for research purposes involving radioisotopes. In addition, there are no incidents at 

either building referenced in the 2006 document. 

A detailed description of Building 4064 and Building 4023 operations, processes, releases, and 

incidents are provided in this section. 

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Section 15.1 

Building Number:

Building Alias: 

4064 

064/T064 

Building Name: Fuel Storage Facility 

Building Function: Nuclear Material / Source Radioactive Material Storage

Radioactive Waste Storage

Contained materials for TRUMP-S Tests 

Notes: Tritium & Neptunium

Drainage through Area III to Area II Ponds 

Radionuclides of Concern: U-233, U-234, U-235, U-236, U-238, Th-228, Th-232, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Ra-226, Cs-134, Cs-137, Sr-90, H-3, Na-22, K-40, 

Mn-54, Fe-55, Co-60, Ni-59, Ni-63, Eu-152, Eu-154, Pm-147, Ta-182, Neptunium 

15.1.1 Description of Operations & Processes 

Building 4064 was constructed in two phases; the first in 1958 and the second in 1963. It was 

built to meet AEC criteria for vaults for the storage of fissionable material, equipped with 

intrusion alarms, and fenced. 1 The HSA provides a detailed description of the building’s 

features, equipment, schematics, and processes. 

Building 4064 was designed and built as a special nuclear material and source radioactive 

material storage building. It was used for storing packaged items of source material (natural 

uranium, depleted uranium, thorium) and special nuclear materials (enriched uranium, 

plutonium, and U-233) of various forms and configurations. Originally, the north (Room 114) and 

south (Room 110 and 104 vault) contained steel racks for storing material. The south side was 

primarily used for the storage of highly enriched uranium and plutonium bearing items; the north 

side was primarily used for source material and “low” enriched uranium storage.2 

Enriched uranium powders and source material powder packages were split into smaller units or 

combined into larger units in a glove box located in Room 104. The glove box was removed 

from Building 4064 at an unknown time. During shutdown and termination of the SNAP

program, excess Zr-U (enriched uranium) alloy material was sectioned into lengths suitable for 

1 EPA/HGL, Final HSA 6, p. 158, citation 5:

5 Boeing Environmental Affairs, Fact Sheet, Building 4064 – Fuel Storage Facility, February 10, 2000. 

2 EPA/HGL, Final HSA 6, p. 159, citation 4:

4 Remley, M. E. General Storage Data for Buildings 064 and 022, October 7, 1977. 

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packaging for shipment in DOE containers. This was done in Room 104; the floor was covered 

with plastic sheeting and the Zr-U was sectioned using a hack saw.1 

During the early 1960’s, the metal racks in the south half of Room 110 were removed in order to 

store material in “bird cages” and drums. This storage included large quantities of special 

nuclear material recoverable scrap. At this time, the fenced yard areas in the front, side, and 

back of Building 4064 were used to store 55-gallon drums of low-level enriched recoverable 

scrap. This material was shipped off-site in the mid-to-late 1960’s and early 1970’s, and may 

have relied on packaging containers constructed by employees at the Building 4163 Box Shop. 

Rockwell International claims that residual contamination from handling bare metallic pieces 

was from enriched uranium, natural uranium, depleted uranium, and thorium. Most reactor 

contracts had ended by the early 1980’s. After 1980, no special nuclear material powders were 

handled or repackaged at this location. After all fissionable material had been removed, 

miscellaneous equipment and containers of radioactive waste (principally soil) were stored in 

Building 4064. In 1993, all nuclear material was removed and Building 4064 was 

decontaminated. 

In 1989, EPA visited this location during the process of cleaning it up. EPA observed an area of 

radiation measuring 60 μR/hr. EPA used a shovel to dig at the area, about 1-foot into the soil 

and was able to increase the surface reading to 200 μR/hr. The technician conducting the onsite 

survey of the area declared that about 50 pCi/gm of beta radioactivity had been at this site.At 

the time of EPA’s visit, SSFL personnel described a spill that had occurred in the early 1960’s.2 

The HSA provides a detailed description of previous radiological investigations and 

decontamination/cleanup of releases. In brief, Rockwell International discovered high alpha 

radioactivity on the concrete ramps leading to rooms 110 and 114 in 1988. The alpha activity 

was concluded to arise from natural elements in the concrete that could not be removed. 

Ambient gamma exposure rate measurements showed a contaminated area bordering and 

outside the eastern fence, but according to Rockwell, no contamination was detected inside the 

fenced-in storage yard. Ambient gamma exposure rate measurements made outside the fence 

in the 2-acre area showed an area of about 4,000 square feet contaminated with Cs-137. Two 

soil samples collected from an area of greatest exposure rate showed a Cs-137 radioactivity 

concentration of 2,500 pCi/g, regarded as 2,500 times Rockwell’s background concentration. 

Beta activity was measured at 1,200 pCi/g in the area of greatest contamination.3 

In 1992, Rockwell surveyed a 6,580-square-foot area comprising the fenced in yard that 

surrounded Building 4064 to assess its radiological condition. All alpha surface activity levels 

were found to be below “Rockwell International’s acceptance limit” of 5,000 dpm/100 cm2. All 

beta surface activity levels were found to be below “Rockwell International’s acceptance limit” of 

1 EPA/HGL, Final HSA 6, p. 159, citation 3:

3 Montgomery Watson Harza, DOE Leach Fields (Area IV AOC) RCRA Facility Investigation Report, Santa 

Susana Field Laboratory, Ventura County, California, Draft, October 2003, pp. 2-3. 

2 Dempsey, Gregg: “Site Visit to Santa Susana Field Laboratory,” EPA Memorandum to Daniel M. Shane, On-Scene 

Coordinator, Emergency Response Unit. July 28, 1989 

3 EPA/HGL, Final HSA 6, p. 164, citation 1:

1 Chapman, J. A., Radiological Survey of the Source and Special Nuclear Material Storage Vault-Building 

T64, GEN-ZR-0005, August 19, 1988, pp. 2, 91-92. 

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5,000 dpm/100 cm2. All gamma exposure rates were found to be below “Rockwell’s acceptance 

limit” of 5 μR/h above Rockwell’s background rate of 15.3 5 μR/h. Rockwell concluded the 

fenced-in yard of Building 4064 met its criteria for release for unrestricted use.1 

In 1992 ORISE concluded Rockwell’s cleanup was incomplete upon identifying three Cs-137 

hotspots in the site yard. Rockwell subsequently remediated the hotspots and revised the 

Building 4064 side yard guidelines to meet a more restrictive 10 mrem/yr maximum dose rate 

for a residential scenario.2 

In 1993, Rockwell removed all known radioactively contaminated equipment, components, and 

structures from Building 4064, as well as asbestos containing materials. All radioactive waste 

was packaged and shipped to an unnamed disposal facility. This activity may have relied on the 

construction of shipping containers by employees in the Building 4163 Box Shop.3 

The preliminary MARSSIM classification for the building 4064 area is Class 1, due to its location 

within the ETEC and its former use as storage for radioactive source and special nuclear 

materials and radioactive waste. 

15.1.2 Building 4064 Radiological Incident Reports 

There have been several incidents associated with Building 4064 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

15.1.3 Building 4064 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0523 3/10/1959 SS Vault SSFL Personnel entered vault area without 

required film badges.

A0028 2/18/1963 Yard SS Vault MFP Cask drain plug rusted out, draining 

contaminated water onto the ground.

N/A 5/3/1963 SS Vault U Cans of UC exploded inside SSFL vault.

1 EPA/HGL, Final HSA 6, p. 164, citation 8:

8 Kneff, D. W., Tuttle, R. J., and Subbraman, G., Radiological Assessment of the Building T064 Fenced-in 

Yard, Rockwell International, N704SRR990035, January 12, 1994, pp. 10, 40-41. 

2 EPA/HGL, Final HSA 6, p. 165, citation 1:

1 Vitkus, T. J., Verification Survey of the Old Conservation Yard, Building 064 Side Yard, and Building 028, 

Santa Susana Field Laboratory, Rockwell International, Ventura County, California, ORISE 93/J-107, 

October 1993, pp. 5-11. 

3 EPA/HGL, Final HSA 6, p. 165, citation 3:

3 Dahl, F. C. and Tuttle, R. J., Final Radiological Survey Report of Building 064 Interior, ETEC Report No.

SSWA- ZR-0001, January 14, 1994, pp. 17-19. 

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A0468 10/8/1964 SS Vault U Can of UC exploded inside a “birdcage” 

shipped to SSFL Vault.

A0622 7/20/1967 SS Vault SSFL U308 Outside transfer of U powder caused 

increased activity in vegetation samples.

A0095 1/12/1982 Vault RAS Alarm Response.

A0109 9/21/1982 Vault Response to RAS alarm at SS Vault.

A0218 6/8/1992 Storage Racks U Employee stacking fuel storage racks cut 

hand.

A0663 8/8/1995 Side Yard Cs-137 Performing the modified Area IV transient 

survey in grid R24S25, a hot spot was 

found.

A0687 9/28/1998 Side Yard Cs-137 Excavated 6 cubic yard of Cs-137 

contaminated soil.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

*Isotopes are written as they are presented in the incident database. The HSA research team believes that “MFP” is 

an acronym for Mixed Fission Products. 

• On March 10, 1959 an employee entered the storage vault without a film badge and it was not 

known how long he was in the vault. The employee was reminded on other occasions to wear 

his film badge (A0523).1

• On February 18. 1963 soil east of the exclusion fence at the SS Vault was found to be 

contaminated during a routine survey. Three additional surveys revealed an area of 700 

square feet of asphalt and soil contaminated with mixed fission products to a maximum of 700 

mrad/h at 2 inches. It was deduced that the contamination resulted from a cask containing 

irradiated Seawolf submarine reactor fuel pins. The cask was received at the vault, stored, 

transferred to the Hot Lab for inspection, and returned to the vault for storage. After storing the 

fuel pins for 1.5-2 years it was shipped back to Westinghouse around May, 1962. Sometime 

during this last storage period, it is believed that the drain plug rusted through, permitting fluid 

contaminated within the cask to spill onto the asphalt. A soil sample showed 8,340 pCi/g betagamma gross radioactivity, including 1x106 dpm/g of Cs-137 and 2x105 dpm/g of Cs-134. 

Alpha activity was negligible in all samples. Approximately 2,365 gallons of soil and asphalt 

were removed at the vault to reduce the maximum contamination level to 0.5 mrad/h. Three 

inches of soil was removed from all the contaminated soil areas and 80 percent of the areas 

were found to be free of contamination upon resurveying. The remaining areas were 

excavated up to 1.5 feet. Fill soil was brought in by the maintenance department to return the 

1 EPA/HGL, Final HSA 6, p. 161, citation 1:

1 Loba, M. L., Internal Letter, Re: Violation of Health Phys. Practices, March 13, 1959. 

1 EPA/HGL, Final HSA 6, p. 161, citations 2 & 3:

2 

Badger, F. H., North American Aviation Internal Letter, re: Contamination Incident SS Vault Santa Susana, 

November 11, 1965.

3 Moore, J. D., Atomics International Internal Letter, re: Radioanalysis Report, February 22, 1963. 

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soil levels to their previous level (A0028).1

• On May 3, 1963 an explosion occurred in the north vault of Building 4064. The sources of the 

explosion were a 1-gallon can containing 8.3 kilograms of uranium carbide (3.7 percent 

enriched U-235) in the form of fines and pieces, and a 1-gallon can containing 2.5 kilograms of 

uranium carbide (4.9% enriched U-235) comprising sintered pieces. The fuel in both cans was 

immersed in Redline 60 oil. The explosion was believed to have been caused by the release 

of hydrogen gas from degrading uranium carbide. After 20 minutes elapsed, the Atomics 

International Fire Department entered the vault and smothered the cans with Metal X. 

According to Atomics International, a smear survey confirmed that contamination had been 

confined to the interior of the vault, however the Incident Report indicates that contamination 

was noted some contamination on the grill of the ventilation exhaust. The cans were removed 

and the vault was decontaminated using kerosene and K-pads. Atomics International planned 

to remove approximately 10 square feet of asphalt floor tiles where the cans had been 

located.2

• On October 8, 1964 a custodian found black oxidized powder deposited on the inside of a 

“bird cage” container. Investigation revealed that uranium carbide had oxidized blowing the lid 

off one 1-gallon can and warping the bottom of the container. The inside of the can was 

surveyed at 5 x 103 dpm/100 cm2 alpha. Contamination levels on the concrete dock of the 

building 4064 increased to 200 dpm / 100 cm2 alpha from a “clean” level. Evidence suggested 

that an inert atmosphere had not been sufficiently established in the 1-gallon container 

(A0468).3

• On July 20, 1967 an investigation was performed to identify reasons for increased alpha 

activity in vegetation samples between Building 4064 and the fire station. It was determined 

that uranium carbide and uranium oxide powder had been transferred from one 55-gallon 

drum to another in the Building 4064 storage yard. The drum from which material had been 

removed was still located in the yard on a piece of plastic sheeting and visible amounts of 

uranium oxide powder remained. The plastic was folded carefully and disposed of as 

radioactive waste. No “significant” surface contamination was found in areas other than the 

plastic sheeting and vegetation samples collected between Building 4064 and the fire station. 

Personnel were cautioned about procedures for transferring radioactive materials or opening 

2 EPA/HGL, Final HSA 6, p. 161, citation 4:

4 Coonce, G. L., Atomics International Internal Letter, re: UC Fire at Building 064, May 8, 1963. 

3 EPA/HGL, Final HSA 6, p. 162, citation 1:

1 Owen, D. E., Atomics International Internal Letter, re: Incident Report, Building 064 Vault, 10-8-64, October 

27, 1964. 

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drums of radioactive materials (A0622).1

• On January 12, 1982 an alarm went off in Building 4064. Crews arrived at the building to 

check it out and the all-clear was given within an hour of arriving on the scene. The cause for 

the alarm was unknown and the alarm was to be checked by maintenance (A0095).2

• On September 21, 1982 an alarm went off in Building 4064. Crews arrived at the building to 

check it out and the all-clear was given within an hour of arriving on the scene. The cause for 

the alarm was unknown and the alarm was to be checked by maintenance (A0109).3

• On June 8, 1992 an employee was stacking storage racks on a pallet when he cut the palm of 

his hand. No detectable activity was noted on the employee’s hand or the storage racks, 

despite concern for fixed U-235 contamination under paint on the fuel storage racks (A0218).4

• On August 8, 1995 a hot spot of approximately 1 square foot was identified during an Area IV 

survey in grid number R24S25. The hot spot was measured at 10,000 cpm or 46 μR/h. Per 

Rockwell’s procedure, a 1 meter ambient survey was performed, but it was below the 5 μR/h

above background limit for classifying as a hot spot. However, the Building 4064 sideyard was 

a formally remediated facility, so the discovery was reported to the Area IV survey project 

manager. The hot spot was roped off and a sample was taken. The sample contained 271 pCi/

g of Cs-137. The Cs-137 concentration was too low to be considered an inhalation hazard and 

the surface radiation was only three times background, so there were no specific safety 

concerns. According to Rockwell, the discovery of contamination did not warrant a formal 

occurrence report because building 4064 was not yet released for unrestricted use. The hot 

spot had been identified in a verification survey by the Oak Ridge Institute for Science and 

Education (ORISE), but had not been properly communicated to Rockwell. The discovery was 

cited as an example of the importance of a walk-about surface gamma surveying used in the 

Area IV characterization survey (A0663).5 The exact location of this hot spot has not been 

identified but it appears to be west of Building 4064.

1 EPA/HGL, Final HSA 6, p. 162, citation 2:

2 Alexander, R. E., Atomics International, re: Radiation Safety Unit Weekly Newsletter for Period Ending July 

22, 1967, August 2, 1967. 

2 EPA/HGL, Final HSA 6, p. 162, citation 3:

3 Bradbury, S. M., Rockwell International Internal Letter, re: Radiological Safety Incident Report, Building 

064, 1/12/82, Unknown Date. 

3 EPA/HGL, Final HSA 6, p. 162, citation 4:

4 Bradbury, S. M., Rockwell International Internal Letter, re: Radiological Safety Incident Report, Building 

064, 9/21/82, September 22, 1982. 

4 EPA/HGL, Final HSA 6, p. 162, citation 5:

5 Wallace, J. H., Rockwell International Internal Letter, re: Radiological Safety Report, T064 South Vault, 

6/8/92, June 17, 1992. 

5 EPA/HGL, Final HSA 6, p. 163, citation 1:

1 McGinnis, E. R., Rockwell International Internal Letter, re: Radiological Incident Report, Soil Contamination 

Found in T064 Sideyard, 8/8/95, August 25, 1995. 

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• On September 28, 1998 ORISE discovered an area of elevated soil contamination in the side 

yard. The area measured approximately 18 feet by 6 feet and was located on the sloping bank 

between 10th street and the Building 4064 parking lot. A total of 6 cubic yard of soil were 

excavated until radiation levels reached background levels. ORISE resurveyed the area, 

verified that radiation levels were normal, and collected two soil samples. Rockwell collected 

additional samples to confirm that the remediation had been successful (A0687).1 

1 EPA/HGL, Final HSA 6, p. 163, citation 2:
 

2 Rutherford, R., Rockwell International Incident Report, re: Verification Survey, September 28, 1998.
 

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Section 15.2 

Building Number:

Building Alias: 

4023 

023/T023 

Building Name/Function: SNAP Liquid Metals Component Test Building

Corrosion Test Loop

Sodium Test Loop 

Notes: Tritium / Neptunium

SNAP: DOE-NASA (NAA S&ID Rocketdyne Personnel) 

Radionuclides of Concern: Possible radionuclides of concern identified include americium-241 

(Am-241), isotopes of plutonium (Pu- 238, Pu-239, Pu-240, and Pu-241); fission products 

cesium-134 (Cs-134), Cs-137, strontium-90 (Sr-90); source and uranium products, thorium-228 

(Th-228), Th-232, uranium-234 (U-234), U- 235, U-238; isotopes of neptunium (Np-237); and 

activation products, cobalt-60 (Co-60), manganese-54 (Mn-54), europium-152 (Eu-152), 

Eu-154, and tritium (H-3), nickel-63 (Ni-63), iron-55 (Fe-55), and tantalum-182 (Ta-182).1 

15.2.1 Description of Operations & Processes: 

Building 4023 served as the Liquid Metals Component Test Building and the Corrosion Test 

Loop. The “old” portion of Building 4023 was constructed in 1962, and the “new” building section 

was constructed in 1976. While the 1969 site layout does not show a radioactive waste holdup 2

tank, a plan-view of the building from a 1993 final survey report does. The sodium test loop was 

located in a small partitioned area of the “old” portion of the building, which has also been used 

1 EPA/HGL, Final HSA 5-A, p. 22, citations 63-67: 

63 Cabrera Services, Inc., Final Status Survey Report: Final Status Survey Post Historical Site Assessment 

Sites, Block 1, March 2007.

64 Rocketdyne, Rockwell International, Building 023 Final Survey Procedure, 023-SP-001, August 3, 1993

65 Rockwell International, Revised TRUMP-S Project Radioactive Materials Usage Application, 

190TI000001, October 31, 1989.

66 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005

67 Sapere Consulting, Inc. referenced a January 20, 1993, internal letter regarding potential contaminants at

T023. The research team has not located this document in the documents provided to date. 

2 EPA/HGL, Final HSA 5-A, p. 12, citations 12 & 13:

12Drawings appear to indicate that A.C. refers to “asphaltic concrete.” The ditches surrounding this building 

were paved and likely served as a pipe trench for the cooling system return and supply lines between the 

building and the evaporator (Atomics International Drawing 303-027-C2, Topographic Plan, March 4, 1963).

13 DOE, Certification Docket for the Release of Building 023 at ETEC, Docket No. DOE/CD-ETEC-023, 

February 1997. 

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for storage. The “new” building section held an analytical chemistry laboratory and a storage 

set-up room.1 

The old portion of the building housed a small sodium loop to conduct studies of radioactive 

contamination transport. A 1976 Use Authorization allowed the use of a small section of 

activated stainless steel Experimental Boilers Reactor fuel cladding in a small sodium test loop 

to gather data on transport characteristics of radiological contamination in sodium loops.2 

According to a 1987 Site Consolidation Assessment, Building 4023 served as a development 

and demonstration test facility in support of the Rocky Flats Plutonium Recovery Project in 

1987. These operations were located in the east bay and chemistry laboratory, while the west 

bay of the building was being used for the storage of power jet hardware. Available information 

indicates the project included the mechanical and structural design of 24 glove boxes, special 

manipulators and conveyor systems. The project also included the development of design 

criteria, definition of process equipment and glove boxes to manage plutonium, development of 

flow diagrams, design and specification of equipment and glove boxes, safeguard evaluations, 

experimental verification designs, reliability, availability and maintainability evaluations, and 

preparation of conceptual design reports.3 

In 1989, reports appear to indicate that Building 4023 served as a support facility for the 

Transuranic (TRU) Management by Pyropartitioning - Separation (TRUMP-S) operations in 

Building 4020 (Hot Lab). Rockwell International requested DOE’s approval to utilize the facilities 

for a 2-year period beginning July 1988 for the Kawasaki Heavy Industries (KHI) and the Central 

Research Institute of the Electrical Power Industry (CRIEPI) of Japan-sponsored ‘pyrochemical 

partitioning of actinides from PUREX waste” program. According to the reports, inductivelycoupled plasma (ICP) analysis at Building 4023 was not higher than 10 μCi of activity. The ICP

analyzer was located in Room 102 A. The material used in this experiment was listed as 

including uranium, neptunium (Np-237), plutonium (Pu-239), and americium (Am-241). 

1 EPA/HGL, Final HSA 5-A, p. 12, citation 14:

14 ORISE, Verification Survey of Buildings 005, 023, and 064, Santa Susana Field Laboratory, Rockwell 

International, Ventura County, California, 94/K-14, October 1994. 

2 EPA/HGL, Final HSA 5-A, p. 13, citations 20, 21, 22: 

20 Information contained in historical documents obtained by HGL to date have not provided information 

regarding specific building operations that occurred from 1962 through 1976. Additional information 

regarding the operations during this time period are needed.

21 DOE, Certification Docket for the Release of Building 023 at ETEC, Docket No. DOE/CD-ETEC-023,

February 1997.

22 Liquid Metals Test Building – 023, Santa Susana Facility, Ventura County, California, Building Piping Plan, 

303- 023-M5, January 1, 1963. 

3 EPA/HGL, Final HSA 5-A, p. 14, citations 26 & 27:

26 ETEC, Site Consolidation Assessment, April 16, 1987. p. 13.

27 

Litwin, R.Z., Rockwell International Background and Capability to Develop a Weapons-Grade Plutonium 

Fuel Cycle and Disposal Evaluation for the PDR, October 5, 1992. 

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The process of transporting the material from Building 4020 Hot Lab to 4023 was 

identified as follows:1 

“When aqueous samples are transferred from Building T020 to Building T023 for analysis and 

back to T020 for solidification and disposal, no more than 10 μCi are contained in the 20-

milliliter aliquot needed for rare earth analysis in Building T023. The aliquot will be placed in a 

25-mL glass vial and sealed with a screw top cap. The glass vial will be placed in a plastic bottle 

with a screw-top cap. R&NS will smear the bottle to make sure it is uncontaminated, and the 

plastic bottle will be sealed in a plastic bag. The bag will be placed in a carrier which will be 

transported to Building T023. At Building T023, a sphincter port will be used to introduce the 

sample and its plastic bottle into the nebulizer glove box. Excess sample and waste will be 

bagged out of the nebulizer glove box and taken back to T020 [Hot Lab] in a similar manner to 

that described above.” 

While program was listed as being part of “DOE’s Private Sectors Initiative,” and DOE 

agreed to provide the experimental materials and dispose of the waste. At the 

conclusion of the work, the facilities were to be decontaminated to a “level that they may 

be released for unrestricted use.” The confirmatory survey was to have been funded by 

the project. The duration of these activities in Building 4023 could not be located. 

A March 1991 environmental, health, and safety self-assessment of ETEC provided a 

five-year plan that included activities for the decommissioning of Building 4023. 

According to this document, activities planned for Building 4023 included the 

surveillance and maintenance of “remaining radioactively contaminated systems in and 

around” Building 4023, including a disconnected radioactively contaminated hood inside 

the building and the associated piping, and a radioactive liquid holdup tank, all of which 

were included in the Fiscal Year 1993 plan for decontamination and decommissioning. 

In 1992 and 1993, an analytical and ICP instrument was used to analyze small amounts 

of radioactively-contaminated solution from the Molten Salt Oxidation project for trace 

metals. According to a final radiological study completed in 1994, the work was 

reportedly completed without incident and the remainder of the radioactive solutions and 

containers were transferred to the Radioactive Material Handling Facility (RMHF). The 

1 EPA/HGL, Final HSA 5-A, p. 14, citations 28 & 29: 

28 Rockwell International, Safety Analysis for Building T020, Operations, TRUMP-S, 190ER000011, 

December 4, 1989.

29 Rockwell International, Revised TRUMP-S Project Radioactive Materials Usage Application, 

190TI000001, October 31, 1989. 

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study did not indicate the quantity or nature of the radioactive solutions and containers 

transferred to the RMHF. The building was released in February, 1998.1 

15.2.2 Building 4023 Potential for Off-Site Contamination 

According to the EPA Area IV Study and HSA, Building 4023’s 1964 and 1994 plot plans show 

an “A.C. ditch” that is also depicted in aerial photos. EPA’s research team did not locate any 

documented information regarding the contraction of the ditch, which terminated at a catch 

basin and then flows into an 18-inch corrugated metal pipe. The pipe directed flow to a drainage 

channel that ultimately led directly to the “R-2A Pond,” located in Area II. The R-2A Pond was 

part of the Site Wide Reclaim Water System, designed and constructed by AEC-DOE and 

Atomics International to keep Area IV from flooding and to provide much needed water to the 

rocket engine test stand workers (employed by North American Aviation) throughout Areas I, II 

and III.2 

The preliminary MARSSIM rating of Building 4023 is Class 1. 

15.2.3 Building 4023 Radiological Incident Reports 

There have been several incidents associated with Building 4023 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

15.2.4 Building 4023 Radiological Incident Report Summary: Data Provided by Boeing 

Incident File 

Name

Date of Incident Location of Incident Isotopes Description of Incident

A0084 12/18/1980 R/A SOD UM LOOP Active Corrosion 

Products

Rinse water surged out of deconned loop leg 

contaminating the area

A0257 4/22/1981 Sodium Loop Co 60/Mn 54 Five grams of hot radioactive contaminated sodium 

leaked from loop and burned

1 EPA/HGL, Final HSA 5-A, p. 15, citations 33-37: 

33 Rockwell International, Document No. GEN-AR-0023, An Environmental, Health, and Safety SelfAssessment of the Energy Technology Engineering Center, Volume 2, March 18, 1991.

34 Flore, J.J., Department of Energy, Docket No. ETEC-023, Certification of the Radiological Condition of

Building 023 at the Energy Technology Engineering Center Near Chatsworth, CA, March 12, 1997.

35 Rocketdyne Report, Final Radiological Survey Report of Building 023, 023-ZR-0001, March 1, 1994.

36 Liddle, Roger, DOE/OAK, Letter, “Release of Facilities for Unrestricted Non-Radiological Use,” April 21, 

1997. 

37 Wong, Gerald, DHS/RHB, Letter, “Boeing’s Request for Concurrence in Release for Use Without 

Radiological Restriction, Rocketdyne Santa Susana Field Laboratory Building T023,” February 19, 1998. 

2 U.S. DOE Office of Environmental Audit, “Environmental Survey Preliminary Report, U.S. DOE Activities at SSFL,” 

1989. 

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16.0 Miscellaneous Operations 

Neither the Conservation Yard nor the Sodium Disposal Facility were intended for use with 

radioactive materials, but both were inadvertently contaminated. 

Section 16.1 

Building Number:

Building Alias: 

4583 

583/T583 

Building Name/Function: “Old Conservation Yard” / “ESG Salvage Yard”

Salvage, Scrap & Barrel Storage Yard

Fuel Tank Farm 

Notes: Tritium 

Drainage to Area II Ponds 

Conservation Yard Radionuclides of Concern: U-238, U-234, U-235, U- 236, Pu-239, 

Pu-240, Pu-241, Pu-242, Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe- 59, Co-60, 

Kr-85, Sr-89, Sr-90, Sb-125, I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, Nb- 95, 

Ru-103, Ru-106, Xn-133, Xe-135, Pm-147, Sm-151. 

16.1.1 Description of Operations and Processes 

This location consisted of a 3-acre area comprising mostly natural terrain used between 

1952-1977 as a salvage yard. It was later known as the “Old ESG Yard.” Radioactive and non 1 -

radioactive materials from the SRE and other programs were brought to this site for storage. 

Excess salvageable items were stored at this location and eventually, it spread to include 

surrounding areas. Although not operated as a radiologically-controlled facility, Rockwell 

surveyed these areas for contamination because, according to Rockwell, although there was no 

“deliberate” dumping or placing of radioactive materials at this location, since there were no 

controls in place, radiologically contaminated items were found. In addition, items from the 

DeSoto Facility were stored at this location. 

In 1969, the barrel storage yard was converted to a material storage yard for Plant Services, 

which relinquished control of the area in 1986 to Rockwell’s Transportation Department. The Old 

Conservation Yard / ESG Salvage Yard was used extensively during the 1960’s and 1970’s 

before it was closed in 1977 and cleared of materials. In 1988, shipping trailers and casks were 

stored in the old barrel storage yard. In the early 1980’s, this location became a fuel tank farm, 

which was demolished in 1999. 

15.1.2 Conservation Yard Radiological Incident Reports: There has been one incident 

associated with the Conservation Yard that could have resulted in a release to the environment 

and worker exposure. The following table provides information presented in an incidents 

1 EPA/HGL, Final HSA 6 p. 186, citation 1: 

1 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. 

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database provided by Boeing. Summaries of all available incident reports are provided in 

Attachment A. 

The preliminary MARSSIM classification for this location, due to its former use as a storage area 

for radioactive materials and because it is not apparent that a full radiologic characterization of 

the area was conducted, is Class 1. 

The HSA provides detailed information and a chronology of previous radiological investigations 

conducted at this location. However, in brief, Rockwell International performed a radiological 

survey at this location in 1988 to determine whether residual contamination remained from 

former storage operations. Rockwell found that the average ambient gamma exposure rate in 

an area of the yard was 27.9 μR/h (Rockwell’s acceptable background was 15.3 μR/h, and 

under 5 μR/h above). Rockwell also discovered Cs-137 contamination. 

16.1.3 Conservation Yard Radiological Incident Report Summary

Data Provided by Boeing 

*Isotopes are written as they are presented in the incidents database. The HSA research team believes that MFP is 

an acronym for mixed fission product. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0288 1/15/1976 conservation yard MFP Contaminated drums and soil discovered 

at Conservation Yard, etc.

• On January 15, 1976 a radioactive spill was detected at the conservation yard with a pug 

survey meter. The spill spread across the conservation yard access road and terminated in an 

equipment holding area. Contamination was found on barrels (up to 800 mrad/h), pallets (up to 

2 mrad/h) and in the surrounding area (up to 35 mrad/h). Other contaminated items were 

discovered at the conservation yard, although it is not clear from the incident report when these 

items were discovered. Contaminated items included a barrel (5 mrad/h), several pallets (up to 

8 mrad/h), a concrete block (3 mrad/h), and a radioactive liquid waste drum (5 mrad/h). A barrel 

was found down the hill on the Simi Valley side of the conservation yard (2 mrad/h) and a pair of 

gloves belonging to the conservation yard operator (.3 mrad/hr). Externally contaminated barrels 

appeared to have resulted from a double stacked barrel of radioactive liquid that corroded and 

leaked out onto the drums and pallet below. As for the source of the contaminated liquid that 

caused the spill, there were not any significant leads. However, the liquid was thought to be a 

cleaning solution. The area was secured and radioactively contaminated asphalt, soil, barrels 

and pallets were removed to the RMHF for disposition (A0288).1

1 EPA/HGL, Final HSA 6 p. 187, citation 1: 

3 Harris, J. and Badger, F., Rockwell International Internal Letter, Re: Conservation Yard Spill, February 6, 

1976. 

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Section 16.2 

Building Number:

Building Aliases: 

4886

886 / T886 

Building Name: Sodium Burn Pit 

Building Function: Waste Disposal Via Combustion, Incineration, Burial 

Radionuclides of Concern: U-238, U-234, U-235, U-236, Pu-239, Pu-240, Pu-241, Pu-242, 

Am-241, Th- 232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-55, Fe-59, Co-60, Kr-85, Sr-89, Sr-90, 

Sb-125, I-131, Cs-134, Cs-137, Ce-141, Ce-144, Ba (La)-140, Nb-95, Ru-103, Ru-106, Xe-133, 

Xe-135, Pm- 147, Eu-152, Eu-154, and Sm-151 

16.2.1 Description of Operations and Processes 

The “Old Sodium Burn Pit” is an area where radiologically contaminated materials had been 

dumped at some point in time. At some point, a protective dike or berm around this area had 

washed away and material from this pit was allowed to move off the site in an uncontrolled 

fashion.1 EPA noted that Rocketdyne had not collected soil or water samples in this area to be 

analyzed for tritium. If the materials dumped at this area and others contained tritium, there was 

no way in Rocketdyne’s measurement protocol to detect it. 

The Sodium Burn Pit (Building 4886) comprised approximately 2 acres. It operated from 1956 

until 1978. In the late 1950’s, a 16,700-square-foot area on the western side of the site was 

developed into an unlined pond. In the 1960’s, a rectangular concrete-lined pool was 

constructed at the southeast corner of the site, surrounded by a concrete slab. A shed was 

located south of the concrete lined pool and a steel protective shield containing portholes was 

located at the east end of the pool. The steel shield also had a hand hole to accommodate a 

hose nozzle. 

In the late 1960’s two man-made shallow water-filled unlined ponds, known as the upper and 

lower ponds, were constructed. The lower pond was located in the middle of the site; the upper 

pond was located just north of the concrete-lined pool. These ponds, (lower 8,250 square feet 

and upper 6,500 square feet) were connected by piping. 

In May 1956, Atomics International decided to locate a waste disposal area on the western 

region of SSFL because this area was underlain by shale of the Santa Susana-Martinez 

Formation. Atomics International believed that the percolation rates in the shale were low and 

that the rock appeared to be unfractured.2 

1 Dempsey, Gregg: “Site Visit to Santa Susana Field Laboratory,” EPA Memorandum to Daniel M. Shane, On-Scene 

Coordinator, Emergency Response Unit. July 28, 1989 

2 EPA/HGL, Final HSA 8 p. 40, citation 1: 

1 

Jarrett, A. A. and Roth, J. N., Quarterly Progress Report, January, February, March 1956, Atomics 

International, May 8, 1956, pp. 1-2. 

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In 1987, Rockwell reported that the Burn Pit was used extensively during the 1960 to 1970 

period for the disposal of combustible materials such as kerosene, Na, and NaK from scrap test 

components such as pumps, valves, etc., from the SRE, SNAP, and other nuclear program 

operations. The area comprised a concrete pad adjacent to a concrete pool containing water 

and upper and lower pond area, both of which down-sloped from the pool. A steam lance was 

used to clean the equipment. Components to be cleaned were placed on the slab, opened to 

expose the Na or NaK, and then washed off with water. The water reacted with the sodium to 

generate hydrogen, which often burned in air. The washed items were placed into the concrete 

pool, where the reaction with water continued, and were removed from the pool and placed into 

the upper or lower pond, where they stayed until all residual sodium had reacted. Some items 

were retrieved and disposed of off-site as solid waste. An area adjacent to and west of the 

upper pond area was used for the storage and burial of materials. 

In 1988, Rockwell reported that “Santo-wax,” used as a coolant for organic moderated reactors, 

was also burned in the area. Rockwell also reported that occasionally, firearms were used on 

vessels to open containers to the atmosphere. Those items, after cleaning, were removed to a 

dumpster for a scrap dealer. Some large components and vessels were buried in place. Some 

barrels and scrap were buried on the far western side of the site. According to Rockwell, 

material was also dispersed onto surrounding land by explosions that extended to Building 

4009. The facility was also made available for the open burning of any combustible material. 

Radiological surveys had shown that the upper and lower ponds were contaminated with 

Cs-137 and zirconium hydride slugs contaminated with 93% enriched uranium. Radioactive 

contamination was suspected in the area because of the potential for radionuclide transport and 

migration from the ponds in the direction of surface-water runoff, and because of the dispersion 

and scattering of radioactive material during cleaning operations.1 

In 1978, Rockwell initiated cleanup of the Sodium Disposal Facility. Up until the mid 1980’s, 

pieces of debris such as pipes, elbows, machined metal parts, and tubes were dug up, “pushed 

over,” and reburied in the upper and lower ponds using bulldozers. All visible scrapped tanks 

were moved to Building 4133 for further disposition. The western p ond area was excavated, 

and hazardous materials and trash were removed. The concrete-lined pool was drained of water 

by a hazardous waste disposal company. Its walls were found to be radioactively contaminated 

and were scabbled. The lower and upper ponds were surveyed; both were found to be 

radiologically contaminated with the lower pond showing the highest readings. Cs-137 was 

identified as the principal gamma-emitting isotope.2 

1 EPA/HGL, Final HSA 8 p. 41, citations 1 & 2: 

1 Chapman, J. A., Radiological Survey of the Sodium Disposal Facility–Building T886, ETEC Report No.

GEN-ZR- 0004, June 3, 1988, pp. 7-8, 19-20.

2 Rockwell Internal Letter from F. H. Badger to R. J. Tuttle, Re: Radiological Information on Old Sodium

Disposal Area, April 23, 1987. 

2 EPA/HGL, Final HSA 8 p. 45, citations 1 & 2: 

1 

Klein, A., Final Report for Decontamination and Decommissioning of Former Sodium Disposal Facility

(FSDF) – B4886, Boeing Report No. EID-04628, November 17, 1999, p. 12.

2 Chapman, J. A., Radiological Survey of the Sodium Disposal Facility–Building T886, ETEC Report No.

GEN-ZR- 0004, June 3, 1988, pp. 7-8, 19-20. 

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In March, 1987 Rockwell conducted a site characterization study wherein 23 trenches were 

excavated between 1 and 7 feet in depth inside and north of the upper and lower ponds. A

trench was also excavated along the southern and western boundaries to channel precipitation 

around the ponds, deflect run-on rainfall, and minimize possible material dispersion. Laboratory 

analysis of soil samples found Cs-137 activity levels as high as 200 pCi/g. While the trenches 

were open, items of debris were uncovered together with differing colored soils and pungent 

odors. The exposure rate in this area was 80 μR/h. Rockwell estimated background gamma 

radiation to be 15 μR/h. In one trench in the lower pond, gamma radiation levels were found to 

be 2 to 4 times background while beta radiation levels were up to 14 times background at a 1-

foot depth. Items of radioactive debris found included zirconium hydride reactor fuel end caps 

contaminated with U-238 and a thoriated oxygen sensor used in the sodium loops. These items 

were not removed at the time of the investigation.1 

In February 1989, DOE conducted the first phase of an environmental survey of Area IV and 

concluded that the removal of contaminated equipment from the Sodium Burn Pit area was 

incomplete. Additionally, the diversion ditches, which had been installed to channel storm water 

away from the contaminated areas, had been breached during cleanup operations. As a result, 

storm water was free to run onto the site and potentially remove contaminated soils and 

sediments. This was designated a Category II finding, having the potential for release of 

contaminated runoff due to inadequate controls. Contaminant concentrations in water runoff 

exceeded drinking water standards for arsenic, chromium, and thorium. Cs-137 was identified 

as the main radioactive contaminant in soil.2 

On July 13, 1989 Gregg Dempsey from the EPA Office of Radiation Programs collected 

duplicate moist samples from the upper pond. These were analyzed for gamma emitting 

isotopes and H-3 (tritium). Results indicated normal or background levels of radioactivity in the 

area, for example. H-3 indicated a maximum of 0.59 pCi/g and K-40 indicated a maximum of 

28.81 pCi/g. Cs-137 was found at levels consistent with those from atmospheric fallout from 

nuclear weapons testing (Cs-137 maximum 0.94 pCi/g). A dry soil sample collected from the 

lower pond was analyzed for gamma emitting isotopes. The lower pond gamma levels were 

found to be roughly twice those of the upper pond. Mr. Dempsey recommended that additional 

water samples be collected and analyzed for H-3 and Sr-90.3 

In 1991, Rockwell conducted a baseline beta-gamma radiological survey of the Sodium Burn Pit 

Area, at near the soil surface. In the lower pond, the survey found elevated gamma activity at 

five locations; one location measured about twice normal background (27.5 μR/h) for gamma 

1 EPA/HGL, Final HSA 8 p. 46, citation 1: 

1 Schiffman, J., United States Department of Energy, Environmental Survey Report, June 16, 1989, pp.

8-10, 16, 20- 21. 

2 EPA/HGL, Final HSA 8 p. 46, citation 1: 

1 Schiffman, J., United States Department of Energy, Environmental Survey Report, June 16, 1989, pp.

8-10, 16, 20- 21. 

3 EPA/HGL, Final HSA 8 p. 46, citations 3 & 4: 

3 Dempsey, G., Report on Environmental Samples Collected at the Rocketdyne Santa Susana Field

Laboratory July 1989, , November 8, 1989, pp. 3-4.

4 Dempsey, G., Site visit to Santa Susana Field Laboratory Operated by Rockwell Rocketdyne, 

Memorandum, July 28, 1989, pp. 5-6. 

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contamination, and substantially elevated above background (6,215 dpm/100 cm2) for surface 

beta contamination.1 

The HSA provides additional details about further radiological investigations conducted at the 

Sodium Burn Pit area. In addition, it appears retroactive licenses may have been issued to 

address multiple fission product contamination in the soil.2 

The preliminary MARSSIM Classification for Building 4886, the Sodium Burn Pit Area, is Class 1 

because of its former use as a disposal area for non-radioactive and radioactive materials, and 

because it is not apparent that a full radiologic characterization of the area had been conducted. 

16.2.2 Radiological Incident Reports 

There have been several incidents associated with the Sodium Burn Pit that could have resulted 

in a release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing. Summaries of all available incident 

reports are provided in Attachment A. 

16.2.3 Sodium Disposal Facility Radiological Incident Report Summary

Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0464 06/11/1964 Sodium Disposal 

Area

I-131 Employee spilled caustic R/A solution on 

ground while cleaning tank.

A0630 05/03/1968 Burn Pit Three contaminated drums discovered 

being burned at the Sodium Burn Pit.

A0075 10/06/1978 Burn Pit & Shed Mixed Fission 

Products

Scrap removed from Sodium Burn Pit 

found contaminated, as was the pit.

• On June 11, 1964 an employee spilled approximately 2 gallons of caustic radioactive solution 

labeles “0.15 mrad/h,” on the ground while cleaning a stainless-steel tank. The incident report 

did not mention any efforts made to clean up the spill. By the time Health & Safety personnel 

arrived, the employee was inside a tank wearing red-line coveralls, gloves, and rubber boots. 

He then left the tank, rinsed off his gloves and boots, and walked around the area. The 

ground-surface reading measured approximately 0.1 mrad/h with a G.M. and window probe. A

reading of 100 dpm/gram (units may be incorrectly presented in the source document) was 

recorded in the center of the spill. It was recommended that bioassays be conducted on both 

employees assigned to the area. There was no follow-up to the incident report concerning the 

1 EPA/HGL, Final HSA 8 p. 47, citation 1: 

1 Collins, J., Baseline Radiological Survey of the Sodium Disposal Facility (T886), Rockwell International

Report No. N704SRR990034, August 31, 1992, pp. 1, 6-12, 18. 

2 EPA/HGL, Final HSA 8 p. 48, citation 3: 

3 Rockwell International Authorization Review by F. E. Begley, February 6, 1989. 

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bioassay (A0464).1

• On May 3, 1968 three radiologically contaminated drums were discovered being burned in the 

burn pit. The site was then sampled for radiological contamination. At the time of the HSA

investigation, EPA awaited receipt of Incident Report A0630.

• In 1978, a survey of old sodium barrels located at the Sodium Burn Pit was conducted 

pending outside vendor bidding for salvage rights. Three radiologically contaminated sodium 

barrels and a pallet with pieces of pipe, valves and trash were found. The barrels had a 

maximum activity of 1 mrad/h. A subsequent survey found the upper half of the walls of the 

concrete-lined pool to be contaminated to 1,000 cpm (0.4 mrad/h). The contaminated waste 

had spread south of the concrete-lined pool because a survey in the region of Building 4730 

identified two radioactive items. The metal waste storage area at Site 4886 contained five 

pallets of finned tubing contaminated to 5 mrad/h. All radioactive items were removed and 

moved to the RMHF for disposal (A0075).2

1 EPA/HGL, Final HSA 8 p. 43, citation 1: 

1 Atomics International Incident Report from D. E. Owens to R. M. Hill re: Sodium Disposal Area spill on 

June 11, 1964, dated June 18, 1964. 

2 EPA/HGL, Final HSA 8 p. 44, citation 2: 

2 Rockwell International Internal Letter from F. H. Badger to W. R. McCurnin re: Preliminary Radioactive 

Survey Sodium Burn Pit, dated October 6, 1978. 

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17.0 Additional Processes, Programs & Associated Facilities 

This section contains a partial listing of additional facilities, operations, job processes, and 

incidents that were excluded from the 2006 Site Description / Site Profile. 

The operations and facilities that were excluded include Coal Gasification facilities, the Sodium 

Components Test Installation (SCTI), Sodium Pump Test Facility (SPTF), the Small Components 

Test Laboratory (SCTL), Laser Testing operations and facilities, and several miscellaneous 

locations that supported a variety of programs and purposes; the Radioactive Laundry, Shipping 

and Receiving, and Hazardous Waste Materials Facility (HWMF) among them. 

Presumably, many of these locations were excluded from the Site Profile on the premise that 

they were not considered “radiological facilities” or integral to the nuclear programs at SSFL. 

However, several of the facilities have histories of supporting the nuclear programs at SSFL, or 

handled radioactive materials regardless of being designated a “non-radiological” location. Many 

of these facilities have documented radiological incident reports contained in the Incidents 

Database. In some cases, the excluded facilities were located in close enough proximity to a 

known source of radiation exposure (like the RMHF or an experimental reactor building) to 

warrant concerns about worker exposure and appropriate monitoring. 

Given the nature of SSFL operations, worker rotation, diverse job duties among the workforce, 

and the shifting nature of building function as programs and projects overlapped and changed 

over the years, a robust Site Profile cannot exist without some reference to all processes that 

were undertaken at the site, and a thorough review of building/facility history over the course of 

SSFL operations.

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Section 17.1 - COAL GASIFICATION 

The Rockwell International Molten Salt Coal Gasification Process (Rockgas Process), the 1-Ton 

and 1/4 Quarter Ton-Per-Hour (TPH) Coal Hydropyrolysis Pilot Test Facility, and coal gasification 

support facilities were operated under ETEC and DOE.1 

According to Rockwell International’s Energy System Division, under contracts with the DOE, 

the division advanced the development of two processes for the extraction of clean-burning gas 

(or liquid hydrocarbons) from coal at SSFL. DOE-ESG developed a system that used a molten 

salt mixture at a high temperature to convert any kind of coal to low or medium Btu industrial 

gas. The molten salt, from which sulfur and ash are removed for disposal, was continuously 

regenerated and recycled. These processes are further explained in Section 8.4 pertaining to 

Building 4005.2 

In 1978, DOE and Rockwell International negotiated the construction and modification of 

facilities at the SSFL Energy Bowl Area.3 Several coal gasification pilot plants, testing facilities, 

and support facilities were operated at SSFL beginning in the mid-1970’s.4 Two vertical test 

stands (VTS) were modified to accommodate coal gasification and liquefaction pilot test plants. 

In 1975, coal hydroliquefaction, a process where pulverized coal was put into a high-pressure 

environmental to produce benzene, toluene, and napthalene, was carried out. The liquids were 

then condensed, and the products were transferred into about one hundred 55-gallon drums. 

Some of the condensed liquids were sent out for qualification and quantification analysis. During 

the transfer of the liquids to the 55-gallon drums, workers had contact with the liquid, breathed 

the fumes, and spilled the liquids onto their clothing.5 

To obtain pulverized coal, rock coal was ground through a fine mesh in a pressurized system. 

Workers were required to open plugged lines to obtain the pulverized coal. Workers were 

provided with dust masks, but their use could not be confirmed. The coal hydroliquifaction 

operations were conducted from approximately 1976 until 2005.6 Coal gasification processes 

are related to radium exposure and directly related to occupational skin cancer. 

1 Rockwell International for DOE, “Advancement of Flash Hydrogasification: Task VIII — Performance Testing” by A.Y. 

Falk, M.D. Schuman, and D.R. Kahn. June 1986. Work performed under DOE Contract #DE-AC21-78ET10328 

2 Rockwell International Energy Systems Group (ESG) Annual Report. File: ESG_Report.pdf 

3 Internal Letter, Rockwell International, “One-Quarter-Ton-Per-Hour Coal Hydropyrolysis Conversion Test Facility” 

September 15, 1978 Contract EX-77-C-01-2518 

4 S.J. Yosim and K.M. Bartley, “The Gasification of Various Coals in Molten Salts,” Rockwell International, Energy

Systems Group, Canoga Park, CA. 

5 United States District Court for the Central District of California, No. CV-97-1554. Undated. Document # 

BNA08759547. Retrieved from Department of Toxic Substances Control (DTSC) publicly accessible Historical 

Archives Database, File HDMS-01690476-122119551.pdf 

6 Pratt & Whitney, “Status of the Pratt & Whitney Rocketdyne/DOE Advanced Single Stage Gasifier Development 

Program,” Presented at the Gasification Technologies Conference October 12, 2005. File: 

Pratt_Rocketdyne_Coal.pdf 

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Section 17.1.1 

Building Number:

Building Alias: 

4048

048/T048 

Building Name: Plant Development Unit (PDU) 

Building Function: Instrumentation Building 

Radionuclides of Concern: None identified. 

17.1.1.1 Description of Operations and Processes 

Building 4048 appears to have been constructed in approximately 1978 and is identified in a 

1978 industrial planning map as the Plant Development Unit (PDU) instrumentation building.1 

The PDU was designed and operated by Rockwell for DOE to demonstrate the technical 

feasibility of producing sulfur-free, low-Btu product gas by partial combustion of Illinois No. 6 

coal in a sparged bed of molten, sodium carbonate salt. These operations are described in 

detail in Section 8.4 pertaining to Building 4005. 

The PDU was designated as a “non-radioactive” facility that operated from November 1978 to 

June 1981. 2 The MARSSIM Classification for Building 4048 is Class 1 due to its proximity to 

Building 4005 and location within ETEC. 

17.1.1.2 Radiological Incident Reports 

No radiological incident reports were located for this building. 

1 EPA/HGL, Final HSA 5-A, p. 148, citation 633:

632 SSFL Area IV, ETEC Industrial Planning Maps, 1962-1992. 

2 EPA/HGL, Final HSA 5-A, p. 148, citation 634:

634 Rockwell International, Environmental Monitoring Program Plan, Santa Susana Field Laboratory, Area 

IV, ER- AN-0006, September 30, 1992. 

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Section 17.1.2 

Building Number: 4049

Building Alias: 049/T049 

Building Name: Hydraulic Test Facility (1959)

Control Center, Piqua Test Loops (1968-1977)

PDU Control Center, Coal Gasification (1977-1988) 

Building Function: Various

Notes: Drainage to the Area II R-2 Pond 

Radionuclides of Concern: C-14, S-35, P-32, Fe-59, Co-60, Mn-54, Ni-59, Fe-55 

17.1.2.1 Description of Operations and Processes 

Building 4049 was constructed in 1959 as a hydraulic test facility control center for an outdoor 

vertical test stand (VTS). It was located within the fence-line boundary of Building 4005. 

According to a 1988 radiological survey of the building, the building was not associated with the 

SNAP program. The outdoor test stand was used for tests with terphenyl organics and finned 

sintered-aluminum-product cladding materials, sodium-water reaction tests, and a variety of 

sodium and NaK hydraulic tests. From 1968 to 1977, Building 4049 was used as a control 

center for Piqua Test Loops. In 1977, Building 4049 was designated as a control and test center 

1 for the PDU coal gasification process. An undated table listing buildings that generate waste at 

2 the SSFL site indicates that Building 4049 also served as a computer services building. By 

1988, according to the August 1988 radiological survey, the building was abandoned. Building 

4049 was demolished in 1999. 

An undated laboratory status report identified burning high boiler residue contamination at 

Building 4049 included the radionuclides of concern, listed above, in addition to activation 

corrosion products (ACPs). According to Boeing, ACPs include, at a minimum, Co-60, Mn-54, 

Ni-59, Ni-63, and Fe-55. 

The preliminary MARSSIM classification for Building 4049 is Class 1 because of previous 

incident reports, its location within ETEC, stains and open storage surrounding the building in 

aerial photographs, and the building’s location within the fence-line boundary of Building 4005. 

17.1.2.2 Radiological Incident Reports

There have been several incidents associated with Building 4049 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

1 EPA/HGL, Final HSA 5-A, p. 150, citation 640:

640 ETEC Document, GEN-ZR-0013, “Radiological Survey of Buildings T049, T042, T027, T032, and T025,” 

August 26, 1988. 

2 EPA/HGL, Final HSA 5-A, p. 150, citation 641:

641 Unknown, Table of SSF Waste Generators, Undated. 

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presented in an incidents database provided by Boeing.

17.1.2.3 Building 4049 Radiological Incident Report Summary - Data Provided by Boeing 

*Activation corrosion products have been defined as typically comprising Co-60, Mn-54, Ni-59, Ni-63, and Fe-55. 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0425 3/31/1960 Support Lab ACP* “Empty” pipe opened in clean lab 

contaminated employee with H.B.R.

A0479 4/1/1960 Building 4049 HBR ACP* Employee contaminated clothing while 

working around open HBR pipe.

A0476 4/17/1960 Building 4049 None Identified Vacuum surge tank sprayed on 

employees’ personal clothing.

A0358 4/27/1960 Pad SS None Identified HBR was sprayed on employee when it 

was opened.

• On March 31, 1960 a pipe containing high boiler residue was opened, with high boiler residue 

spilling onto someone’s shoes and pant legs. Activated corrosion product (ACP) contamination 

levels were recorded at 500 dpm/100 cm2. These ACPs have been defined by Boeing as 

typically comprising Co-60, Mn-54, Ni-59, Ni-63, and Fe-55 (A0425).1

• An incident occurred on April 1, 1960 during which an employee came in contact with an open 

pipe containing high boiler residue. According to the incident report, there were high boiler 

residue spots on his undershirt and work shirt. A smear of his skin showed no skin 

contamination and the shirt and undershirt showed no detectable contamination (A0479).2

• On April 19, 1960 machine oil from the vacuum surge tank sprayed an employee’s shirt. The 

shirt was surveyed and smeared and found to contain no detectable contamination of 

radiation. The shirt was returned to the employee (A0476). 3

• On April 27, 1960 an employee opened the high boiler residue line and had high boiler residue 

spray on the employee’s shoes, pants, and shirts. The articles of clothing were surveyed and 

smeared, and following cleaning, there was no contamination detected. The incident report did 

not indicate the contamination levels detected prior to cleaning, if any (A0358).4 

1 EPA/HGL, Final HSA 5-A, p. 150, citation 643:

643 Warren, J.W., Internal Letter Re: Radiological Safety Incident Report,A0425, April 22, 1960. 

2 EPA/HGL, Final HSA 5-A, p. 150, citation 644:

644 Warren, J.W., Internal Letter Re: Radiological Safety Incident Report A0479, May 19, 1960. 

3 EPA/HGL, Final HSA 5-A, p. 150, citation 645:

645 Warren, J.W., Internal Letter Re: Radiological Safety Incident Report A0476, May 19. 1960. 

4 EPA/HGL, Final HSA 5-A, p. 150, citation 646:

646 Warren, J.W., Internal Letter Re: Radiological Safety Incident Report, Building 049 Pad-SS, April 27, 

1960. 

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Section 17.1.3 

Facility Name:

Building Numbers/Function: 

SSFL Energy Bowl Area Test Facility: 

B-900 Control Room

B-901 

B-923 VTS I, VTS II, VTS III Coal Gasification

B-924 Bowl Steam Cont. Building

B-934 Bowl VTS-III Control Room 

Laser Engineering Test Facility (LETF)

Components Test Laboratory I (CTL-I)

Components Test Laboratory V (CTL-V) 

Operations/Processes: 1/4-Ton Per Hour Coal Hydropyrolysis Conversion Test Facility

1-Ton Per Hour Gasifier Reactor Development Program

4-Ton Per Hour Gasifier Reactor Development Program 

Contaminants of Concern: Possible Radium, Benzene, Coal Dust, Coal Tar, Coal Tar 

Creosote, Distillates, Sodium Hydroxide, Xylene (BTX), Char Residue 

17.1.3.1 Description of Operations & Processes 

The Bowl Area is comprised of three Bowl Area test-stand structures, a retention pond, a skim 

pond, a control center, and buildings used to support DOE coal gasification facilities. It was 

located within the ETEC operations area.1 The facility was used during the 1950’s and 1960’s to 

test rocket engines, and trichloroethylene (TCE) was used for each flushing (approximately 

50-100 gallons per rocket engine test). During the rocket engine testing period, the TCE and 2

water run-off from the flame bucket cooling and unused fuels were discharged to the unlined 

Bowl Skim Pond (capacity approximately 200,000 gallons). The pond was part of the Site-Wide 

Water Reclaim System, where Area IV industrial effluent, waste water, and reactor blowdown 

was directed. The surface of the skim pond was periodically ignited to allow the accumulated 

fuel to burn off. This pond discharged to the R-1 Pond (Area II) and ultimately to the Perimeter 

Pond (Area I) before draining off-site. 

In December, 1975 the Rockwell One-Quarter Ton Per Hour (1/4 TPH) Coal Hydroyrolysis 

Conversion Test Facility was activated by Rocketdyne to develop a hydrogenation process for 

coal liquefaction and gasification under a DOE-sponsored hydrogasification program.3 The 

facility was used to develop process system and subsystem design data required to advance 

the development of a single-stage hydrogasifier (reactor).4 

1 Author unknown. “29 DOE Buildings Smith Briefing / 30 Bldgs on IL’s from ETEC, Status as of 8/9/95”. Document 

retrieved from publicly accessible historical archives at California Department of Toxic Substances Control (DTSC). 

File: HDMSE00375150.pdf 

2 November, 1992. Author Unknown. “Bowl Area (Bowl Retention Pond, Bowl Skim Pond, and Bowl Test Stands,” 

November 1992. Document retrieved from publicly accessible Historical Archives Database, California Department of 

Toxic Substances Control (DTSC), File: HDMSp01739799.pdf 

3 DOE Contract #EX-77-C-01-2518 

4 Ibid. 122119551 

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In 1978, plans were begun to modify the facility toward further characterizing the performance 

and requirements of the reactor system, provide additional data for process optimization, and for 

design of the 4-TPH reactor that was slated for construction at the Bowl Area VTS-III facility in 

1979 under contract with DOE. By January, 1979 the 4-TPH Gasifier Reactor Development 

Program was underway. 

The expansion of the 1/4 TPH facility established a dual test position test facility, with provisions 

for a coal flow loop. The expansion allowed for additional contract and company-sponsored 

programs to be conducted simultaneously with the existing gasification reactor development 

contract. The expanded facility made use of common test and instrumentation systems for 

multiple shift operations for employees (i.e. first shift dedicated to DOE Reactor Development 

Program; second shift dedicated to the hydropyrolysis process applications and development. 

The facility was located at the SSFL Energy Bowl Area Test Facility.1 2 

According to a Rockwell International Annual Report, under contracts with DOE Rockwell 

advanced the development of two processes for the extraction of clean-burning gas (or liquid 

hydrocarbons) from coal. The process used a molten salt mixture at high temperature to convert 

any kind of coal to low or medium Btu industrial gas. The testing of the process progressed at a 

rate of 24-tons-per-day.3 This process is related to, and more fully described in Section 8.4 

related to Building 4005. 

From 1978-1983 the facilities were operated by Rocketdyne and DOE to demonstrate high 

temperature reaction of pulverized carbonaceous feedstocks with hydrogen in a short residence 

time reactor. A key element of the coal gasification reactor concept is the injector/reactor 

technology pioneered and developed by NASA and Rocketdyne. It was first used in liquid-fueled 

rocket engines. 

Waste was stored in drums at the floor of the Bowl Area, and were apparently stored onsite for 

some period before being disposed as hazardous waste. Disposal reportedly occurred some 

time between 1980-1983; Rocketdyne was unable to locate disposal manifests or other 

records.4 

There are indications that Rockwell-DOE coal gasification research at SSFL continued until 

approximately 2005, under Pratt-Whitney Rocketdyne.5 

No incident reports were located for the locations described above.

1 Rockwell International for DOE, “Advancement of Flash Hydrogasification: Task VIII — Performance Testing” by A.Y. 

Falk, M.D. Schuman, and D.R. Kahn. June 1986. Work performed under DOE Contract #DE-AC21-78ET10328 

2 United States District Court for the Central District of California, No. CV-97-1554. Undated. Document # 

BNA08759547. Retrieved from Department of Toxic Substances Control (DTSC) publicly accessible Historical 

Archives Database, File HDMS-01690476.pdf 

3 Rockwell International, Annual Report. Page 7. File: ESG_Report.pdf 

4 November, 1992. Author Unknown. “Bowl Area (Bowl Retention Pond, Bowl Skim Pond, and Bowl Test Stands,” 

November 1992. Document retrieved from publicly accessible Historical Archives Database, California Department of 

Toxic Substances Control (DTSC), File: HDMSp01739799.pdf 

5 Pratt & Whitney Rocketdyne DOE Advanced Single Stage Gasifier Development Program presentation, Gasification 

Technologies Conference, October 12, 2005. File: Pratt_Rocketdyne_Coal.pdf 

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17.2 SODIUM COMPONENTS TEST INSTALLATION 

The SCTI Building Complex and later Power-Pak installation included Buildings 4006, 4355, 

4356, 4357, 4359, 4360, 4361, 4392, 4457, 4656, 4756, and some facilities at the ETEC SSFL

Energy Bowl Facility. Although mostly designated as “non-nuclear” or “non-radiological”

facilities, some of the locations have documented use of radioactive materials and radiological 

incidents. Those locations with radiological incidents are included below, for addition to the 

SSFL Site Profile. 

The proximity of the SCTI and Power-Pak to nearby SNAP Buildings 4010, 4012, 4019, and 

4024 raised EPA’s concern about radioactive contaminant migration and airborne emissions. In 

addition, some SCTI buildings were constructed atop the former SNAP facilities that were noted 

to be extensively contaminated with radioactivity. 

The SCTI was a development test facility for liquid metal (sodium) system components. It 

provided a test site for the non-nuclear developmental testing of typical liquid metal reactor 

components, primarily steam generators. The SCTI Complex was originally designed to test 

sodium-heated steam generators and sodium-to-sodium intermediate heat exchangers under 

simulated sodium-cooled nuclear power plant operating conditions at a power level of 35 

megawatts thermal. It operated from 1964 through 1995 and underwent system upgrades to the 

steam and feed water systems in 1973. Further modifications in the late 1970’s and early 1980’s 

included the addition of a second sodium heater and the replacement of most of the steam/feed 

water system and the water make-up system. Operations included sodium tank cleaning, 

generation of steam from a sodium heat source, water treatment, X-Ray operations and 

radiography. 

In general, locations associated with the SCTI are designated “non-radiological.” However, EPA

included worker testimony from the DOE-EPA interview component of the Area IV Study and 

employees state that although they were not monitored for radiation exposure at this location, 

some of the tanks held radioactive materials. In addition, EPA based MARSSIM Classifications 

of the SCTI buildings on the lack of radiological investigations as well as concerns about the 

proximity of the SCTI Complex to SNAP Critical Testing Facilities that were noted to have 

extensive radiological contamination. Based on EPA’s assessment, the proximity of the SNAP

facilities presented potential radiological contaminant migration pathways via surface water run 

off and airborne releases. 

Due to the nature of SCTI operations, it was common for SCTI employees to be involved in 

DOE Coal Gasification processes at ETEC. It is common for employees of the SCTI and Coal 

Gasification facilities to have utilized Time Clock Locations in Area I. And, there is growing 

evidence, particularly during the Site Remediation period at SSFL, that SCTI employees should 

have been consistently monitored for radiation exposure. Under contract with DOE-ETEC, 

Boeing was required to provide all employees engaged in Site Remediation activities with 

adequate radiation protection. 

SCTI’s size, operations, processes and incidents necessitate its addition to the SSFL Site 

Profile to ensure a comprehensive characterization of facility operations.

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Section 17.2.1 

Building Number:

Building Alias: 

4356 

356/T356 

Building Name/Function: Sodium Components Test Installation (SCTI)

Steam Generation / Components Cleaning

X-Ray Operations 

Associated Buildings: Building 4656 Cooling Stacks 

Radionuclides of Concern: Sealed source use of Cs-137 and Ir-192 are noted below in 

Radiological Use Authorizations and Radiological Incident Reports. This suggests potential use 

of other sealed sources as well. Radionuclides associated with potential migration from Systems 

for Nuclear Auxiliary Power (SNAP) Buildings 4010 and 4012 include: Sb-125, Am-241, Cs-134, 

Cs-137, Co-60, Eu-152, Eu- 154, Pu-238, Pu-239, Pu-240, Pu-241, Sr-90, tritiated titanium H-3, 

and natural and enriched U-234, U-235, U-238 

17.2.1.1 Description of Operations & Processes 

Building 4356 was constructed in 1959 as the Sodium Components Test Installation (SCTI) High 

Bay. Although this location was designated as, “Non-Nuclear,” several radiological incidents are 

noted below. Workers at this location were not considered to be “radiation workers,” and were 

not routinely or consistently monitored for radiation exposure. 

Building 4356 did not contain any restrooms or heating / ventilation features. The ceiling height 

was 65 feet. The HSA provides a detailed description of the SCTI’s features and schematics, 

and references four sodium drank tanks; a reaction tank in below-grade concrete pits; 22 

aboveground tanks containing sodium, sulfuric acid, calcium hydroxide, sodium hydroxide, 

morpholine, brine, gaseous argon, hydrazine, deionized water, and diesel oil. 

The SCTI was a development test facility for liquid metal (sodium) system components. It 

provided a test site for the non-nuclear developmental testing of typical liquid metal reactor 

components, primarily steam generators. The SCTI Complex was originally designed to test 

sodium-heated stem generators and sodium-to-sodium intermediate heat exchangers under 

simulated sodium-cooled nuclear power plant operating conditions at a power level of 35 

megawatts thermal. It operated from 1964 through 1995 and underwent system upgrades to the 

steam and feed water systems in 1973. Further modifications in the late 1970’s and early 1980’s 

included the addition of a second sodium heater and the replacement of most of the steam/feed 

water system and the water make-up system. Operations included sodium tank cleaning, 

generation of steam from a sodium heat source, water treatment, X-Ray operations and 

radiography. 

Radiological Use Authorizations for this location include the permitted use of two 250 μCi 

Cs-137 sealed sources and one 100 μCi Cs-137 sealed source that were used as sodium level 

gauges. Sealed source use of Cs-137 and Ir-192 are noted at this location, and suggest 

potential use of other sealed sources as well. Additionally, radionuclides associated with 

potential migration from SNAP Buildings 4010 and 4012 include Sb-125, Am-241, Cs-134, 

Cs-137, Co-60, Eu-152, Eu-154, Pu-238, Pu-239, Pu-240, Pu-241, Sr-90, H-3, and natural and 

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enriched U-234, U-235 and U-238. Because the general slope of Area IV is in a southerly 

direction, runoff from SNAP Buildings 4010 and 4012 could potentially impact Building 4356. 

The SNAP buildings handled radioactive materials and generated radionuclides, which means 

there is a possibility, according to the EPA research team, of radiological contamination in the 

area of Building 4356. 

The preliminary MARSSIM classification of Building 4356 is Class 1 because of sealed 

radioactive material sources that were used at this location, and the potential for radioactive 

material migration via surface water flow and airborne releases from SNAP Building 4010 and 

4012, in addition to the lack of site investigation. 

17.2.1.2 Building 4356 Radiological Incident Report Summary - Data Provided by Boeing

There have been several incidents associated with Building 4356 that could have resulted in a 

release to the environment and worker exposure. The following table provides information 

presented in an incidents database provided by Boeing.

17.2.1.3 Radiological Incident Reports - Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0639 10/7/1974 SCTI TANK #5 Cs-137 Source removed from SCTI surge tank 

and stored without approval.

A0093 12/8/1981 TEST STAND Ir-92 Radiographic source pigtail uncoupled 

and fell to deck.

A0138 5/11/1985 SCTI 3RD DECK Ir-92 Source stuck in short radium tube during 

radiographic operation at SCTI

A0153 2/20/1986 SCTI Ir-92 Radiography barrier removed; 

employees entered area.

• On December 8, 1981, an iridium-192 (Ir-192) radiographic source pigtail uncoupled and fell to 

the deck of the SCTI test stand. A pigtail is a source assembly that consists of the sealed 

source and a connector that attaches the source to the control cable. The HSA research team 

could not locate the original incident report (A0093).1

• On May 11, 1985 an employee gammagraphing between the 2nd and 3rd deck of the SCTI 

could not retrieve the 45 curie (Ci) Ir-192 source. The source had become stuck in the short 

radium tube during radiographic operations. The employee was advised to stay cleaer of the 

area and security and management were notified. Survey meters and dosimeters were 

brought to Building 4356 and security road blocks were put in place. The dose rate at the 

radiographer’s barrier rope on the east side indicated 12 mR/hr. The radiographer pointed out 

the source location below the floor on the 3rd deck along the steam generator. The dose rate 

was approximately 3 R/h at 4 feet from the back side of the source. A bend in the tube was 

pried back to allow the source to be cranked back into the lead camera. One employee 

1 EPA/HGL, Final HSA 5B, p. 67, citation 3:

1 Interview No. 279 conducted by DOE in 2010. 

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received 2 mrem during the source recovery for a total of 13 mrems for the day. The health 

physicist on site received 18mrem during the recovery (A0138).1 

1 EPA/HGL, Final HSA 5B, p. 68, citation 1:

1 Internal Correspondence from Badger, F.H. to Radiation & Nuclear Safety Group, Rockwell International, 

Re: Radiological Safety Incident Report, T-356 SCTI, May 11, 1985, May 15, 1985. 

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Section 17.2.2 

Building Number:

Building Alias: 

4006 

06/T006 

Building Name/Function: Sodium Laboratory 

Radionuclides of Concern: Building 4006 was predominantly a non-radiological facility, but 

there are records of minor uses of radioactive materials, including UO2, Mn-54, H-3 foils in gas 

chromatographs, and sodium loop level gauges possibly employing cesium-137 (Cs-137) 

sources. Radionuclides that may be present as a result of residual radioactivity in Building 4006 

include: Cs-137, Mn-54, H-3, and natural and enriched uranium (U-234, U-235, U-238). 

Radionuclides associated with potential migration from Systems for Nuclear Auxiliary Power 

(SNAP) Buildings 4010, 4012, and 4024 include: Sb-125, Am- 241, Cs-134, Cs-137, Co-60, 

Eu-152, Eu-154, Pu-238, Pu-239, Pu-240, Pu-241, Sr-90, H-3, and U-234, U-235, U- 238. 

17.2.2.1 Description of Operations & Processes 

Building 4006 was constructed between 1957 and 1959 as a sodium laboratory. Its nearby 

Cooling Tower (Building 4616) was demolished in the 1980’s to make room for a Power-Pak 

associated substation. The HSA provides a detailed description of Building 4006’s features and 

schematics. 

Building 4006 was considered a “non-nuclear” sodium laboratory. However, Use Authorizations 

provided for use of radioactive materials. In addition to a documented radiological incident 

contained in the Incident Reports Database, a 2008 Resource Conservation and Recovery Act 

(RCRA) Facility Investigation Report (RFI) describes other incidents associated with Building 

4006 that include a sodium fire on November 25, 1987 resulting in a release of sodium metal, a 

mercury spill in a sink on January 21, 1992, and a tetralin fire on August 16, 1959.1 

• 1973 Radiological Use Authorization No. 66 (September 28, 1973) specified that sodium (Na) 

would be added to canisters containing uranium oxide (UO2) in Building 4006.2 

• 1974 Radiological Use Authorization No. 81 (June 26, 1974) permitted the use of tritiated 

titanium (H-3) foils as gas chromatograph detectors, used at the location until 1986.

• 1976 Radiological Use Authorization No. 101 (April 8, 1976) permitted the handling of 0.5 μCi 

of manganese-54 (Mn-54) contained in sections of activated piping with frozen sodium. The 

unpacking of the pipes occurred in Building 4006. 

The preliminary MARSSIM classification of Building 4006 is Class 2 because regulated 

radiological material was handled at this location; Building 4006 aerial photographic analysis 

1 EPA/HGL, Final HSA 5B, p. 103, citation 3:

3 Group 5 – Central Portion of Areas III and IV RCRA Facility Investigation Report, Santa Susana Field 

Laboratory, Ventura County, California, Volume V – RFI Site Reports, Appendix I, Coal Gasification Process 

Development Unit, CH2M Hill, Draft in Progress November 2008, p. Table I.2-1. 

2 EPA/HGL, Final HSA 5B, p. 104, citation 4:

4 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. O-5. 

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noted possible staining; there is potential for radioactive material migration via surface water 

flow and airborne release from SNAP Buildings 4010, 4012, and 4024. 

17.2.2.3 Radiological Incident Reports Summary - Provided by Boeing 

There has been one incident associated with Building 4006 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing. 

17.2.2.4 Radiological Incident Reports - Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Istopes Description of Incident

A0484 8/11/1960 BLDG. MET. 

LAB

Activation 

Products

Polishing samples of SRE Moderator Can 

contaminated “clean” area.

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Section 17.2.3 

Building Number:

Building Alias: 

4355 

355/T355 

Building Name/Function: Sodium Components Test Installation - Control Center SCTI

Small Components Test Loop

17.2.3.1 Description of Operations & Processes 

Building 4355 was the SCTI Control Center. It was used to monitor and control operations in 

Building 4356. It contained the control room, data acquisition system, offices for facility 

management and staff, restrooms, a small chemistry laboratory, and facility record storage.1 

During the DOE-EPA Employee Interviews, Employee No. 195 indicated that, although workers 

at this location were not required to wear monitoring or film badges, some of the sodium tanks 

had radioactive material “to measure the level of the materials in the tanks.” Other employees 

indicate their need for radiation badges at other parts of the site, but as employees assigned to 

SCTI they were considered “non-nuclear” workers.2 

Radiological use authorizations permitted the operation of Bowed Tubes Measurement. 

Authorization 117D, dated July 1, 1984, permitted the use of a 1.0 microcurie cobalt-60 (Co-60) 

sealed source that was checked annually to ensure that no leakage occurred.3 

The MARSSIM Classification of Building 4355 is Class 1 because of its use of sealed 

radioactive material sources, and the potential for radioactive contaminant migration via surface 

water flow or airborne release from SNAP Buildings 4010 and 4012. According to EPA, because 

the general slope of Area IV is in a southerly direction, runoff from SNAP Buildings 4010 and 

4012 could potentially impact Building 4355. 

17.2.3.2 Radiological Incident Reports

There has been one incident associated with Building 4355 that could have resulted in a release 

to the environment and worker exposure. The following table provides information presented in 

an incidents database provided by Boeing.

1 EPA/HGL, Final HSA 5B, p. 60, citation 1:

1 Kneff, D.W. et al., Sodium Component Test Installation (SCTI) Demolition Final Report, EID-08336, 

October 1, 2003, pgs. 13-14. 

2 EPA/HGL, Final HSA 5B, p. 60 & 61, citations 3-1:

3 Interview No. 195 conducted by DOE in 2010.

1 Interview No. 63 conducted by DOE in 2010. 

3 EPA/HGL, Final HSA 5B, p. 61, citation 6:

6 Review of Radiation Safety Records Management System, 2010. 

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17.1.2.3 Building 4355 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Istopes Description of Incident

A01044 4/15/1988 Outside Barrier Ir-92 passed through 

radioactive barrier of radiographer.

[Redacted]

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Section 17.2.4 

Building Number: 4357 

Building Alias: 357/T357 

Building Name/Function: SCTI Heat Transfer Loop Control Building 

Radionuclides of Concern: Sealed source use of Cs-137 & Ir-192 suggests potential use of 

other sealed sources as well. Radionuclides associated with SNAP Buildings 4010 and 4012 

present concerns at this location.

17.2.4.1 Description of Operations & Processes 

Building 4357 was constructed in 1958 as the Heat Transfer Loop Control Building, and later a 

Pump Bearing Test Facility Control Building for the Liquid Metal Engineering Center (LMEC) and 

Energy Technology Engineering Center (ETEC). It provided a test site for the non-nuclear 

developmental testing of typical liquid metal reactor components, primarily steam generators.1 It

later became part of the Sodium Component Test Installation (SCTI) Complex, serving as a 

storage building. 

17.2.4.2 Radiological Incident Reports

There have been no radiological use authorizations or incident reports found for this location. 

1 EPA/HGL, Final HSA 5B, p. 71, citations 7 & 8:

7 Sapere Consulting, Inc. and The Boeing Company, Historical Site Assessment of Area IV Santa Susana 

Field Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. P-25.

8 Kneff, D.W. et al., Sodium Component Test Installation (SCTI) Demolition Final Report, EID-08336, 

October 1, 2003, pgs. 5, 13. 

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Section 17.2.5 

Building Number:

Building Alias: 

VTS-III (ETEC SSFL Energy Bowl Area) 

Steam Accumulation Blowdown Evaluation Rig (SABER)

Building Function: Large Scale Steam Valve Testing 

Associated Buildings: Building 923 - Civil Works / Structure

Building 934 - Control Room

Building 901 - Pre-Test Building

Building 924 - Steam System 

17.2.5.1 Description of Operations & Processes 

The Steam Accumulation Blowdown Evaluation Rig (SABER) was located in the ETEC area 

known as the SSFL Energy Bowl Area. Its purpose was to conduct the Large Scale Steam Valve 

Test (LSSVT) to demonstrate large steam Valve Test Article performance, and to obtain twophase flow design data that would otherwise be unavailable. The test system design, 

procurement and construction management was performed by ETEC.1 

The program involved searching for a test facility that could supply the steam flow and steam 

quality conditions required for the valve tests. The search identified the SABER in the SSFL

Energy Bowl Area as the only existing site that could meet the required test conditions. 

This facility supported DOE POWER-PAK operations. In 1986, ETEC / Rocketdyne / DOE 

installed the Power-Pak Facility at the SCTI for the purpose of supplying 25 megawatts of 

electrical power, which was intended to be sold by the government to the Southern California 

Edison Company. Because the new equipment would become an “integral part of [the] existing 

steam generator test complex and, as such, a specialized test facility,” no plans were made to 

submit designs for checks and County building permits were not requested, according to 

Rockwell International. It is assumed that this precedent prevailed for other operations that met 2

this criteria. 

There were no incident reports discovered for this location.

1 Steam Accumulator Blowdown Evaluation Rig, Large Scale Steam Valve Test Action Description Memorandum. 

Document retrieved from publicly accessible Historical Archive courtesy of California Department of Toxic Substances 

Control (DTSC). File #HDMS00019780.pdf 

2 Internal Letter, Rockwell International to Ventura County. “In Reply, Refer to 86RC00245,” January 10, 1986.

Document retrieved from publicly accessible historical archives courtesy of California Department of Toxic 

Substances Control (DTSC). File: HDMSE00024518.pdf 

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17.3 SMALL COMPONENTS TEST LOOP (SCTL) 

BUILDINGS: 4026, 4226, 4826, 4358 

The Small Component Test Loop (SCTL) Complex is comprised of the four buildings listed 

above, and may involve other Area IV locations. Research on the SCTL for inclusion to the 

SSFL Site Profile is incomplete. CORE Advocacy has included some buildings associated with 

SCTL operations where radiological incidents are documented, and will update this section of 

the 2016 Site Description as more information is obtained. 

Building 4358 was originally built to support the Sodium Component Test Installation (SCTI) and 

was later moved to support the SCTL. The SCTL originally began as the Large Component Test 

Loop (1957). In 1970, it was modified and became the SCTL. 

The SCTL complex consisted of an enclosed component test area, a sodium transport and 

storage system, an instrument and control system, and other interfacing systems. The sodium 

transport and storage system consists of four test loops, a thermal conditioning loop, and two 

liquid sodium storage/drain tanks that act as reservoirs for normal and emergency draining. 

The SCTL’s principal purpose was to provide a test bed for non-nuclear qualification testing of 

typical liquid metal fast breeder reactor components and to obtain test data for verification of the 

elevated temperature design criteria of the piping design guide. In 1974, the facility was again 

modified and expanded. In 1978, a new control room was added, giving the complex its final 

appearance. 

Testing within the facility continued periodically from 1959 until 1985. The facility was 

maintained in inactive “standby” status, with the sodium systems under an inert cover gas, until 

late 1995. 

Although designated a “non-nuclear” or “non-radiological” facility, radiological incidents are 

documented. Its proximity to known sources of radioactivity and radionuclide generation (SNAP

Buildings 4010, 4012, 4019, 4024) may have resulted in radioactive contaminant migration due 

to surface water flow and airborne releases. 

Worker records show that employees associated with the SCTL, designated as “non-radiation” 

workers, were frequent participants in Site Remediation activities that should have necessitated 

consistent radiation monitoring. 

The SCTL should be included in the SSFL Site Profile to ensure an accurate characterization of 

facility operations. 

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Section 17.3.1 

Building Number:

Building Alias: 

4026 

026/T026 

Building Name/Function: Small Components Test Loop (SCTL)

Control Building

Associated Buildings: Building 4726 Electrical Cooling Station

Building 4805 Time Clock Shack

Building 4426 Uninterruptible Power Supply (UPS) 

17.3.1.1 Description of Operations & Processes 

Building 4026 was originally built as the Large Components Test Loop. Operations began in 

1959 with a pump test. This facility used to test small components such as valves and pumps in 1

liquid sodium. 2 The HSA provides a detailed description of Building 4026’s features and 

schematics. 

Although designated a “non-radiological” facility, radiological incidents are contained in the 

Incident Report Database that suggest the use of sealed sources at Building 4026. Aerial 

photographs of this location depict staining and a probable open storage area. 

The preliminary MARRSIM classification for Building 4026 is Class 2 due to sealed source use, 

documented incidents, lack of site investigation, and potential radioactive material migration via 

surface water flow or airborne releases from SNAP buildings nearby. 

17.3.1.2 Radiological Incident Reports - Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Istopes Description of Incident

A0465 9/10/1969 Gammagraph Ir-192 Employee disregarded warning signs and 

entered field of radiographic operation.

A0238 10/3/1979 SCTL Ir-192 Stuck gammagraph source recovered.

1 EPA/HGL, Final HSA 5B, p. 126, citation 7:

7 U.S. Department of Energy, Small Component Test Loop, Energy Technology Engineering Center (ETEC) 

Website, http://www.etec.energy.gov/History/Sodium/SCTL.html, accessed November 25, 2009, p.1. 

2 EPA/HGL, Final HSA 5B, p. 126, citation 8:

8 Rutherford, P.D., Site Environmental Report for Calendar Year 1998 , DOE Operations at Rocketdyne 

Propulsion & Power, RD99-115, The Boeing Company, Rocketdyne Propulsion & Power, September 22, 

1999, p. 2-10. 

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17.4 DOE LASER ENGINEERING 

SSFL-ETEC Energy Bowl Facilities were involved in the research and development of laser 

technology: The Hydrogen Fluoride Chemical Laser, the Sigma Tau Laser, the “RACHL,” and 

others. Some of these projects involved DOE. Others appear to have been carried out in 

facilities owned or operated by DOE, or situated on DOE-optioned land, although DOE 

involvement and proprietary interests in the laser engineering program remains unclear at this 

time. This section is incomplete. CORE Advocacy will provide more information on the SSFL

ETEC Laser Engineering Facilities as it is obtained. 

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Section 17.4.1 

Building Number:

Building Alias: 

LETF 

Laser Engineering Test Facility

The “RACHL” 

Building Function: Hydrogen Fluoride Chemical Laser 

17.4.1.1 Description of Operations & Processes 

The Laser Engineering Test Facility (LETF) was located in the ETEC area known as the SSFL

Energy Bowl Area. The program required Rockwell International to develop a fluorine atom 

diagnostic and a fluorine molecule diagnostics system. Modification of the facility was to 

incorporate a Steam Ejector system. 1 There were several entities involved in the RACHL, 

including Rockwell International, Department of Defense (DOD) and DOE. The LETF Building 

was later used for DOE coal gasification operations. 

1 Internal Letter, Rockwell International, “Hydrogen Fluoride Chemical Laser Technology,” June 30, 1976. Retrieved 

from publicly accessible Historical Document Archive, courtesy of Department of Toxic Substances Control (DTSC). 

File #HDMS200410653.pdf 

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17.5 Miscellaneous Additions - Various Processes and Operations 

Several facilities that participated in various processes or serviced and supported diverse 

programs over the decades of SSFL’s operations were excluded from the 2006 Site Profile. On 

the following pages, CORE Advocacy has identified a partial list of the excluded locations, and 

provided descriptions of their operations and processes. As more information is obtained, this 

section will be updated accordingly. 

Some of these locations may have been designated “non-radiological” facilities but could have 

handled radioactive materials at some point in time. In addition, regardless of facility designation 

there may be documented incidents involving radiological materials. The locations of some of 

the facilities and their proximity to various sources or generators of radioactivity may result in 

concerns about worker exposure, regardless of the facility designation or processes. 

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Section 17.5.1 

Building Numbers:

Building Alias: 

4273 / 4283

273/T273 - 283/T283 

Building Name/Function: Radioactive Laundry 

Associated Buildings: Building 4316 Maintenance Skid Shack 

Notes: Tritium

Drainage to the Area II Ponds 

Radionuclides of Concern: U-238, U-234, U-235, U-236, Pu- 239, Pu-240, Pu-241, Pu-242, 

Am-241, Th-232, H-3, Na-22, Na-24, Cr-51, Mn-54, Fe-59, Co- 60, Kr-85, Sr-89, Sr-90, Sb-125, 

I-129, I-131, Cs-134, Cs-137, Ce-141, Ba (La)-140, Nb-95, Ru-103, Ru-106, Xn-133, Xe-135, 

Pm-147, Sm-151 

17.5.1.1 Description of Operations & Processes: 

Buildings 4273 and 4283 were constructed in approximately 1957 for use as a radioactive 

laundry. 1 The building included a radioactive laundry and protective clothing storage area. 

Contaminated laundry from the SRE Building 4143, the Engineering Test Building 4003, and the 

RMHF Building 4021 was brought to this location for cleaning. Operations were discontinued in 

1971.2 

The HSA research team could not locate a more detailed description of these buildings. These 

buildings may have had a septic tank and a leach field. 

The preliminary MARSSIM classification for Buildings 4273 and 4283 area is Class 1. Items of 

radioactive laundry were washed and dried in Buildings 4273 and 4283 for approximately 14 

years. It is not apparent where wastewater from laundry operations discharged or that a full 

radiological characterization of the area was conducted. 

17.5.1.2 Building 4273 Radiological Incident Reports 

None found.

The HSA provides a detailed chronology of previous radiological investigations and 

decontamination/cleanup of releases. DOE released these locations for unrestricted use in 

1985.

1 EPA/HGL, Final HSA 6, p. 146, citation 1: 

1 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. 

2 EPA/HGL, Final HSA 6, p. 146, citation 2: 

2 Chapman, J. A., Radiological Survey of the T513 Parking Lot; Old R/A Laundry Area; Plot 333; and Areas 

between the SRE to RMDF, and KEWB to RMDF, Energy Technology Engineering Center Report No. GENZR- 0009, August 26, 1988, pp. 8, 19, 54, 61, 63-64, 77, 79. 

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Section 17.5.2 

Building Number:

Building Alias: 

4133 

133/T133 

Building Name: Hazardous Waste Materials Facility (HWMF) 

Building Function: Contaminated Sodium Cleaning

Burn Pit Facility (Waste Disposal)

Drum Storage 

Radionuclides of Concern: Although Building 4133 was considered a non-radiological site, 

low levels of Cs-137 were discovered here. Radiological surveys were conducted at the building 

area because of inadvertent radiological contamination of the facility (Building 4029) that 

provided the residual sodium for treatment in Building 4133.1 

17.5.2.1 Description of Operations & Processes: 

This facility is technically considered to be a “non-radiological” facility, and was not included in 

the 2006 SSFL Site Description or Site Profile. Boeing Employment Summaries may indicate 

that employees assigned to this building were not radiation workers, not monitored for radiation 

exposure, or were not at risk of being exposed to radiation. In addition, the processes and 

potential releases resulting from the operations of the HWMF may not be appropriately included 

in SSFL data used in dose reconstruction. However, Building 4133 received radioactively 

contaminated components as standard operating procedure.2 

A portion of Building 4133 was initially constructed as Building 4724, the Contaminated Sodium 

Cleaning Facility used to support the Sodium Reactor Experiment (SRE). When Building 4724 

was closed, the upper portion of the building was decontaminated and released for unrestricted 

use. This portion of the building was used for the construction of Building 4133. 

In December 1977, Building 4133 was constructed at its current location. It has been identified 

as the Sodium Burn Facility on industrial planning maps, but also is collectively referred to as 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 180, citation 3:

3 Morton, J., Draft Verification Survey of Building 4059 (Phase 1), Building 4133, and the 17th Street 

Drainage Area, Santa Susana Field Laboratory, The Boeing Company, Ventura County, California, Oak 

Ridge Institute for Science and Education, February 2000, pgs 2-3. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 173, citations 5 & 6: 

5 B/133 Status Report, Unknown Author, March 15, 1993 through October 29, 1993, HDMSe00381574 – 

HDMSE 00381640.

6 Soucy, R.C., ETEC Waste Verification Plan for the Hazardous Waste Management Facility (HWMF), GENAN- 0041, Energy Technology Engineering Center, November 29, 1993, p. 5. 

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the Hazardous Waste Management Facility (HWMF) with Building 4029. Unlike the Sodium 

Burn Pit, this facility treated sodium and other metals inside the building structure.1 

Building 4133 was designed for the treatment of non-radioactive alkali metals, including sodium, 

sodium-potassium alloy (NaK), and lithium. The processes used to size-reduce solid 

components resulted in metal oxides that were removed from the air by a Peabody Variable 

Throat Venturi air scrubber and spray tower located outside Building 4133. Rinse water drained 

to a neutralizing pit or storage tank, and could be pumped to the aboveground tank when the 

other was full. Substances that may be contained in the scrap materials to be treated in the 

facility include: sodium, lithium, potassium, NaK, and lithium hydride. Building 4133 was also 

used as a drum and equipment storage area in the 1960’s and 1970’s.2 The RFI report also 

notes that prior to 1978, the Building 4133 area was used for parking and equipment/drum 

storage in support of the Sodium Reactor Experiment (SRE) operations. 

The HWMF began operation in 1978 as a drum storage yard and was fully permitted in 1983 as 

a RCRA hazardous waste treatment and storage facility for non-radiological chemical wastes 

generated onsite. However, status reports for 1993 indicate that while operating as the HWMF, 

Building 4133 received pipes and tubes that were radioactively contaminated above background 

levels. Standard operating procedure at Building 4133 dictated that if an object with sealed, 

inaccessible, internal surfaces was to be cut open at Building 4133, then a radiation survey 

overcheck was to be performed during and after cutting operations. Any material found to have 

radioactive contamination was transferred to the Radioactive Material Handling Facility 

3 (RMHF). A 1993 Rockwell letter states that water used to wash buried objects from the Building 

4886 Sodium Disposal Facility at Building 4133 was found to contain low concentrations (less 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 160, citations 1-6: 

1 

Sapere Consu t ng, Inc. and The Boe ng Company, Historical Site Assessment of Area IV Santa Susana Field 

Laboratory, Ventura County, California, Volume 2 – Area IV Site Summaries, May 2005, p. G-11.

2 3

McG nn s, E.R., Building 4133 Radiation Survey Report, RS-00015, The Boe ng Company, January 14, 2004, p. 8. 

Santa Susana Area IV, Atom cs Internat ona /Energy Systems Group P ann ng Maps, March 1962–November 1992.

4 

Knudsen, K.T., Safety Analysis Document – Building 133 Hazardous Waste Management Facility, 133-ZR-0003, Rev. C, 

Energy Techno ogy Eng neer ng Center, June 28, 1991, Rev sed Apr 1, 1996, p. 5.

5 

The Boeing Company Closure Plan Hazardous Waste Management Facility, Buildings T029 and T133, Santa Susana 

Field Laboratory, Ventura County, California, MWH, December 2003, p. 3-5.

6 

MWH, Group 7 – Northern Portion of Area IV RCRA Facility Investigation Report, Santa Susana Field Laboratory, 

Ventura County, California, Volume II – RFI Site Reports, June 2009, pg. B.2-2.

7 

Energy Techno ogy Eng neer ng Center, Operation Plan: Hazardous Waste Management Facility, 133-AN-0001, March 

3, 1991, p. 25. 

2 EPA/HGL, Final HSA 7-3-NBZ, p. 173, citation 7:

7 

Science Applications International Corporation, Final RCRA Facility Assessment (RFA) Report for Rockwell 

International Corporation, Rocketdyne Division, Santa Susana Field Laboratory, Ventura County, California, 

May 1994, p. 7-3. 

3 EPA/HGL, Final HSA 7-3-NBZ, p. 173, citations 5 & 6: 

5 B/133 Status Report, Unknown Author, March 15, 1993 through October 29, 1993, HDMSe00381574 – 

HDMSE 00381640.

6 Soucy, R.C., ETEC Waste Verification Plan for the Hazardous Waste Management Facility (HWMF), GENAN- 0041, Energy Technology Engineering Center, November 29, 1993, p. 5. 

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than 10 pCi/L) of cesium-137 (Cs-137), suggesting a means of potential radioactive 

contamination at Building 4133.1 

The preliminary MARSSIM classification for the Building 4133 area is Class 1, due to the 

possibility of cutting of radioactive contaminated pipes and tubes, and proximity to other 

radiological facilities. 

1 EPA/HGL, Final HSA 7-3-NBZ, p. 174, citation 7:

7 Internal Correspondence from Tuttle, R.J. to Rutherford, P.D., Rockwell International, Re: Disposal of

Potentially Contaminated Wash Water at T133, June 3, 1993. 

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Section 17.5.3 

Building Number:

Building Alias: 

4011 

011/T011 

Building Name/Function: Radiation Instrument Calibration Laboratory 

Associated Buildings: Building 4403 Traffic Dispatch Building

Building 4711 Electrical Substation 

Notes: Possible Tritium 

Radionuclides of Concern: Am-241, Cs-137, Co-60, Eu-152, Eu-154, Ir-192, Pb-210, Pu-238, 

Pu-239, Pu-240, Pu-241, K-40, Ra-226, Sr-90, Ta-182, Tc-99, Th-230, Th-232, H-3, and U-234, 

U-235, U-238. 

17.5.3.1 Description of Operations & Processes 

Building 4011was used to support various non-nuclear programs until 1984. From 1984 to 1996, 

the north section of Building 4011 was used for calibration and repair of radiation detection 

instruments. Calibration and repair would have involved instruments containing radioactive 

contaminants from any of the Area IV nuclear facilities. Service work used only sealed 

radioactive sources and electroplated calibration sources. After 1996, laboratory activities were 

transferred to Building 4100 (AETR). 

The Property Inventory and Control Department used the south section of the building. The 

south side of Building 4011 was used as a materials warehouse for non-radiological materials. 

Industrial planning maps indicate that Building 4011 was an administration and services building 

in the late 1960’s and early 1970’s, a development support shop in the mid-1970’s, a 

manufacturing support shop in the late 1970’s and early 1980’s, a machine shop in the mid-tolate 1980’s, and a storage facility in the early 1990’s.1 The HSA provides a detailed description 

of Building 4011 features and schematics. 

The preliminary MARSSIM classification for Building 4011 is Class 1 because of sealed and 

unsealed radioactive material sources, reported radiological incidents, and previous 

investigations. 

17.5.3.2 Radiological Incident Reports - Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Istopes Description of Incident

A0531 4/26/1960 West of Bldg.11 MFP Leak test of OMRE shipping cask spilled 

radioactive liquid on the ground.

1 EPA/HGL, Final HSA 5B, p. 184, citation 4:

4 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. 

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*Isotopes are listed as written in the Incident Report Database. The HSA research team believes that “MFP” is an 

acronym for “mixed fission products.” 

A0519 7/14/1964 SS 19th & F St. Empty but labeled radioactive shipping 

containers found in private car.

A0083 7/2/1980 RIHL [Hot Lab] 

and Bldg. 11

Ta-182 Unexplained exposure, over guidelines.

A0318 4/13/1985 Calibrator Cs-137 Calibration source came lose from actuator 

rod.

A0651 1/6/1994 Inst. Calb. Lab Internally contaminated air sampling pumps 

not labeled.

A0658 12/6/1994 T011 Calib. Lab Cs-137 28 Ci Cs-137 calibration source dislocated 

from release pull rod.

Incident File 

Name

Date of 

Incident

Location of 

Incident

Istopes Description of Incident

• On April 28, 1960 during the final leak test of the Organic Moderated Reactor Experiment 

(OMRE) trailer cask cooling system, a union between the two circulating lines was not 

tightened. When the pump was turned on, water ran out from the system onto the trailer bed 

and blacktop through bolt holes where the cask was secured. The water was soaked up 

immediately and confined to an area of about 1 square foot. Approximately 1 pint of 

radioactive liquid containing mixed fission products was spilled on the ground to the west of 

Building 4011. All loose dirt was cleaned up and a s ample of the blacktop was taken. Results 

shoes 216 dpm beta-gamma. The incident report notes that this type of operation should have 

been done in an area that had the proper facilities for handling contaminated equipment.

(A0531).1

In addition to the incidents noted in Boeing’s Incident Database, the HSA research team 

identified two other incident pertaining to Building 4011. 

• Sludge in a sink trap in Room 120 (Building 4011) was found to contain 27.2 pCi/g of U-234. 

The sink and trap and a portion of the drain line were removed and disposed of as radioactive 

waste.2 It was not clear to the HSA research team from figures included in the HSA whether 

the drain line was connected to the former septic tank, or the sewer line.

• On January 15, 1976 a radioactive spill in the Rocketdyne conservation yard was discovered. 

In the search for the origin and the extent of contamination, numerous radioactively 

1 EPA/HGL, Final HSA 5B, p. 185, citation 1:

1 Klostermann, J.P., Radiological Safety Incident Report, Bldg. #11 (West Side), Atomics International, April 

28, 1960. 

2 EPA/HGL, Final HSA 5B, p. 185, citations 2, 3, &4:

2 Barnes, J., Final Radiological Survey Data Package for Building 011, Santa Susana Field Laboratory, The 

Boeing Company, July 28, 1998.

3 Rash, M., Final Oversight Verification and Confirmation Radiological Survey Report for Buildings T-011, 

T-019, T-055, and T-100, Tetra Tech EM, Inc., December 20, 2002, p. 9.

4 Rash, M., Final Rocketdyne Technical Support and Field Oversight Document Review for Buildings T009, 

T011, T019, T055, and T100, Tetra Tech EM, Inc., December 20, 2002, pgs 8-9. 

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contaminated barrels and pallets were recovered from a wide area of the Santa Susana Field 

Laboratory (SSFL) complex on the whole. A pallet in the Building 4011 receiving yard with two 

acetone barrels on it was discovered to be radiologically contaminated at approximately 800 

mrad/hr beta. According to the incident report, all contaminated items that had been detected 

were transferred to the Radioactive Materials Handling Facility (RMHF) for disposal with the 

exception of three large concrete structures and a 1 mrad/hr spot on the asphalt in the 

Building 4011 yard. The incident report does not describe how these items were handled. 

Externally contaminated barrels appeared to have resulted from a double-stacked barrel of 

radioactive liquid that corroded and leaked. 

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Section 17.5.4 

Building Number: 4641 

Building Alias: 641/T641 

Building Name/Function: Shipping and Receiving 

Notes: Building 4641 served as a transfer point for all SSFL incoming/outgoing 

shipments. 

Radionuclides of Concern: All radionuclides used at SSFL Area IV should be considered 

potentially present at any shipping/receiving or waste disposal location (i.e. RMHF). 

17.5.4.1 Description of Operations & Processes 

Building 4641 was constructed in 1964 to be used for shipping and receiving. It served as a 

transfer point for all SSFL incoming and outgoing shipments. Through 1985, this included 

radioactive materials. However, it should be noted that the documented incident below involved 

radioactive materials in 1989. 

Non-radioactive materials were reportedly stored in the warehouse, while radioactive and 

nuclear shipments were reportedly only handled on the outdoor dock. However, the storage 

area may have been used to store drums containing mixed fission products. Radioactive 

materials included individual gamma-graphic sources, radioactive laundry, and shipping casks. 

While reportedly “no radioactive containers were opened in the area of Building 4641,” and “the 

building had a radiation detector alarm system in the dock, and the alarm system was never 

triggered,” these assertions are inconsistent with building function and known operations. In 

addition, the failure of an alarm to be triggered may not be a reliable indication of safe radiation 

levels in the area. To the contrary, given the building’s function, it could suggest a faulty alarm 

system or poor calibration of detection equipment. 

17.5.4.2 Building 4641 Radiological Incident Reports 

There has been one documented incident associated with Building 4641 that could have 

resulted in a release to the environment and worker exposure. The following table provides 

information presented in an incidents database provided by Boeing. 

17.5.4.3 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0581 6/20/1989 Bldg. 4641 Traffic 

SSFL

Be-7 Radioactive material arrived at SSFL 

Traffic without any labels or control.

• On June 20, 1989 13 mCi of solid Be-7 arrived at the facility in a box without any labels or 

controls, in violation of State of California regulations. The incident was summarized as “a loss 

of control and security required of radioactive materials.” There were no known releases to the 

environment or worker exposure documented, as a result of this incident. 

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Section 17.5.5 

Building Number:

Building Alias: 

4066 

066/T066 

Building Name/Function: Instrumentation Repair / Calibration 

Notes: Designated “Non-Radiological” Facility

Documented Radiological Processes & Incidents

Rocketdyne Employees

Workers Not Monitored for Radiation Exposure 

Radionuclides of Concern: Uranium, due to documented incidents / processes. In addition, 

this building’s proximity to Building 4059 raises concerns about the following radionuclides: U238, U-234, U-235, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Cs-137, Sr-90, H-3, Fe-55, Co58, Co-60, Ni-63, Ba-133, Eu-152, Eu-154, Eu-155, Pm-147, Ta-182 

17.5.5.1 Description of Operations & Processes 

Building 4066 area comprised Building 4066, Time Clock Building 4806, and the land 

surrounding these two buildings. Building 4066 was constructed in 1963 for use in calibrating 

and testing non-radiological equipment. It was a 4,800-square-foot steel building with concrete 

foundations and floor, corrugated steel siding and roof, housing a 3,524-square-foot laboratory.1 

Building 4066 contained internal partition walls with wood framing and drywall. One room had 

copper shielding in the walls to provide radio-frequency shielding for instruments. A steel teststand was located outside the south side of Building 4066. 

There were no Radiological Use Authorizations issued for Building 4066. The preliminary 

MARSSIM Classification for the Building 4066 area is Class 1 due to its location within ETEC, 

close proximity to SNAP Reactor Building 4059, and because no site investigation has been 

conducted. 

Surface water in the vicinity of the former Building 4066 likely flows east into a storm drain on 

SSFL’s 20th Street, that directs the water southeast via an intermittent stream through Area III to 

a pond in Area II. The storm drain collects storm water from the northern and eastern portion of 

EPA’s Subarea HSA-5C. Storm water originating from the former SNAP reactor Building 4059 

discharges into this storm drain. Consequently, this storm drain may provide a pathway for the 

migration of radionuclides originating from the former research reactor. In addition, the sanity 

sewer lines located along the northern and eastern side of Building 4057 presented some 

concern to EPA because, if radioactive materials were released into the sanitary sewer system, 

residual contamination may exist in the materials surrounding the sewer lines. 

1 EPA/HGL, Final HSA 5C, p. 41, citations 1&2:

1 Pendleberry, S. L., Removal of DOE Buildings, Demo Pak A, Boeing Report EID-04366, May 18, 1999, pp.

4-6.

2 Santa Susana Area IV, Atomics International/Energy Systems Group Planning Maps, March 1962–

November 1992. Ref. 122119552 

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17.5.5.2 Building 4066 Radiological Incident Reports 

There has been one documented incident associated with Building 4066 that could have 

resulted in a release to the environment and worker exposure. The following table provides 

information presented in an incidents database provided by Boeing. In addition, there has been 

one historical document located that describes ongoing radiological processes at this location. 

17.5.5.3 Radiological Incident Report Summary - Data Provided by Boeing 

Incident File 

Name

Date of 

Incident

Location of 

Incident

Isotopes Description of Incident

A0599 10/31/1966 Instrument Shop U Contaminated furnace switch taken to 

clean instrument shop for repair.

• In October 1966, an in-line vacuum switch was removed from the tiltpour pumping system and 

hand-carried to Building 4066. When the instrument technician opened the switch to calibrate 

it, a fine black powder, assumed to be uranium oxide, leaked out onto the technician’s clothing 

and work bench. The technician’s clothing was removed; he showered in Building 4020 (Hot 

Laboratory), and was subsequently found to be free of contamination. The area around the 

work bench was decontaminated (A0599).1 

•It should be noted that as a “non-radioactive” location, workers of Building 4066 were 

considered “non-radiological” workers and were not monitored for radiation exposure. In 

addition, the worker involved in this contamination incident was sent to Building 4020 (Hot 

Laboratory) for decontamination procedures. However, Building 4020 was considered a 

restricted area, and was not supposed to be accessible to workers that were not monitored for 

radiation exposure.

• Atomics International’s Safety Progress Report acknowledged Building 4066’s routine use of 

radiological materials by specifying that the Radiation Safety Department would need to be 

actively involved before employees were allowed to work on instruments or equipment. In 

addition, the report specifies ongoing work and repair of the radioactive exhaust system. Its 

need for repair suggests that it had been formerly used, and it is therefore reasonable to 

assume that it was radioactively contaminated as a result.2

1 EPA/HGL, Final HSA 5C, p. 41, citation 3:

3 Anonymous, Atomics International Monthly Progress Report to M. E. Remley, re: In-Line Vacuum Switch, 

November 8, 1966. 

2 Atomics International Monthly Progress Report, October 1966. R.E. Remley. File: HDMSP00051133.pdf 

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18.0 SANTA SUSANA FIELD LABORATORY’S ASSOCIATED FACILITIES: 

Downey • Canoga • VanOwen • DeSoto 

18.1 Downey Facility 

The 2006 Site Profile provides insufficient and incomplete information about operations at the 

Downey Facility. The current SEC is in effect until 1955. According to Boeing, AEC operations 

and the Water Boiler Neutron Source (WBNS) reactor existed and were operational at Downey 

until 1956, after which time suitable site remediation activities likely would have involved 

employees of North American Aviation, the DOE Contractor.1 It would be reasonable to consider 

these employees eligible to EEOICPA. 

Downey Facility is located on Lakewood Boulevard in Downey, California. The 2006 Site 

Description states that AEC activities were limited to only a “small portion of a large building.” 

However, that assertion is contradicted throughout Boeing’s 2001 radiological survey of the 

Downey Facility, which documents radioactive materials storage, handling, use, and nuclear 

research in several uncontrolled areas that were accessible to all employees. There is no 

indication that these operations were carried out in a “secluded” area, or that they were 

conducted in a location that was deemed suitable for the safe handling of radioactive materials 

and nuclear research. 

According to Boeing, the WBNS reactor was developed and operated on the only loading dock 

of the Downey Facility. As the only loading dock, it was considered the shipping and receiving 

point of entry and exit for all operations. In addition to the operation of the WBNS on the loading 

dock, the “exponential pile” of source material was stored on the loading dock. Given the normal 

function of a “loading dock,” is likely that this area was frequented by employees that were not 

appropriately monitored for radiation exposure. 

Downey Facility also hosted a 2 MeV Van de Graaf Particle Accelerator, a radiochemical 

laboratory, a neutron counting room, and a “construction area with a small .05W teaching 

reactor.” There is no indication that a “construction area” was suitably equipped to contain or 

house any type of experimental or teaching reactor, or associated radioactive materials. There 

is no indication that the “construction area” used for the teaching reactor was located in a 

secluded, restricted, or otherwise controlled area. 

The quality of site remediation following the removal of the WBNS and transfer of AEC 

operations to Canoga Park and SSFL remains undetermined. According to Boeing’s radiological 

survey, records of the facility’s early activities are “sketchy, anecdotal, and non-existent.” 

Boeing references the storage of the exponential pile on the loading dock, and additional 

radioactive materials used in the “Pearlman Laboratory” and the basement of the “Finger 

Building.” In addition, the basement was identified as the location of the film dosimetry 

processing and calibration source storage. Pearlman’s Laboratory was identified as the location 

of a radiochemistry lab. These locations are not indicative of seclusion, nor a “small area in a 

1 Boeing, Engineering Product Document #RS-00019, Radiation Survey of Downey Facility prepared by P. Liddy & P. 

Rutherford, May 4, 2001. 

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large building.” Photographs of the WBNS on the loading dock show workers standing around 

the reactor, holding survey equipment. The reactor is surrounding by several wooden shipping 

containers (indicative of typical loading dock operations). One worker can be seen atop several 

stacked wooden shipping containers, using survey equipment near the top of the reactor. The 

final designation of the wooden shipping containers is unknown.1 

Given the involvement of NAA nuclear space propulsion systems that involved both DOE and 

NASA, the development of the Space Shuttle and International Space Station in DOE locations 

at SSFL Area IV, DOE involvement and proprietary interests in Downey operations after 

1955-56 should be evaluated. As more information is obtained about agency and contractor 

projects, there may be justification to add years to the covered time period for Downey workers 

(potentially under the SEC class). 

18.2 Canoga Park / Van Owen Facility 

The 2006 Site Profile provides insufficient and incomplete information about operations at the 

Canoga / Van Owen Facility, which is not covered by EEOICPA after 1960. CORE Advocacy is 

currently evaluating historical documents related to Canoga Facility. There are some indications 

that Canoga and Van Owen Facility may have supported DOE operations at SSFL after 1960. A

newly obtained incident report documenting a 1967 uranium fire at the Canoga Facility is 

provided below. 

The Special Nuclear Materials License (No. SNM-21) authorized Atomics International Division 

of North American Aviation to receive and possess 50 grams of uranium enriched in 

uranium-235 (U-235) for use in fission counter tubes. SNM-21 was amended eight times to 

increase the number and type of nuclear materials that could be handled at both the Canoga 

Park and SSFL sites, and it was in effect until September 27, 1996.2 3 In addition, in 2005 

DEEOIC acknowledged the original North American Aviation - AEC contract, wherein North 

American Aviation was permitted its discretionary use of its facilities and locations owned by the 

company or leased by AEC to perform functions affiliated with fulfilling its contractual 

obligations.4 It would be reasonable that Canoga and VanOwen facilities would be used, as 

required, to support SSFL or DeSoto DOE operations. 

The 2006 Site Description indicates that AEC-sponsored activities at Canoga Facility occurred 

in the VanOwen Building from 1954-1960 following the move from Downey Facility in 1955. In 

addition, the 2006 Site Description indicates the principal work in the VanOwen Building 

included design, development and operation of small aqueous fuel reactors, fuel development, 

radiochemistry, and beryllium machining. The reactor activities at Canoga Facility involved the 

1 Boeing, Engineering Product Document #RS-00019, Radiation Survey of Downey Facility prepared by P. Liddy & P. 

Rutherford, May 4, 2001. Page 11. 

2 U.S. Nuclear Regulatory Commission, Materials License, SNM-21, September 15, 1977. 

3 U.S. Nuclear Regulatory Commission, NUREG-1077, Environmental Impact Appraisal for Renewal of Special 

Nuclear Material License No. SNM-21, June, 1984. 

4 Division of Energy Employees Occupational Illness Compensation (DEEOIC) Peter M. Turcic Memorandum to 

Christy Long, District Director of Seattle, Subject: Atomics International and Energy Technology Engineering Center, 

September 7, 2005 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 

 

Page 404 of 418 

construction of small 10-W aqueous uranyl sulfate homogenous reactors (L-47 / L-77) for use in 

training institutions. According to the 2006 Site Description, reactor fueling only occurred after 

the reactors were installed at the training institutions (contradicting the previous statement that 

the reactors operated at Canoga Facility). Reportedly, only a few fuel elements for the organic 

moderated reactor (OMR) were fabricated at the Canoga Facility. 

The 2006 Site Description indicates that AEC-processes at VanOwen occurred in the machine 

shop (including beryllium machining), a radiochemical laboratory, and office space. In addition, 

the 2006 Site Description indicates wastewater from all facilities except the laboratory was 

discharged into a common treatment plant; wastewater from the radiochemical laboratory went 

to a clarifier where it was tested to ensure safe release to the sanity sewer); and sludge from the 

clarifier was disposed off-site at a licensed low-level facility. 

18.2.1 L-47 Reactor 

In 1996, ORAU reviewed 59 terminated research and test reactor license docket files and 

concluded that License R-19 provided insufficient decommissioning and disposition information 

about the L-47 reactor licensed to Atomics International for operations at Canoga Facility 

between 1957-58.1 2 Upon ORAU’s visit to the Canoga Facility, the L-47 reactor could not be 

located. No information on its disposition, or that of its fuel, could be located.3 

According to the NRC, in December 1956 and January 1957 Atomics International submitted 

three applications to the AEC to construct, manufacture, possess, and use a 5-watt nuclear 

research reactor; the Model L-47 intended for demonstration, experimentation and sales 

purposes. Construction Permit No. CPRR-14 was issued August 2, 1957 (the license for the 

reactor erroneously stated that the Permit was issued July 2, 1957), to North American Aviation 

for construction of the 5-watt (thermal) utilization facility; a “homogenous solution-type reactor 

utilizing highly enriched uranyl sulfate as fuel in distilled light water.” 

Records indicate the recommended amount of fuel to be purchased for the L-47 was 2,000 

grams of fully-enriched uranium-235 (U-235) in the form of uranyl sulfate. The amount of fuel 

actually purchased was not clearly identified. The L-47 was installed in the Reactor Room at 

VanOwen and began operating August 29, 1957. It operated for 73 watt-hours in 1957. 

Operations data was not located for 1958. License R-19 was terminated by the AEC June 30, 

1958. 

The L-47 was considered by some to be a “poor man’s reactor;” the original name of the reactor 

was the “Penny Ante.” It was quickly replaced by the more powerful 10-watt Model L-77.4 

1 AEC License R-19 authorizing operation of the L-47. August, 1957. File: Vanowen_AEC_1957-08-05.pdf 

2 AEC Docket No. 50-50 License R-19 Termination Request. June 30, 1958. File: Vanowen_AEC_1958-06-30.pdf 

3 Nuclear Regulatory Commission memorandum to Michael F. Weber, Chief, Low-Level Waste and Decommissioning 

Projects Branch Division of Waste Management re: Rockwell International Site Visit. January 29, 1996.

File: Vanowen_NRC_1996-01-29.pdf 

4 AEC License R-40 authorizing operation of the L-77. May, 1958. File: Vanowen_AEC1958-05-17.pdf 

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18.2.2 L-77 Reactor 

The L-77 Reactor license to operate at Canoga Facility was amended to allow for operations at 

DeSoto Facility, issued June 28, 1960. 1 To date, CORE Advocacy has not located any 

information on L-77 operations at the Canoga / VanOwen Facility prior to its transition to the 

DeSoto Facility. 

18.2.3 Canoga Facility - Incidents After 1960 

• A May 19, 1967 Atomics International internal memo references commendations for several 

employees involved in a uranium fire at the Canoga Facility on May 17, 1967.2 

1 AEC License Amendment, L-77 Operations to DeSoto Facility. June, 1960. File: Vanowen_AEC_1960-06-28.pdf 

2 Atomics International Internal Memo, “Commendation Re: Uranium Fire, Canoga Facility, May 17, ,1967” May 19,

1967. File: 810_Uranium_Fire_Canoga.jpg.pdf 

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18.3 DeSoto Facility 

The 2006 Site Profile contains insufficient and incomplete information about operations, 

processes, and incidents at the DeSoto Facility that is likely to compromise the dose 

reconstruction process. It should be noted that currently, the SSFL Site Profile lacks historical 

sources associated with DeSoto operations. 

Currently, CORE Advocacy’s review of DeSoto Facility operations is ongoing. Some newly 

obtained information is provided, below. CORE Advocacy anticipates a submission to NIOSHABRWH pertaining to DeSoto Facility, exclusively. It should be noted that several reports 

contained in the Incident Database (some of which are referenced below) document insufficient 

sampling and worker monitoring practices well after the expiration of the 1964 SEC. This may 

support a need to expand the current SEC into subsequent years. 

The 2006 Site Description indicates that radiological operations occurred at the DeSoto Facility 

between 1959 to the mid-1980’s; Atomics International used nuclear fuel material and other 

radioactive materials in Buildings 101 and 104 between 1959-1983. In addition, the 2006 Site 

Description states a reduced level of work was continued by Rocketdyne (Building 104) into the 

mid-1990’s. 

However, given SSFL’s site remediation activities and NASA operations in Area IV during this 

time period, for which DeSoto played a supportive role (and vice versa), newly obtained 

information should be carefully reviewed to verify the accuracy of the 1983 end-date for nuclear 

operations, and the reduction of Rocketdyne work into the 1990’s at DeSoto. 

Before 1984, Buildings 101 and 104 were designated as 001 and 004. Much of the historical 

documentation available refers to these building numbers. 

18.3.1 L-77 Nuclear Reactor (NRC Licensed) 

The 2006 Site Profile describes the L-77 as a small, low-power (10 W) research reactor that 

was operated in Room 416-61 of Building 104/004 from 1960-1976; it used enriched uranyl 

sulfate solution. The L-77 is described as a prototype teaching reactor sold to many universities 

and around the world. Reportedly, the laboratory housing the L-77 was decommissioned and 

decontaminated in the late 1970’s, and the NRC released the facility for unrestricted use and 

terminated the reactor’s license in February of 1982. The 2006 Site Description provides no 

sourced historical documentation of facility operations, or incident reports associated with the 

L-77. 

18.3.2 Advanced Test Reactor (ATR) Fuel Fabrication and Supportive Activities 

The 2006 Site Profile indicates that the NRC-licensed fuel fabrication operations were 

conducted in the northern section of the first floor of Building 101, with radiochemistry support 

operations conducted on the first and second floors of Building 104. These support activities 

included hot chemistry laboratories (east section of second floor), an emission spectroscopy 

laboratory (Room 411-72, first floor), and an X-ray diffraction laboratory (Room 411-58, first 

floor). The fuel fabrication facility reportedly produced a variety of different fuel elements for test 

reactors. Many fuel manufacturing programs began in 1959 using 2% to 93% enriched uranium 

metal and composites. Some of the work involved developing uranium-aluminum allows and, 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
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because of the uranium, occurred in sealed glove boxes. One of the larger programs was fuel 

manufacture for the ATR using uranium-aluminum (UAIx) powder with an enrichment of 93% (an 

incident involving UAIx is described in the Log Book entry summary, below). Fuel manufacturing 

ended in 1983, according to the 2006 Site Profile. In addition, the AEC-sponsored work 

involving the manufacture of beryllium-containing parts took place at DeSoto Facility. 

Reportedly, D&D of Buildings 101 and 104 included removal of all fuel and radioactive materials 

and waste; removal of contaminated equipment, drain lines, tanks and ventilation ducts, and 

cleaning of all surfaces including floors, walls, and ceilings. The preceding two paragraphs is the 

only information about ATR operations and fuel fabrication / supportive activities provided in the 

2006 Site Profile. The 2006 Site Description provides no sourced historical documentation of 

facility operations, or incident reports associated with the ATR. 

18.3.3 Gamma Irradiation Facility (GIF) 

The 2006 Site Description indicates that the GIF was a state-licensed above-ground vault in 

Building 104 that used sealed Cs-137 and Co-60 sources for radiation hardening tests of 

electronic components and for food irradiation research. The GIF consisted of Rooms 41M-11 

and 41-M11A on the northeast corner of Building 104. Activity is reported to have ceased in the 

late 1980’s and the sources were apparently shipped off the site for recycling in the early 

1990’s. Bi-annual leak checks of these sources detected no leaks. In 1995, a Rocketdyne 

survey of the GIF reportedly verified that it was not contaminated. However, the 2006 Site 

Description provides no sourced historical documentation of facility operations, or incident 

reports associated with this operation. 

18.3.4 Mass Spectroscopy Laboratory (Helium Laboratory) 

The 2006 Site Description indicates that up until 1995, Rocketdyne used the state-licensed 

Mass Spectroscopy Laboratory to analyze miniature radioactive specimens of neutron-irradiated 

nonfissile metals from DOE and international reactors for helium content. The Mass 

Spectroscopy Laboratory consisted of Rooms 414-69, 416-72, 414-75, 416-76, 416-76A, 

414-77, 418-80, 416-80A, and 414-81, in the northeast section of the 1st floor of Building 104. 

The 2006 Site Description does not provide a detailed description of operations or processes 

associated with these locations, or any incident reports associated with the operations of the 

Mass Spectroscopy Laboratory / Helium Laboratory. 

According to the 2006 Site Description, operations at this location generated very small 

quantities of radioactive, hazardous and mixed waste, including radioactive solvents, solvent 

wipes, and acids shipped from other U.S. and international research organizations. In 1995, the 

equipment was shipped to Pacific Northwest National Laboratory in Richland, Washington, 

where it continues to be used. 

The 2006 Site Description indicates that fuel fabrication was terminated in 1984, but small-scale 

laboratory research work on gamma irradiation and analysis of radioactive samples continued 

until 1995. Remedial activities occurred at various times in the 1980’s followed by NRC license 

termination. In the mid-1998, all remaining equipment, interior walls, and drain lines were 

removed. The facility was reportedly decontaminated; residual contamination was reported to be 

“low-level and confined to the laboratory. No elevated radiation levels or contamination outside 

the laboratory were detected before or during D&D.” 

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The 2006 Site Description indicates that no contaminated water held in tanks was ever above 

the MPC, but this assertion is contradicted in the log book, which documents water sampling 

practices and radiation readings, prior to release into the municipal sewer system.1 

18.3.5 DeSoto Facility, Select Log Book Entries - Various Locations 

A DeSoto Facility Log Book (October 28, 1964-November 19, 1965) chronicles several incidents 

that could have resulted in a release to the environment and worker exposure. In addition, 

sampling of the contaminated water holdup tanks and releases to the municipal sewer system 

are documented. Below is a summary table of select Log Book entries from the first 50 pages of 

the document (which exceeds 300 pages). The document is provided in its entirety on the disk 

accompanying the 2016 SSFL Site Description. A partial list of additional incidents involving the 

DeSoto Facility is provided, at the conclusion of the table summary below. 

18.3.6 DeSoto Log Book Entry Summary (Select Entries - First 50 Pages) 

Date Location Isotopes Entry Summary

10/29/1964 119-336 unknown [REDACTED] reported to H.P. with some grey dust on his hands which had 

fallen out of elephant trunk in Room 119-336. Smears of the dust indicated 

>1600 dpm/10c

most of the dust off with alcohol and kimwipes and then washed hands in the 

change room using soap and scrub brush. 2nd spears indicated no 

detectable contamination alpha, beta or gamma. Meter check also indicated 

no detectable radiation. He was released.

10/29/1964 11D-08D unknown Smears of the vac pump in 11D-08 indicated internal contamination ~600 

dpm/100cm2

beta-gamma. The log book entry indicates no alpha contamination was 

detected. 

10/29/1964 11D-08D Cs-137 Smears on floor 11D-08D indicated general (beta-gamma only) 

contamination from 150 dpm / 100 cm

made temporary shoe-covered area until maintenance can scrub and wax.

10/29/1964 11D-08D MFP Smears on floor 11D-08D indicated general (beta-gamma only) 

contamination from 150 dpm / 100 cm

made temporary shoe-covered area until maintenance can scrub and wax.

10/30/1964 ... smears taken in the room indicated a maximum of 70 dpm alpha and 100 

dpm beta gamma. Maintenance personnel were therefore not required to 

wear respirators. Smears of old motor indicated 102 dpm alpha and 210 dpm 

beta-gamma. The motor will be decontaminated in the filter bank room. 

Maintenance was unable to decontaminate the motor. It will be left in the 

room and decontaminated sometimes during the week. 

11/2/1964 Smeared & monitored piece of SNAP-10 fuel rod (plastic bag container) ok, 

tagged it @ 150 mr/hr at surface.

1 De Soto Facility, Atomics International Log-Book, Building 104. Document retrieved from publicly accessible 

historical archive, courtesy of the California Department of Toxic Substances Control (DTSC). File: 

HDMSp01640471.pdf 

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11/3/1964 Checked 3 containers of Hallam elements for shipment to SS [Santa Susana] 

Materials Vault at SanSu [Santa Susana] (4064). Total activity = 2770 mr 

maximum surface intensity = 2.2 mR/h. At one meter from container surface 

= 0.6 mR/h.

Checked cask being trans-shipped to San Su vault, contained less than 1 

curie of irradiated stainless steel. 65 mr/h at cask surface & 5mr/h at one 

meter.

11/4/1964 BeO Checked irradiated BeO to be machined in [REDACTED]’s lab. 25 mr/h @ 

surface, < .05 mr/hr at 1 meter.

11/4/1964 416-41 EU Spill in Rm 416-41. Etching acid with enriched uranium spilled in hood and 

onto floor. No smears taken before because of obvious contamination. Floor 

scrubbed by D/766. Smears taken after scrub revealed no detectable 

contamination. Sampled and released FED II to city sewer system.

11/5/1964 119-41

(1114)

Received a call from Rm 119-41 that smoke was coming from the [illegible] 

ventilation ducts. HP called control center and responded with repirator. It 

was found that ventilation from F.B. 40 was off. HP requested that rooms 

119-41 & 45 be evacuated. In checking the roof it was found that the main 

switch was on but the filter bank off. Maintenance was called. [Redacted] 

checked the emergency off switch in Room 119-42 and found someone had 

tripped it. F.B. to back in service at 09:30. Suggest a protective cover be 

placed over the switch. Control canter called out all clear.

11/5/1964 Bldg. 004 UC Shut down bank at 21:45. Started back up 22:35. Finished filter change on 

filter bank #5 Building #004. Smears of the floor of the absolute side read 12 

counts over background beta-gamma. Smears taken of the pre-filter floor 

read background for beta-gamma. Neither side showed any alpha count. The 

dirty filters were reading 0.70 mr at3 inches inside the filter bank and 

somewhat less outside the bank. The men were in the bank about an hour 

and half. While two men were helping from outside - which helped to speed 

up the change. The used filters were all tagged and wrapped in sheets of 

poly for protection against contamination. The start-up of the change was 

slowed down because of a shortage of useable filters. The men had to make 

several trips to the Nordhoff warehouse to pickup additional filters.

11/6/64 0900-1045. Control center put a call in for RA [illegible] on the traffic code, 

HP responded. It was found a R.A. liquid waste drum was leaking through 

the [illegible] hole. The spillage was controlled and liquid transferred to a new 

drum. Smears of the area were [illegible]. The liquid was from maintenance 

liquid waste. Firemen washed the area down with water. All clear at 10:45.

11/10/1984 

[sic?]

U

MFP

Checked recorder for [Redacted]. No contamination detected. Smears of 

MFP operation indicated contamination (22,431 dpm beta-gamma, 616 dpm 

alpha) limited to small ram block. Smear after draining indicated no 

detectable contamination. Checked can of normal Uranium for transfer to 

Bingham. No contamination. 4 mr @ surface. ~0.6 [illegible] sampled Fed I to 

[Redacted]. Smeared glove box 1242 (now 11H-42a) for SNAPTRAN; 

external background alpha, beta, gamma, internal indicated 150 dpm alpha / 

1600 dpm beta-gamma.

Date Location Isotopes Entry Summary

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11/11/64 11H-37 Obtained several smears in room 11H-37 in preparation for outside 

contractor work. Only the panel around furnace is significantly contaminated. 

When guard screens are removed, piping thereunder may also be 

contaminated, but no smears could be taken at this time. [Redacted] has 

been informed of results.

11/12/1964 1030 U

NaK

Fire - Room 1030. Apparently butyl alcohol in induction furnace system 

overheated, blew off tygon line and released = 7,000 of NAE into steel catch 

pan; 25 grams of normal uranium in furnace at time; no problem with it. Fire 

extinguished by fire department. Much smoke - plastic induction enclosure 

was damaged. Maintenance cleaned up after project personnel removed 

NaK butyl solution to safe place. No injuries.

11/12/64 115-13 Thorium Smears of Room 115-13 indicate contamination to 300 dpm / 100 cm2 alpha. 

Contamination apparently resulted when thorium oxide was being sifted. The 

sieves did not fit tightly and airborne contamination caused the room to 

become contaminated. Lab was decontaminated by project personnel and 

subsequent smears revealed contamination reduced to acceptable limits.

11/14/64 EU Monitored shipment to Montebello SPERT III rods 4-8% Enriched U. 2-3 mc .

616 [illegible] 5.0 mrem/h at surface of rods; .2 mr/h at external shipping 

container. Changed filter paper in constant air monitor in hallway of 

radiochemistry Building #4 because no air flow could be detected and unit 

seemed to be overheating. F/M reported louder than normal noise. After 

sample paper change, noise subsided. Filter Bank # 52 down at 12:10 hrs to 

change bearings. The timer to the #2 alpha [illegible] is not working. Bank 52 

will not be back in service until some time Monday. Signs have been posted 

at the effected rooms. At the start of the bearing change I was told that only 

the bearing and the [illegible] will be worked on. At the end of the shift when I 

had not been notified of the bank being turned back on, I checked the bank. 

In addition to the motor and bearings they also had been working on the fan. 

Smears taken of the fan indicated 8 dpm alpha and 42 dpm beta-gamma 

indicating that any tools used could be contaminated and that some 

contamination has gotten past the absolute filters. Since no one was at 

maintenance, I didn’t have any way to check their tools.

11/16/64 11H-37B

11H-22

Spill in SNAP Area 11H-37B not contaminated ran into brown room d/789. 

Mopes and buckets were contaminated (15 dpm alpha, no beta-gamma). 

Maintenance mopped up 11H-37B, will mop brown room. ... Spill of 

radioactive waste in 11H-22 and aisle, 11H indicated 90 dpm beta-gamma 80 

dpm alpha. Localized; maintenance will mop up.

11/16/1964 4.91% 

UC

Spill involving 100 grams of 4.91% vitro UC fuel stock in room 115-09 (1003). 

See report for details. [Report was not located].

Date Location Isotopes Entry Summary

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11/17/1964 11D-08D H&S provided a R/A smear & meter DEP survey in Lab #11D08D. The survey 

was performed during and after the welding of R/A capsules. A meter survey 

revealed 110 mR/h beta/gamma at approx. 6” from the capsules. A smear 

survey revealed the following:

Welding Tip - 16,240 dpm/100cm

Clamp - 900 dpm/100 cm

Welders Glove - 26,200 dpm/100 cm

Welding gun 5,000 dpm / 100 cm

Floor Area <30 dpm / 100cm

Work Bench 200 dpm / 100 cm

A smear survey of a glove box in Lab 11D-08D revealed the following:

Outside Areas - 150 dpm/100cm

the inside due to the hi-level contamination.

11/18/1964 U Monitored and prepared 8 ea. birdcages for transfer to Santa Su. All 

birdcages speared <1 d/m 100/cm2 alpha and <30 d/m /100 cm2 betagamma. The maximum cm [sic] reading at the surface revealed .4mR/h 

gamma. The birdcages are definitely contaminated on inside. As per 

[Redacted] H&S was unable to check the contamination level insisde due to 

the cans being previously closed.

11/18/1964 11D-08D A smear survey was conducted in the northern corner of Lab 11D-08D after 

removal of the glove box from the corner. The survey revealed a maximum of 

200 dpm/100cm

was notified.

11/18/1964 H&S advised and assisted in preparing the service grinder in Lab 11H & 428 

for grinding a 6 mR/h sample. Sample will be ground on 11-19-64.

11/18/1964 Aisle 115 1700: Notified Captain [Redacted] of extreme dust raised by sand blasting 

Aisle 15. Eye & throat irritation apparent. Visibility under worst condition ~20 

ft. Conversation between [Redacted - four employees] resulted in continued 

operation until 1900. Blasting is supposed to resume this weekend.

11/19/1964 11H-37B Enriched 

U

H&S provided coverage in Lab 11H-37B during the period of time a piece of 

93% Enriched “U” was being cut in half. Air samples were taken and revealed 

434x10-11

a maximum of 6 mR/hr gamma at 2” and 100 dpm/100cm

100cm2 beta-gamma. A smear survey was taken of the area around the 

service grinder and revealed a maximum of 12 dpm/100cm

100 cm2 beta-gamma.

11/19/1964 11H-428 A smear survey was conducted by H&S of misc. tools in Lab 11H-428 and 

revealed a maximum of 240 dpm/100 cm

gamma. Advised operations to clean immediately.

11/29/1964 Routine radiation survey of areas designated “clean” where there is possibly 

some non-belonging radiation revealed 0.1 mR/h above bkg. 

11/30/1964 11H-37 Smears taken of pit in 11H-37 and of plate covering pit indicated 400 dpm 

alpha; 500 dpm beta-gamma on top of side of plate (now facing down); 

bottom side of plate & walls of pit indicated contamination within limits for 

yellow tag area. Water sample from pit indicated 0 dpm alpha, beta, gamma/

cc of water.

Date Location Isotopes Entry Summary

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12/3/1964 119-33 Cs-137 Smears of Cs-137 hood in 119-33 indicated no contamination on floor in front 

of hood, but ~3,000 dpm beta-gamma on all inside surfaces (walls, floor, but 

not ceiling) of hood and ~3,000 dpm geta gamma on [ineligible] exterior.

12/3/1964 119-33 Cs-137 Smears of Cs-137 hood in 119-33 following alconox scrubbing reveal 100 to 

200 dpm on floor of hood and outer lip. No other significant contamination on 

other surfaces.

12/5/1964 1059 At the request of [Redacted], next tenant of 1059, smears were taken of the 

walls. Results are as follows, left side, middle, right side:

North Wall 12.3-102 / 16-68 / 8-48 

East Wall [no reading indicated] / 36-102 / [no reading indicated]

South Wall 82-224 / *3,345-11,934 / 4

*probably from cutting of R/A ducting

12/8/1964 Filter 

Bank #53

Room 

1126

Filter Bank #53 down to change filters. What appeared to be coolant water 

was found in the bottom of the bank. Upon removal of the pre-filters, water 

marks could be seen from the exhaust trunk leading from room 1126. A visual 

inspection, smear, and meter survey of the exhaust outlet revealed that 

contaminated water had been confined to the wooden surface directly below 

the exhaust outlet. Smears taken indicated as follows: [Ranges provided for 

purposes of this document]:

0 to 2,000 dpm alpha / 0 to 2,400 dpm beta-gamma

A visual inspection of the ducting in Room 1126 to determine the cause of the 

high smear results revealed what appeared to be water marks, indicating a 

leak in the ducting. This probably explains the reason for the north side and 

top of the bank being contaminated. It probably should be noted that the fan 

and ducting on the clean side of the absolute filters are now highly internally 

contaminated. Due to lack of true decontamination of the outside of the bank, 

could not be completed. To preclude a similar recurrence, a sign has been 

placed on the bank room switch prohibiting the use of the bank until the 

cause of the occurrence can be determined.

12/8/1964 1131 

(119-45)

Vent filters in Room 1131 (119-45) were changed. FED II back in service @ 

2300 hours. Smears taken indicated general contamination of aisle 119 to 

room 1131. Contamination was highest at the filter bank doors (250 dpm 

alpha) and progressively decreased to background at Room 1131 and at the 

north double doors. Smears taken inside the filter bank room indicated 

contamination of 2,000 dpm alpha on the floor and 1,000 dpm alpha on the 

bank itself. Due to the lack of sufficient time, neither a complete 

decontamination of the filter bank room or a complete smear survey of same 

could be performed. At the end of the 3rd shift, three decontaminations had 

succeeded in reducing the contamination levels to 40 dpm alpha at the bank 

doors, 10 dpm at Room 1139, and the guard post, and background at 1131 

and the north end of the double doors. Maintenance is holding two men over 

to finish the decontamination of the area. The release of the area will have to 

be initiated on the first shift.

12/11/1964 Shoes of guard on 3rd shift working detox in Bldg HQ 001 SNAP & FED II 

area were found to be contaminated to 20 dpm alpha. Cleanings reduced this 

to < 1.0 dpm. He walked patrol without shoe covers.

[Note: Guard involvement in detox procedures].

Date Location Isotopes Entry Summary

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12/19/1964 Room 

426-07

At 1515 hours [Redacted] of D/734 entered room 426-07 and after being in 

the room approximately 3 minutes began to feel groggy. He recognized that 

the symptom was probably due to insufficient oxygen content in the room air, 

and at the same time heard hissing from a needle valve used to purge on of 

LASERS with nitrogen. He shut off the valve, exited from the room, and 

called the control center. After a telephone call to [Redacted], it was decided 

that [Redacted] should be taken to the West Valley Park Hospital as a 

precautionary examination. [Redacted] was released from the hospital at 

1800 hours on 12/19/64. [Redacted] when questioned as to the probable 

cause of the valve being left opened states that [Redacted] had probably 

used the LASER last and could have left the valve open. He also stated that 

if this were the case, the valve had been open and filling the room with 

nitrogen for 20 hours.

12/21/1964 119-41 Routine check of respirators in 119-41 indicated general contamination in 

and out on 3 of 4 checked, of about 350 dpm alpha, beta, gamma. The fourth 

mask marked “*” indicated 600 dpm alpha, and 900 dpm beta-gamma on 

exterior; interior indicated same as others.

12/21/1964 11D-27D [Redacted] D/766 got dust in his right eye from ductwork in 11D-27D. First 

check indicated 856 dpm alpha, no beta/gamma. After 1, 2 and 3 washings 

all indicated 40 dpm alpha. After 4th washing, 16 dpm alpha. After 5th 

washing indicated 4 dpm alpha.

1/5/1965 Loading 

Dock

Ir-192 Source - approx. 10 curies arrived dock & delivered to X-Ray w/out 

identification on the outside of the container. A letter will be written from 

Santa Susana. Informed [Redacted] of situation in 1159 i.e. 57 smears 

indicated general floor contamination to ~500 dpm beta gamma / 400 dpm 

alpha. Spot contamination to 4,500 dpm alpha; 7,000 dpm beta-gamma

1/14/65 Smears taken of the cars in the HQ-1 parking lot indicated from 0 to 127 dpm 

beta-gamma (no alpha was detected). Since [redacted]’s smears indicated 

just about the same levels, the panic button was not pushed at SS. 

[Redacted] suggested also that an air sampling be taken since [redacted] 

would probably want such results. A few minute sampling was taken but the 

results could not be calculated since the hurricane [illegible] in excess of the 

[illegible] of [illegible] at which it is calibrated.

2/2/1965 Smear from trousers of [redacted] revealed alpha contamination to 300 dpm/

100 cm2 and 900 dpm / 100 cm2 beta-gamma. Company clothes were 

provided and an attempt is being made to decontaminate pants in question. 

Decontamination unsuccessful. Trousers were confiscated by H&S.

2/6/1965 Bldg. 001 

119-33B

U-235 Called in by [Redacted] to give coverage to maintenance on the removal and 

replacement of a scrubber system in Room 119-33B Bldg. 001. The water in 

the bottom of the scrubber was checked with the 2650 and showed a reading 

of 50 mR/hr (beta gamma) at 6”. Smears were taken of the U-235 carbon 

oxide just inside of the scrubber itself, showing a reading of 11,000 dpm / 

cm2 beta gamma. The working crew were instructed to wear gloves at all 

times along with respirators while cutting into the pipe to keep internal 

contamination to a minimum. Nasal smears and a complete survey was 

conducted of all men working in the area at the close of the working day. No 

contamination was found. Work on this will continue through tomorrow until 

completion.

Date Location Isotopes Entry Summary

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 414 of 418 

2/7/1965 Bldg. 001 

119-33B

U-235 Instructed crew to decontaminate area under the new scrubber location. This 

area was reading 11,000 dpm cm

checked out at about 530 dpm/cm

which was considered a much better working level since all of the loose 

U-235 oxide had been removed.

2/7/1965 Bldg. 001

119-33B

U-235 The working crew has been carefully instructed on the use of gloves at all 

times and a survey of the working personnel is made often to keep the 

spread of contamination to a minimum. The changing of dirty shoe covers 

often has helped to keep contamination at a minimum. The removal of 

contaminated pipe and tools coming out of this area made it necessary that 

all items were thoroughly checked out because of the ease with which U-235 

powder is picked up by anything placed on the floor. [Continued following 

page].

2/7/1965 [Continued from previous page]. For example, the used contaminated pipes 

taken from the duct work of the scrubbers had to be re-rapped [sic] in 

plyethylene sheets several times and kept from touching the floor before they 

could be removed from the area - with a clean area background reading for 

beta, gamma, as well as alpha. Cleaned up, and headed for home.

2/9/1965 Survey made of red flammable cabinet outside of reclamation area per 

[redacted] with following results: Outside 30 dpm / 100 cm

< bkgd alpha. Inside 155 dpm / 100 cm2 beta gamma and 16 dpm/100cm

alpha. This cabinet should have been decontaminated before removal from 

11D-08.

2/12/1965 RMDU 

[sic]

shed

Smears of elephant trunk ducting left outside RMDU shed indicated no 

detectable contamination on outside of bag; but ~2,700 dpm alpha and 4,383 

dpm beta-gamma on ducting inside bag. Indicated 7.0 mR/h at contact of bag 

w/ gin [sic] 2651.

2/16/1965 Arrived on the scene a bright and cheerful HP man - only to find the office 

filled with humanity. ... Lunch hour - Called by Control Center to check out an 

individual in the cafeteria dressed in Red-Line coveralls. Found said 

individual to be [Redacted] from Building #004 Room #413-53 (a clean area). 

He works for [redacted] D/734/431 Metallurgy & Mechanics. He seemed 

quite innocent of his sin, since he claimed he had only worked at HQ for five 

years and besides, he in an “Engineer” and was not cognizant of the basic 

health & safety instructions governing tagged areas. Rule #7 covers this 

infraction.

Date Location Isotopes Entry Summary

18.3.7 Partial List of Incidents / Notes of Operations Associated with DeSoto Facility 

• On January 12, 1961 Health & Safety was notified of a radioactive spill of 3% enriched 

uranium carbide in Room 423-29. The Control Center was notified immediately. Upon further 

investigation, it was learned that a [Redacted] had been weighing approximately 4 grams of 

uranium carbide which was contained in a small plastic box when it suddenly ignited. Smoke 

enveloped the immediate area and a small portion of the burning sample spilled on [Redacted] 

lab coat. The lab coat was discarded and rolled up to smother the flame. The employee was 

taken to the Health & Safety Lab for a nasal smear, which indicated significant contamination. 

Further precautionary measures were deemed necessary by the monitor. The employee was 

taken to the Health & Safety Bioassay Lab for further evaluation. Smear survey of the area 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 415 of 418 

revealed approximately 500 d/m beta-gamma and 300 d/m alpha on the floor and surrounding 

tables. (Incident Report #A0550).1

• Transfer of Neutron Source: On August 21, 1963 a request was made by the SNAP

Experimental Unit to continue the development of the neutron scatter gage in Room 426-73. 

The original trasnfer of the 7 Pu-Be neutron sources to the L-77 was approved by Criticality 

Safeguards in May, 1963. Approval was granted. 2 It was determined that the experiments 

would be performed at the L-77 (Bldg. 004), and that no more than three sources would be out 

of the shielded shipping container at any one time.3

• On December 13, 1966 an Air Sample Radioactivity Worksheet for DeSoto Facility, Room 

1110-62 indicated the following readings: Alpha: Total Counts 271 per 5 Minutes - 54 c/m 

gross - 54 c/m net - 226 d/m - 6.3 x 10 11 uc/cc. Beta-Gamma: Total Counts 498 per 5 Minutes 

- 99 c/m gross - 90 c/m net - 320 d/m - 8.0x10 11 uc/cc. UC 20% fire ~ 6.0 Kg total. ( Incident 

Report #A0309).4

• An Atomics International Safety Newsletter (1967) describes the apparent loss of airflow 

across two weighing hoods in Building 001 Room 110-62. Apparently, duct work was not 

connected to the facility exhaust system; it terminated after passing through the false ceiling. 

The condition was corrected by Plant Services.5

• An Atomics International Safety Newsletter (1967) describes smear surveys of Room 1110-62 

that revealed “excessively high” (3.5 x 104 d/m-100 cm2) contamination levels that required 

project personnel to spend a total of four shifts decontaminating the area to acceptable levels. 

According to the report, a contributing factor to these levels was the fact that the shaft seal on 

the ATR chipmunk glove box was installed backward, allowing “gross leakage from within the 

box to the room atmosphere.”

Data from lapel air samplers worn by personnel operating the ATR chipmunk during periods of 

high airborne activity ranged as high as 1.0x10 8 uCi/cc.6

1 Incident Database, provided by Boeing. 

2 Internal Letter, Atomics International. A.R. Moore / A.R. Yarrow, “Transfer of Neutron Source,” August 21, 1963.

Document retrieved from publicly accessible historical document archive, courtesy California Department of Toxic 

Substances Control (DTSC). File: HDMSP001841089.pdf 

3 Internal Letter, Atomics International, “Transfer of Neutron Source Approval,” May 24, 1963. Document retrieved

from the publicly accessible historical document archive, courtesy of the Department of Toxic Substances Control 

(DTSC). File HDMSP001841090.pdf 

4 Incident Database, provided by Boeing. 

5 Internal Letter, Atomics International. “Radiation Safety Unit Weekly Newsletter for the Period Ending July 15, 1967” 

M.E. Remley - R.E. Alexander. July 25, 1967 Document retrieved from the publicly accessible historical document 

archive, courtesy of the Department of Toxic Substances Control (DTSC). File HDMSp01641279.pdf 

6 Internal Letter, Atomics International. “Radiation Safety Unit Weekly Newsletter for the Period Ending July 15, 1967” 

M.E. Remley - R.E. Alexander. July 25, 1967 Document retrieved from the publicly accessible historical document 

archive, courtesy of California Department of Toxic Substances Control (DTSC). File HDMSp01641279.pdf 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 416 of 418 

• An Atomics International Safety Newsletter (1967) describes unusual exposure readings 

observed on an L-77 location badge utilized during the year 1966. Investigation of personnel 

exposures for the period in question indicated normal exposures for those associated with the 

L-77 operations as well as for workers in adjoining labs. The cause of the location badge 

exposures was not identified.1

• On June 18, 1970 a depleted uranium fire occurred in Room 119-42, in an oven. The fire 

occurred during an attempt to “oven dry” 20 pounds of depleted uranium chips, which had 

been generated from a milling operations in room 11-H-28 on the same day.

[Redacted] discovered the fire when he observed flames emanating from the top of the oven 

door. Upon discovering the fire, [Redacted] applied G1 powder to the two stainless steel trays 

containing the uranium. The oven temperature had been regulated for 500 ̊F but had not 

reached that temperature. Estimated temperature upon ignition was 350 ̊F. When the writer 

arrived at the facilities the fire had been extinguished by Protective Services. Floor smears 

taken throughout the room indicated maximum surface contamination levels of <50dm/100 

cm2 and <30 d/m / 100cm2 beta. Two air samples taken between 1330 and 1500 hours 

indicated maximum airborne concentrations of 1.6 x 10 10 uCi/cc on immediate count. A 38-

hour air sample collected over the balance of the weekend indicated airborne concentrations 

of 3.0 x 10 11 uCi/cc. Personnel involved were asked to submit bioassay samples on the 

following Monday. Contamination was below permissible limits (A0636). 2

CORE Advocacy notes that this incident, among others contained in the Incident Database 

that are documented after 1964 (the end of the SEC period), indicates poor worker monitoring 

practices that may compromise the ability to conduct dose reconstruction.

Given the nature of the incident involving contaminated smoke likely directed toward the 

ceiling and ventilation system, the floor should not have been the only surface sampled. In 

addition, workers should have been subjected to nasal smears, but there is no indication of 

such an evaluation. Bioassay samples should have been taken prior to workers leaving for the 

weekend, where several days passed before bioassay samples were collected. 

1 Internal Letter, Atomics International. “Radiation Safety Unit Weekly Newsletter for the Period Ending July 15, 1967” 

M.E. Remley - R.E. Alexander. July 25, 1967 Document retrieved from the publicly accessible historical document 

archive, courtesy of California Department of Toxic Substances Control (DTSC). File HDMSp01641279.pdf 

2 Incident Database, provided by Boeing. 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 Page 417 of 418 

19.0 CONCLUSION 

The 2006 Site Description is insufficient and inappropriate for use in dose reconstruction. It contains an 

incomplete list of DOE facilities, operations, processes, and incidents that resulted in environmental 

releases and worker exposures. 

CORE Advocacy has identified approximately 51 facilities that were excluded from the 2006 Site 

Description. The majority of these omitted facilities were considered “radiological.” Approximately 381 

radiological incidents were also excluded. The omission of these facilities and all corresponding data on 

processes, years of operations, releases, and incidents casts serious doubt upon the validity and integrity 

of SSFL environmental data currently used in TBDs 3-6. 

Among the omitted facilities are a nuclear reactor (HMRFSR), a hot laboratory (Building 4009), an extra 

particle accelerator (Building 4009), and a low-level radioactive waste incinerator which operated for 

nearly 25 years before being dismantled and disposed as radioactive waste (Building 4664). 

Also omitted are all references to a major nuclear incident that occurred in 1969 (S8DR, Building 4059). 

In addition, several processes and programs that involved complexes, installations and facilities of both 

radiological and non-radiological processes have been omitted. 

The current 2006 Site Description has provided a dramatically diminished perception of DOE operations 

and processes at SSFL. Moreover, it has likely contributed to a systematically underestimated probability 

outcome for the majority of SSFL dose reconstructions. Consequently, coupled with challenges SSFL

workers experience from the outset of the EEOICPA claims process, EEOICPA legislation continues to fail 

the employees of SSFL and its associated sites. 

CORE Advocacy identified numerous designated “non-nuclear” facilities that, throughout the history of 

site operations, engaged in radioactive processes without a “radiological” designation. This brings into 

question whether or not the facilities were properly licensed or designed to safely handle or dispose of 

radioactive materials. More importantly, the “non-nuclear” workers assigned to these locations do not 

appear to have been adequately monitored for radiation exposure. There are growing indications that the 

need to monitor workers was determined solely on facility designation. 

In 2014, Boeing acknowledged a “phenomenon” wherein approximately 8,400 “non-nuclear” SSFL

workers have “blank” radiation records in their employment files. 1 Based on Boeing’s explanation, it 

appears non-nuclear employees were anticipated to, at some point, require radiation monitoring. It does 

not appear that worker designations or monitoring practices were consistent or in sync with the ever 

changing operations that are documented at SSFL. This issue alone makes a compelling case for an 

expanded SEC. 

SSFL’s many complexities underscore the need for a comprehensive and robust site characterization. It is 

a privilege to submit the following document for your review and consideration for addition to the SSFL

Site Profile. 

Sincerely, 

D’Lanie Blaze 

CORE Advocacy for Nuclear & Aerospace Workers

COREAdvocacy.org

1 The Boeing Company, “Commentary on the ABRWH Meeting, Redondo Beach, California,” November 6, 2014. 

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description 
 


 


 


 


 


 


 


 


 


 


 


 


 


 


 


 


 


 


 

Page 418 of 418 


 

This document was compiled and prepared
 

in acknowledgment of all personnel of
 

Santa Susana Field Laboratory (SSFL)
 

whose contributions to science and technology
 

continue to guide our exploration of space,
 

influence our pursuit of energy alternatives on earth,
 

and lead our imaginations in our quest for answers
 

to the mysteries of the Universe.
 

Special Thanks:

U.W. Hopson -Rocketman

, , , ,

Terrie Barrie - ANWAG, Deb Jerison - EECAP

, , , 

[Redacted] [Redacted] [Redacted] [Redacted] [Redacted] [Redacted] [Redacted] [Redacted] [Redacted] [Redacted]

CORE Advocacy for Nuclear & Aerospace Workers • Proposed Corrections to TBDs 1 & 2 • SSFL Site Description