SC&A Status Update: SEC-00219 

CPP Class Definition Requiring Evidence of 

External Dosimetry (1963–1974) and 

Evaluation of Areas and Times that NIOSH 

has Determined Doses are Reconstructable

John Stiver

S. Cohen and Associates

Contractor to:

Advisory Board on Radiation and Worker Health/ABRWH

Centers for Disease Control and Prevention

November 18, 2015
Timeline of SEC 00219 Discussions

• INL Work Group Discussions

• April 22, 2015 – Clarification and Technical 

Teleconference between the Work Group, NIOSH, and 

SC&A (not open to the public)

• July 8, 2015 – Work Group Teleconference 

• November 10, 2015 – Work Group Meeting 

• Advisory Board Discussions

• March 26, 2015 – ABRWH Meeting 104 in Richland, WA

• July 23, 2015 – ABRWH Meeting 105 in Idaho Falls, ID

2 
SC&A Evaluation of CPP Class 

Definition Requiring Evidence of 

External Dosimetry (1963–1974) 
SC&A Revision 01 Evaluation Approach

Purpose/Goal: Evaluate if the revised SEC Class definition 

may unintentionally exclude workers from the SEC Class 

due to current dosimetry requirements.

1. Review all currently available claimants with at least 250 Days 

of covered employment 

• Split original SEC period into two components based on 

dosimetry requirements

• Identify claimants that did not meet dosimetry 

requirements in one or both SEC periods for further 

investigation

2. Investigate claimants who did not meet SEC dosimetry 

requirements to determine the potential for internal exposure 

to alpha-emitting contaminants at CPP

4 
Overview of Claimant Population (both 

periods) 

• At the time of its review, SC&A identified 898 total 

claimants with covered employment who worked in 

one or both portions of the SEC period.

• 107 claims did not meet the 250-day criteria in either the 

SEC period or the two periods combined (note: SC&A still 

evaluated 61 of these claims).

• 19 claims appear to only have worked at ANL-W and/or 

NRF (i.e. – no evidence of work at INL). 5 of these claims 

also did not have 250 days of employment.

• 2 claims did not have DOE monitoring records, because the 

claims were filed too recently

• 1 claim had been withdrawn prior to receiving DOE 

monitoring records.

5 
Breakdown by SEC Period: Later Period

(3/1/1970 – 12/31/1974) 

6 
Later Period (continued)

(3/1/1970 – 12/31/1974) 

SC&A had three observations based on it’s review of 

claimants in the latter portion of the SEC period:

• Observation 1: While the class definition provides the 

example of “at least one film badge or TLD dosimeter,” 

SC&A assumed in its review that any evidence of 

monitoring during the latter SEC period (3/1/1970–

12/31/1974) will satisfy the intended criteria. 

• Observation 2: SC&A identified a single claim that 

contained in-vivo dosimetry related to CPP, but did not 

have related external dosimetry. It is recommended that 

this claim be included with the claims requiring additional 

data capture at INL. 

7 
Later Period (continued)

(3/1/1970 – 12/31/1974) 

• Observation 3: Clarification is warranted to establish 

how “temporary” and/or “visitor badges” are utilized 

in the implementation of the class definition during 

the later SEC period (3/1/1970–12/31/1974).

• Clarification was provided by NIOSH at the November 

10, 2015 WG meeting - temporary and visitor badges 

and location file cards are adequate if 250 d 

requirement met

• Observations 1 and 2 raise questions about class 

implementation

8 
Breakdown by SEC Period: Earlier Period

(1/1/1963 – 2/28/1970) 

9 
Earlier Period (continued)

(1/1/1963 – 2/28/1970) 

SC&A had one additional observation related to the 

earlier SEC period:

• Observation 4: Absent additional information to the 

contrary, dosimetry associated with “CADRE” should be 

considered CPP for the purposes of determining SEC 

eligibility. Similar to the CPP dosimetry records, it is 

important to establish that “CADRE” badging records have 

all been captured from INL. 

• At the November 10 2015 WG meeting, the meaning of 

CADRE was discussed but not defined. NIOSH will 

reach out to site personnel for clarification

10
NIOSH and SC&A Cases Identified for Further 

Evaluation with DOE/INL

• NIOSH’s August 21, 2015, Work Group update 

identified 10 claims that required follow-up 

inquiries with INL

• At the November 10, 2015, WG meeting NIOSH 

indicated that 3 of the 10 had external dosimetry in the 

later period and would therefore be included in the SEC 

and did not need further investigation

• Based on its independent review of the claimant 

population, SC&A identified 23 additional cases 

requiring follow-up

• 11 of 23 recommended for follow-up with DOE/INL 

(Observation 6)

11
Summary and Path Forward

• Several Observations related to clarification of the SEC 

class definition:

• Use of annual/career records as well as other records, 

such as in-vivo results in place of individual dosimeters 

(later SEC period)

• Use of temporary and/or visitor badges

• If current class accepted, implementation may be 

challenging

• A combined 18 claims were identified by NIOSH and 

SC&A for further investigation to validate the class 

definition 

• At the November 10, 2015, WG meeting it was 

determined that additional investigation for the 18 

identified claims would be completed prior to the 

January 20, 2016 Board teleconference meeting

12
Evaluation of Areas and Times that 

NIOSH has Determined Doses are 

Reconstructable 
Introduction and Background

Dose Reconstructability/Gap Analysis

• Approach: “Horizontal” and “Vertical” analysis

• Horizontal – examine the DR methodology applied by NIOSH for 

all INL personnel – cross cutting

• Vertical – specific characteristics of the individual areas at the 

INL site

• 6 areas of investigation:

1. Reactor Modeling (horizontal)

2. Test Area North (TAN) (vertical)

3. Central Facilities Area (CFA) (vertical)

4. Fission and Activation Product (FAP) Bioassay Indicator 

Radionuclides (horizontal)

5. Burial Grounds (vertical) pended

6. Chemical Processing Plant (CPP) Pre-1963 (vertical) pended

14
INL Test Reactor Area

Nuclear Modeling 

• This investigation examines one aspect of the dose 

reconstructability assumption for one of the several 

major site areas, the Test Reactor Area (TRA)

• Based on SC&A report, NIOSH SEC-00219 Test Reactor 

Area Modeling, SCA-SEC-2015-0074-C, Rev. 0, 9/28/15)

• Applicability of ORAUT-OTIB-0054 to the more “exotic” 

reactors at Test Area North (TAN) is addressed in a 

separate report
TRA Overview

– Does the methodology of ORAUT-OTIB-0054 (Fission and 

Activation Product Assignment for Internal Dose-Related 

Gross Beta and Gross Gamma Analysis) adequately model the 

reactor characteristics and operations of the Test Reactor 

Area (TRA)?

o Are the reactors and operating scenarios adequately 

enveloped by the OTIB cases so that the isotopic ratios 

are valid? 

o Have all off-normal operating scenarios been identified 

and are they adequately enveloped by the OTIB 

methodology? 

16
ORAUT-OTIB-0054 

• Frequently, air-sampling or urinalysis data on worker 

exposure to mixed fission and activation products 

associated with nuclear reactors or fuel are available 

only in the form of gross beta or gross gamma activity 

unattributed to specific radionuclides. 

• The OTIB provides guidance on assigning radionuclidespecific intakes using ratios to Cs-137 or Sr-90 indicator 

radionuclides. 

• The goal of the OTIB is to reduce a large amount of 

reactor fuel isotopic data into a simplified, 

representative set that could be used by dose 

reconstructors looking at actual claimant cases. 

17
INL Test Reactor Area

Table 5-1 of the Evaluation Report lists the eight TRA reactors, only the 

first three of which were (or are, in the case of the ATR) high-power, 

high-flux reactors used for material testing. The remaining critical 

facilities were low-power reactors used for reactivity measurements 

and are not considered in this analysis. 

SC&A examined whether ORAUT-OTIB-0054 adequately envelopes the 

parameters of the three materials testing reactors, including the fuel 

and cladding composition, power level profile, burnup, and decay time 

following removal from the core. 

– Advanced Test Reactor (ATR)

– Materials Test Reactor (MTR)

– Engineering Test Reactor (ETR)

18
SC&A Evaluation – ATR

Advanced Test Reactor

• OTIB-0054 explicitly models the fuel characteristics 

and operations of the ATR, so it is expected that 

any workers exposed to ATR fuel during reactor 

operations or out of the reactor, but before any 

isotopic separation, would be adequately treated 

by the methodology in the OTIB in order to 

determine internal exposures. 

• SC&A did not find material instances of the ATR 

operating outside of its design envelope.

19
SC&A Evaluation – MTR

 The MTR fuel enrichment, cladding, and plate design were similar in 

important nuclear aspects to those of the ATR, and both reactors were 

light-water cooled and moderated, with beryllium and aluminum 

reflectors. Its power level and burnup, however, were considerably 

less than those of the ATR. 

Hence, the OTIB-0054 methodology, which explicitly models the ATR, 

should also adequately envelope the MTR in considering internal 

exposures when the latter reactor operated with uranium fuel. 

 Although the MTR initially used uranium fuel, in 1958 it became the 

first reactor run with a Pu-239 core. Phoenix experiment with Pu core 

conducted from January to April 1970 

How much different the Pu operations were, and whether the 

differences would be radiologically significant, would require detailed 

comparative ORIGEN runs, which were not done for this report. 

20
Engineering Test Reactor – ETR

• The ETR achieved first criticality on September 19, 1957, and 

operated at a maximum power level of 175 MW until December 

1981. 

• The reactor was light-water cooled and moderated and uranium 

fueled. Beryllium and aluminum neutron reflectors were placed 

outside the core. Initially, a new core contained about 14 kg of 

U-235, which was increased to 23.7 kg in January 1963. 

• As with the MTR operating with uranium fuel, the OTIB-0054 

methodology, which explicitly models the ATR, should also 

adequately envelope the ETR in considering internal exposures. 

21
Summary and Path Forward

• Resolve potential issues of applicability of ORAUTOTIB-0054 to MTR operating with plutonium fuel. 

• Prepare a prioritized list of other reactors that 

may fall outside the envelope of the OTIB-0054 

methodology (November 10, 2015, WG meeting)

22
Review of NIOSH Strategy for 

Reconstructing Doses to Workers at 

Test Area North (TAN) 

Based on SC&A report Review of NIOSH 

Strategy for Reconstructing Internal Doses to 

Workers at Test Area North (September 2015) 
Scope of Programs, Campaigns, Research and 

Activities at Test Area North (TAN)

• Aircraft Nuclear Propulsion Program (ANP) (1952–1961) 

• Initial Engine Test (IET)

• Heat Transfer Reactor Experiments (HTREs)

• Technical Support Facility (TSF)

 TAN 607 Hot Shop

 LOFT (TAN 650)

 Storage Pool

 Storage Pads (TAN 790 and 791)

 Radwaste Liquid Disposal System

 Storage Building

 Radiography Facility (TAN 607)

• Water Reactor Research Test Facility (WRRTF)

 Low Power Test Facility (LPTF)

 Shield Test Pool Facility (STPF)

• Specific Manufacturing Capability (SMC)

24
Focus and Scope of TAN Review

• An evaluation of the completeness of the captured 

external dosimetry data for the various TAN facilities

• Applicability of ORAUT-OTIB-0054 and Tables 5-22 

and 5-23 of internal dosimetry TBD to the 

performance of internal DR for facilities at TAN that 

handled and stored spent and irradiated fuel

• The unique circumstances associated with exposures 

to airborne effluents from the ANP, which are not 

addressed in OTIB-0054

25
Review of TAN External Dosimetry Data

• Although captured SRDB data represent only a 

sampling of site data, NIOSH nonetheless believes 

that they can reconstruct external doses at TAN – 

thus we reviewed the available data upon which this 

assertion rests

• Methods – Performed a search of SRDB records using 

search terms including: dosimetry, dosimeter, 

external, personnel, badge, exposure, and film

 37 documents

 12,177 pages of records

 181,183 beta/gamma readouts

 6,665 neutron readouts

26
Observations on External 

Dosimetry at TAN 

• External dosimetry for TAN as a whole appears to be fairly 

complete from mid-1955 through part of 1970, with a small gap 

from June through December 1961. Neutron dosimetry data 

appears spotty. 

• It is not possible to group the available data by subdivisions of 

TAN; i.e., the available data lack adequate granularity

 Important because activities and research throughout TAN 

were extremely diverse. It will be difficult to build external 

dosimetry co-worker models for each subdivision of TAN.

• Additional data capture will be required if the Board 

determines that a full completeness study is warranted

27
Use of OTIB-0054 to Reconstruct Internal Doses 

at TAN – Applicability

• TAN opened in 1952 for 

ANP, with support facilities 

and programs including TSF, 

IET, WRRTF. These facilities 

handled irradiated and 

spent fuel.

• Irradiated fuel and 

components from 

reactors/research facilities 

were often stored at TAN 

hot cells, hot shop, and 

storage pool. All of these 

facilities had potential for 

internal exposures

• Irradiated and spent fuels 

included: 

 ANP Fuel

 System for Nuclear Auxiliary 

Power (SNAP) and associated 

transients (SNAPTRAN)

 Special Power Excursion Reactor 

Test (SPERT)

 Loss of Fluid Test (LOFT) Facility

 Fuel from Heat Transfer Reactor 

Experiment (HTRE)

 Disassembly of SL-1 

 Storage of TMI fuel and debris

28
Methods*

• Performed ORIGEN simulations to determine if:

 The ratio of inventories of reference FPs to other FPs, as 

used in OTIB-0054 are reasonable, if not bounding, as 

compared to these ratios for TAN irradiated fuel

 The ratio of reference FPs to TRUs as in Tables 5-22 and 5-

23 of the TBD are reasonable, if not bounding, as 

compared to these ratios for TAN irradiated fuel

• Evaluated for the purpose of internal DR based on 

gross beta/gamma urinalyses

*Use of ORIGEN simulation and/or Tables 5-22, 5-23 are not considered 

useful/appropriate for internal exposure assessment for workers handling 

spent/irradiated fuel associated with ANP.

29
Observations/Conclusions

• Based on ORIGEN modeling for conventional reactor fuel, 

OTIB-0054 is generally claimant favorable when the fuel is not 

highly enriched, maintains integrity following burn-up, and is 

at a high power level (e.g., 200 MW)

• Underscores the importance of limiting our observations to 

general trends and consistent behavior

 For example, dose estimates based on a 200-day burn model 

will generally overestimate dose from actinides, when the actual 

reactor was operated for significantly less time

 Modeling exercise does not always comport well with SC&A’s 

analysis of actual measurements (bioassay indicator 

radionuclide study)

30
Aircraft Nuclear Propulsion Program - Heat 

Transfer Reactor Experiments (HTREs)

• To test viability of a reactor for aircraft propulsion, 

3 different reactors were built (HTRE 1, 2, and 3)

• All 3 were direct cycle, air-cooled

 Air from turbojet engine compressed and forced past 

wafer-thin concentric ribbons of nuclear fuel enriched 

to 93.4%

 Air heated to 1,250 °F by fuel temps of up to 3,000°F 

powered turbine engine

31
Initial Engine Tests (IETs)

• Testing program for 3 HTRE assemblies designated as 

Initial Engine Tests (IETs), IET#1 through IET #26

• IETs #1, #2, #5, #7, and #9 did not require nuclear 

power, and had no potential for release of 

radioactivity or human exposure

• IETs #1 and #3 – plumes went offsite - issues closed in 

INL WG discussions

• IET #10 issues still open – NIOSH will prepare a white 

paper (November 10 2015 WG meeting) 

32
Observations and Recommendations

• Independent analyses of airborne emissions 

associated with the major IETs performed by 

SC&A under contract to CDC revealed that the 

DOE significantly underestimated the airborne 

emissions for the IETs with the largest airborne 

emission

• Outdoor exposures associated with releases from 

the ANP need to consider the results of CDC’s 

investigations into these source terms

• Challenges associated with reconstructing 

outdoor onsite exposures associated with these 

releases 

33
Evaluation of Available Survey Data 

for the Central Facilities Area at 

Idaho National Laboratory

Based on the SC&A report: 

Evaluation of Available Survey Data 

for the Central Facilities Area at 

Idaho National Laboratory 

(September 2015)
Introduction and Background

• At the July 8, 2015, INL WG Meeting, SC&A presented a status update 

on their evaluation of the INL SEC Petition Evaluation Report for SEC00219, originally issued by NIOSH on March 14, 2015, and revised on 

July 21, 2015.

• SC&A recommended that the available survey data for the Central 

Facilities Area (CFA), both during operations and prior to demolition 

and dismantlement (D&D), be evaluated to determine the actinide to 

Sr-90 and actinide to Cs-137 ratios and compare these ratios to the 

values in Tables 5-22 and 5-23 of ORAUT-TKBS-0007-5. 

• This presentation provides a summary of NIOSH’s proposed methods 

for bounding operational period internal doses for the CFA, a review 

of the available survey data, and a comparison of the derived values 

to the values in Tables 5-22 and 5-23 of ORAUT-TKBS-0007-5. 

35
Proposed Methodology to 

Assess Actinide Intakes

• Section 7.2.5.2, SEC-00219 ER - describes NIOSH’s methods 

for bounding operational period internal dose for CFA

• Potential missed intakes of uranium, thorium, and 

plutonium for CFA workers are of particular interest

• Proposed method uses mixed fission product (MFP) and 

mixed activation product (MAP) intakes determined from 

the beta-gamma bioassay data to assess actinide intakes 

using the actinide to Sr-90 and actinide to Cs-137 ratios 

and Tables 5-22 and 5-23 of ORAUT-TKBS-0007-5 

36
CFA Survey Data

• Conducted a search of the SRDB for radiological 

surveys of the CFA facilities

• 1954–1956 Contamination Surveys of CFA-669 

Hot Laundry and CFA-674A Chemical Engineering 

Lab (SRDB 139224)

• Post D&D soil sample results from excavation of 

the contaminated sanitary sewer line on the 

north side CFA-669 (SRDB 088224)

37
Contamination Survey Data

• Only beta-gamma and alpha results greater than 

background levels (beta-gamma and alpha) at the same 

sample location were considered. 

• Eighty-five contamination survey results were found that 

met our criteria (listed in Appendix A of report).

• Six smear results from April 12, 1954, are not consistent 

with the other results and may have the alpha and betagamma results transposed. They were not included in the 

analyses.

• Original survey results are given in units of cpm. Limited 

counter efficiency information was used to estimate 

activity

38
D&D Soil Samples

• 19 soil samples from the excavation of the 

contaminated sanitary sewer line on the north 

side CFA-669

• Analyzed for alpha and gamma spectrum and 

strontium-90

• The U-234 and Th-228 levels were not 

significantly different from environmental 

levels 

39
CFA Summary 

• Analyses of the smear data and soil sample results 

indicated general agreement in the magnitude of the 

maximum contamination ratios for U-234 and Pu-238 

given in Tables 5-22 and 5-23 of ORAUT-TKBS-0007-5.

• There are several limitations in the data used:

 The survey data found was very limited and from period of early 

operations. 

 The survey data is written given in units of cpm. Instrument 

parameters uncertain.

 The soil samples were collected during D&D operations. Prefer 

characterization survey, prior to D&D.

 SC&A data capture in January 2016 time frame will seek 

to resolve some of these uncertainties

40
FAP and Actinide (Pu, U, etc.) Intakes 

per NIOSH’s SEC ER

SEC ER: 4 Fundamental assumptions form the basis for 

NIOSH’s ability to reconstruct internal doses 

A. FAP Bioassays – Sufficient workers’ records containing FAP bioassay (in-vitro and in-vivo) results are available to assign intakes and resulting doses from FAP (some 

periods/areas may need an FAP coworker model 

developed). 

B. FAP Intakes – Except for special situations, all the dosimetrically significant FAP intakes are directly tied to 

an indicator radionuclide (Sr-90 or Cs-137). The FAP ratios and intake assignment methods provided in 

ORAUT-OTIB-0054 bound all FAP exposure potentials at 

INL. 

41
FAP and Actinide (Pu, U, etc.) Intakes per 

NIOSH’s SEC ER (cont’d) 

C. Actinide Intakes – Except for special situations, the actinide 

(uranium, plutonium, thorium, etc.) intakes are directly tied 

(in a constant ratio) to the FAP; therefore, actinide intakes 

and resulting doses can be assigned using Table 5-22 (Sr-90 

ratios) and/or Table 5-23 (Cs-137 ratios) of ORAUT-TKBS- 0007-5 (TBD-5).

D. Special Situations Actinides – Personnel involved in 

operations and situations (planned or unplanned) with 

actinides present, that were not directly tied to an FAP in a 

constant ratio, were adequately monitored, and the results 

are available in the workers’ records. Therefore, actinide 

intakes and resulting doses can be reconstructed in these 

special cases.

42
Item A - SC&A Review Approach

• Evaluate a semi-random* sample of INL claimants 

to assess the adequacy and completeness of 

individual records for the purposes of dose 

reconstruction

• Were all relevant workers monitored for FAP?

• Were monitored worker records complete?

• Are coworker models appropriate for areas and time 

periods other than those already designated?

43

*Selected claimants were biased towards claims with multiple employment periods 

during SEC Evaluation period.
Item A - SC&A Review Approach (cont.)

• SC&A determined that there were 973 claimants who 

had covered employment at INL during the evaluated 

SEC period (921 claims with SEC employment greater 

than 90 days)

• SC&A semi-randomly selected and reviewed 92 

claimants from this population (roughly 10%)

• More than 60 percent were trades workers 

• This category had the largest number of employment periods 

per worker and are therefore over-represented in the 

sampling (semi random)

44
Item A - Summary Conclusions and 

Recommendation 

Based on SC&A’s review of 92 randomly selected claimants, it 

was evident that fission and activation product bioassay is 

generally available for a wide variety of job titles. Thus SC&A 

does not believe there are “completeness” issues with the 

dataset of fission and activation product bioassay that would 

preclude its use in developing coworker models. Nor was there 

any indication that specific job titles were systematically 

excluded from the internal monitoring program. However, it is 

SC&A’s opinion that FAP coworker models should be evaluated 

and developed for each relevant INL site area beginning with 

the start of radiological operations for each individual location. 

45
Item A - Issue Status Update

• The issue of coworker model applicability for years 

and areas not originally designated by NIOSH was 

discussed at the November 10, 2015 Work Group 

meeting.

• NIOSH agrees that coworker modeling may be 

appropriate and will assess coworker requirements 

and feasibility for the applicable INL site areas and 

years prior to 1967.

46
47

Items B-D Review - FAP and Actinide Intakes

Evaluation of Ratios using Actual Measurements 

• NIOSH’s ER recommends using Sr-90 and/or Cs-137 bioassay results in 

conjunction with ratios in OTIB-0054 to assign FAP intakes.

• NIOSH’s ER recommends using Sr-90 and/or Cs-137 bioassay results in 

conjunction with ratios in TKBS-0007-5 to assign actinide intakes.

• The ratio values were derived mostly by computer simulation (ORIGEN)

• SC&A searched for documentation that would provide measured 

radionuclide ratios

NOCTS

 SRDB

 INL electronic bioassay database
48

Evaluation of Ratios (continued)

 SC&A located measured quantitative 

radionuclide analyses of:

Nasal swabs

Urinalyses

Fuel element scale

Fuel storage contamination swipes

Air filters

 A total of 42 samples were located and 

analyzed for radionuclide ratios.
49

Summary of Results

1. FAP intakes assigned using NIOSH’s recommendations in OTIB-0054 based 

on Sr-90 intake values are generally (but not always) equal to, or greater 

than, those derived from actual measured values.

2. The Cs-137/Sr-90 ratios are not always 1:1 as assumed in OTIB-0054 and 

TKBS-0007-5; frequently, large variations in the ratio exist. This brings into 

question the validity of using an indicator radionuclide when deriving FAP 

and actinide intakes. This may be the most important result of this study, 

because a Cs-137/Sr-90 value of 1:1 is one of the cornerstones for use of 

the ratio method at the INL. 

3. Actinide intakes assigned using NIOSH’s recommendations in TKBS-0007-5, 

Table 5-22, based on Sr-90 intake values, or Table 5-23, based on Cs-137 

intake values, are sometimes significantly less than those derived from 

actual measured values.

4. Special bioassays. It is difficult to evaluate when “special” (situations 

where actinides were not tied to a fission product in a given ratio) bioassays 

were needed, if they were performed, and if they are indicated as such in 

the bioassay records.
50

Recommendations and Path Forward

1. Determine if records of analyses of dissolver contents (containing the 

fuel elements) are available; preferably for a variety of INL reactor fuel 

elements, and also fuel elements from off-site reactors.

2. Conduct Further INL document research to evaluate NIOSH’s 

recommended ratio values, especially for actinides and Cs-137/Sr-90. 

Records with quantitative radionuclides analyses are especially 

important. 

3. Determine if special or non-routine bioassays were associated with 

special exposure events, such as are referred to in the ER, or if instead 

the term “special,” or “non-routine,” bioassay was applied to the 

priority of processing over “routine” bioassays.

4. SC&A data capture trip in the January 2016 time frame will seek to 

address these concerns. The report will then be revised 
Burial Grounds and CPP Pre-1963 

Analyses and Reports pended until data 

obtained from January 2016 capture can be 

analyzed 

Recap of Concerns/Path Forward

51
Burial Grounds

Concerns

• Evidence exists that a “strict” contamination control 

program was not in place

• Site apparently lacked adequate smear counting capability 

for some length of time before early 1970s

• Radioactive waste was not specifically identified for most 

drums, boxes, and other containers in the early years

• Offsite waste received from commercial, university, ERDA, 

and military sources in 1960–1963 inadequately identified 

• AEC concerned over conflicted role of health physicists at 

the Burial Grounds, who were responsible for much of its 

operation, as well as radiation protection 

• Internal investigations and appraisals bring into question 

“robustness” of HP program and “defense-in-depth” 

approach for radiological controls, as cited in the ER

52
Burial Grounds (cont’d)

Path Forward

• SC&A will conduct a data capture and worker interview trip 

to Idaho Falls in the late January 2016 timeframe

• Conduct additional interviews with former Burial Grounds 

workers with experience during the time period in question 

(1952–1970); emphasis on radiological control program

• Conduct additional data capture with focus on the following:

• Additional evidence of potential intakes to radwaste handlers

• How contamination control was administered

• Available routine and special air-sampling data 

• Robustness of health physics program: independence, 

resources, monitoring practices

53
Burial Grounds (cont’d)

Path Forward, Cont’d

• Evaluate dose assessment feasibility 

• Review external and internal dose electronic 

database when completed by NIOSH. 

• Review historic bioassay procedures and practices. 

• Can all Burial Grounds workers be identified?

• Can all significant radioactive waste source terms be 

identified and addressed?

54
Chemical Processing Plant (Pre-1963) 

Concerns and Focus of SC&A Investigation

• Evaluate contamination incidents and control 

program prior to 1963

• Assess internal dosimetry program for CPP 

workers 

• Characterize temporal changes in source term 

and exposure potential

Status:

• SC&A has reviewed site records available on the 

SRDB and performed a partial claimant survey

55
Chemical Processing Plant (Pre-1963) 

Status (cont.):

• Determined further data capture was warranted: 

• Contamination surveys (specifically alpha surveys)

• Incident reports

• Radiation safety reporting and practices

• Source and exposure potential documentation for alpha 

emitting material

• Interviews and data capture will be conducted during the 

January 2016 trip to INL to gain information on the 

potential for alpha-only internal exposures at CPP

56
Questions Comments?

57