ML20046B501

Evaluation of Potential Internal Radiation Exposures Associated W/Contamination Events in Handling Traversing In-Core Probe (TIP) Tubing at Limerick Nuclear Generating Station & Recommendations for Evaluation & Control of .

ML20046B501
Person / Time
Site:	Limerick
Issue date:	06/27/1993
From:	Skrable K PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
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ML20046B483	List: ML20046B482 ML20046B501
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NUDOCS 9308040295
Download: ML20046B501 (96)

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i EVALUATION OF POTENTIAL INTERNAL RADIATION EXPOSURES ASSOCIATED WITH CONTAMINATION EVENTS IN HANDUNG TRAVERSING IN-CORE PROBE (TIP) TUBING

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I AT UMERICK NUCLEAR GENERA 11NG STATION AND RECOMMENDATIONS FOR THE EVALUATION AND CONTROL OF EXPOSURES ALARA i

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Kenneth W. Skrable June 27,1993 i

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1. INTRODUCTION On January 27,1993, during work on the traversing in-core probe (TIP) tubes at the Unit 2 Limerick Nuclear Generating Station, radioactive contamination released during the work caused significant contamination of the area and facial contamination of four workers. Despite the fact that respiratory protection equipment was not used to protect the workers, subsequent whole body counting results indicated that no significant intake or deposition had occurred in these workers. This incident resulted in a Notice of Violation from the Nuclear Regulatory Commission (NRC) including (1) failure to make adequate surveys to determine that individuals were not exposed to airbome concentrations exceeding the limits specified in 10 CFR 20.103 and (2) failure to adequately inform workers as to the presence of high levels of radioactive contamination or of means to minimize their exposure to such contamination. Specifically, at about 2:00 a.m. on January 27,1993 during mmoval of TIP tubes, the licensee failed to detect the introduction of high levels of radioactive contamination (subsequently measured to be as high as 320 mrad h'1 per 100 cm of removable contamination)into the work area as the TIP tubes were removed. As a result, the NRC concluded that there was a potential for workers performing the task, without benefit of respiratory protective equipment, to sustain a significant intake of rad.ioactive material. As a result of this incident and other findings by the NRC inspector, a subsequent enforcemer conference was held on March 16,1993 tm discuss the safety significance of the licensee's overall performance in the area of radiological controls over the past year, the specific violations, and corrective actions.

This repon provides an evaluation and comparison of the methodology used in the NRC Enforcement Conference Repon Nos. 50-352/93-04 and 50-353/93-04, including the supplement dated 3/29/93 and that in the report by Philadelphia Electric Company (PECo), " Calculation of Airborne Levels Undervessel as a Result of TIP Tube Event," both of which give different scenarios for the evaluation of the potential exposure from this incident after the fact. This report also provides recommendations for the evaluation of potential exposures from combined external and internal sources. Such evaluations are expected to help identify radiation protection requirements considering the expected combinea dose fu i both types of sources.

For a given job or work activity, engineering controls, procedures, and monitoring requirements should be chosen to help assure that the total effective dose equivalent (TEDE) from internal and external sources is maintained as low as reasonably achievable (ALARA),

which is a requirement in the revised 10 CFR 20 regulation effective on January 1,1994. In situations involving exposures to intemal and external sources that are well documented, under i

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control, and properly monitored, and where the external exposure rate greatly exceeds the internal exposure rate, use of respiratory protection devices that hinder workers and increase their total exposure time to accomplish a task may cause workers to receive a TEDE significantly greater than that received without the use of respiratory protection equipment. It also is to be I

noted that past and cunent practices, based upon a control action level requirement in the extant l

10 CFR 20 regulation and undue concern for internal exposure, have often caused workers to receive a TEDE from combined internal and external sources that was not ALARA.

2. DISCUSSION I

2.1 Comparison of Requirements in the Extant and Revised 10 CFR 20 Regulations The requirements for the evaluation and control of extemal and internal exposures under j

the extant and revised 10 CFR 20 regulations differ in several aspects. While the regulatory j

significance of the TIP tube contamination incident must be determined on the basis of the extant i

regulation, my recommendations for the evaluation of potential exposures are based on the

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revised regulation, which becomes effective on January 1,1994. Major differences in the extant O

i and revised 10 CFR 20 regulations are sunimarized as follows.

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The extant regulation provides quarterly intake limits separate from and independent of the external limits specified in a 13 week control period while the revised regulation limits the combined doses from intemal and external sources to certain values from exposures within a one j

year control period. Limits within the extant regulation are based upon recommendations in i

ICRP Publication 2, while those in the revised regulation are based upon recommendations in i

i 1CRP Publication 26 and ICRP Publication 30. Those portions of the extant and revised j

10 CFR 20 regulations that most directly relate to the evaluations and recommendations in ttus l

report are quoted in Appendix A, which also includes pertinent comments relative to their meaning and application. Major differences in the extant and revised regulations are summarized in Appendix A and below as follows.

A comparison of the requirements for the evaluation and control of intema2

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exposures in the extant and revised 10 CFR 20 regulations shows them to differ in the following f

signi6 cant ways.

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2.11 Comparison of the Limits in the Extant and Revised 10 CFR 20 Regulations The extant regulation limits the intake of radionuclides to cenain quarterly intake limits (QILs) independent of any external dose received by workers. The QIL is 1/4 of what would l

correspond to an annual intake limit (AIL) for exposure of Standard Man, which is based upon a j

committed dose to the critical organ corresponding to the maximum permissible dose (MPD) j recommended by the ICRP. Each QIL corresponds to an exposure of 500 MPC-hours ot Standard Man and not the 520 MPC-hours commonly applied by NRC inspectors and the nuclear l.

industry (See comments in Appendix A.). The revised regulation limits the total effective dose i

equivalent (TEDE) to the whole body and dose equivalent to any organ or tissue from combined external and internal sources to respectively 5 rems and 50 rems to the actual exposed workers in any year of practice. Considerably flexibility is given for the acceptable methods for i

demonstrating con pliance with hese dose limits (See comments in ' Appendix A.). Appendix B provides guidance on the evalcation and control of internal radiation exposures on the basis of

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the committed uose limits in ICRP Publication 30 and the revised 10 CFR 20 regulation.

2.12 Comparison of Requirementsfor Monitoring and Assessment of Exposures O

The extant regulation requires that exposures to airborne internal radiation sources be j

. monitored, assessed, and recorded when workers enter an airborne radioactivity area and when j

their exposure in any day exceeds 2 MPC-hours or in any week 10 MPC-hours. The revised regulation requires monitoring, assessments, and recording of intemal radiation doses when it is likely that the intake in any year could exceed 10% of the applicable annual limit on intake I

(ALI). Experience in the U.S. nuclear power industry shows, in fact, that workers are not likely j

to exceed 10% of the ALI. Provided that transuranic radionuclides do not make a significant contribution to the mix of radionuclides, exit portal monitors at nuclear power stations are in i

most cases capable of confirming this fact. Significant exposures of workers above the extant quarterly intake limit have not occurred despite the fact that accidental exposures in situations involving high area surface contamination levels have sometimes been easily detected from the external contamination monitoring of workers, air sampling results, and/or through bioassays.

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,0 213 Comparison of Control Action, Investigation, and ALARA Requirements t

The extant regulation requires the use of respiratory protective equipment if the projected i

exposure in any seven consecutive days is expected to exceed 40 MPC-hours and all other means have already been tanen to limit such an exposure. If an exposure is found after the fact to exceed this control action level of 40 MPC-hours, then an investigation is required to prevent a f

t recurrence and to document the circumstances involved, among other things. Although the j

revised regulation does not specify similar control action and investigation levels,it does require i

that the total effective dose equivalent (TEDE) from combined exposures be maintained

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ALARA. Thus, if the use of respiratory protective equipment or other procedures to limit the internal exposure are shown to cause workers to receive a larger TEDE than the value anticipated when such equipment /and or procedures are not used, then such equipment and procedures should not be used. Such use would not be ALARA, and it might be interpreted as a violation of j

the revis:d 10 CFR 20 regulation (See coinments in Appendix A.).

l The formal requirement in the revised 10 CFR 20 regulation to maintain the total effective t

dose equivalent from combined exposures to intemal and extemal sources ALARA is intended to f

overcome the ALARA deficiencies in the extant 10 CFR 20 regulation. The proper j

implementation of this new ALARA requirement will not be easy. It will require appropriate j

evaluations of potential exposures to internal and external radiation sources. To maintain a proper balance of the requirements for the evaluation and control of external versus internal radiation exposures, training programs for all staff, including management of nuclear facilities will be required to overcome the many misconceptions and the undue concern about internal l

radiation exposures. The Nuclear Regulatory Commission will have to implement similar training programs for inspectors and other NRC staff responsible for implementing the revised l

10 CFR 20 regulation including regulatorv guides on the proper evaluation of potential l

exposures for helping to assure that the teral efective dose equivalent (TEDE) is in fact j

maintained ALARA. It is obvious that honest mistakes will be made and that some of these mistakes, no doubt, will be caused by the undue concerns and perceptions about internal l

radiation exposures.

l These significant differences in the current and revised regulations influence the

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determination of the regulatory significance of the contamination incident resulting from work on the TIP tubes as well as recommendations regarding the evaluation and control of potential t

i radiation exposures to external and internal radiation sources.

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r 2.2 Surface Contamination Guidelinesfor the Evaluation of Potential Exposures, Guidance Q

for Air Sampling in the Workplace, and the Efficacy ofRespiratory Protection i

f Various guidelines have been used for radioactive surface contamination by national and international organizations and individual nuclear facilities for various purposes A summary by l

q Klein and Schmidt " Limits for Radioactive Surface Contamination," indicates that the literature i

contains only a few cases where work in contaminated areas has resulted in a detectable body burden (See Appendix C.). Recommended contamination limits for the same application a

category vary by factors of up to 1,000. Negative findings have been found for situations involving very high levels of contamination, e.g., significant radium burdens were not observed, l

in occupants of a home where the contamination was found in the range of 10,000 dpm/100 cm~

to 100,000 dpm/100 cm with peak levels greater than 1,000,000 dpm/100 cm-They conclude:

"From the limited data available it appears that the surface contamination limits in use in the j

United States are conservative and have been effective in minimizing radiation exposure."

i Nuclear power plants in the United States use conservative contamination guides for posting and establishing protection requirements that typically range.

(1) posting as contaminated for areas > 1,000 dpm/100 cm-O t

(2) requirement for some protective clothing for areas > 10,000 dpm/100 cm-l' (3) requirement for additional protection, possibly respirators for areas > 50,000 dpm/100 cm~.

I am not aware of a single exposure in the commercial nuclear power industry that has ever exceeded the quarterly intake limit in the extant 10 CFR 20 regulation. No doubt, the use of respiratory protection at the control action level shown in 3 above has caused workers to receive a total effective dose equivalent (TEDE) greater than what would have been received if no protection had been used. More realistic procedures established at the Davis-Besse Nuclear Generating Station to meet the ALARA requirements of the revised 10 CFR 20 regulation have demonstrated a saving in the TEDE through the use of considerably less respiratorv protection i

1 than in the past Almost no circumstances occurred in a recent maintenance program where respiratory protection was mund to be ALARA.

In its revised 10 CFR 20 regulation, the Nuclear Regulatory Commission is trying to address the problem of overprotection of workers from intemal radiation exposure at de expense

) Personal communication with Regis Greenwood, telephone: 419-321-8437.

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O a greater externai radiation exposun: by requiring that the Teos from comsined externai and 1

f internal exposures be maintained ALARA. Specific guidance has been provided by the NRC that might be helpful in this regard including (1) a January 1990 NRC report, NUREG C/R 5512,

" Residual Radioactive Contamination for Decommissioning - Technical Basis for Translating l

Contamination Levels to Annual Dose," and (2) Regulatory Guide 8.25, '" Air Sampling in the Workplace."

As stated in Regulatory Guide 8.25, reactor licensees are exempted from the guide:

In addition, this guide does not apply to activities conducted under 10 CFR Part50 at t

i reactorfacilities. Although the provisions of10 CFR Part 20 apply equally to nuclear reactors and to other facilities, the air sampling programs of reactor licensees are well established, and the NRC is satisfied that the quality of air sampling at tuclear reactors is l

adequate. Therefore, no further guidance on air sampling is needed at this time for reactor licensees.

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Despite this disclaimer, Regulatory Guide 8.25 is an excellent guide for implementation of i

requirements of the revised 10 CFR 20 regulation. This guide is supported by the NRC report 7

NUREG-1400, which is described in the guide:

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Draft NUREG-1400, " Air Sampling in the \\Vorkplace," provides examples, methods, and techniques that the licensee mayfind usefulfor implementing the recommendations in this guide.

Of particular significance to the evaluation of potential internal exposures from l

contamination released in the TIP tube incident is Section 1.1 of the Regulatory Guide 8.25, j

When to Evaluate the Need for Air Sampling, where it is quoted partly:

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As a general rule, any licensee who handles or processes unsealed or loose radioactive materials in quantities that during a year will total more than 10,000 times the AUfor inhalation should evaluate the needfor air sampling.. When quantities handled in a year are less than 10,000 times the AU, air sampling generally is not needed. (The basisfor tnis value is that experience has shown that worker intakes are unlikely to exceed one-millionth of the material being handled or processed, as discussed in NUREG-1400.)

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Q Other NRC reports that may be useful in the evaluation and control of internal radiation exposures include:

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Regulatory Guide 8.9 Revision 2, " Interpretation of Bioassay Measurements."

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NUREG/CR-4757, "Line-Loss Determination for Air Sampler Systems."

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NUREG/GR-006, " DEPOSITION: Software to Calculate Particle Penetration Through i

Aerosol Transport Lines."

In addition, several ANSI standards are being considered that will hopefully address the

- i evaluation of potential exposures from external and internal radiation sources. A recent ICRP i

publication titled, " Protection from Potential Exposure: A Conceptual Framework," does provide some general guidance in this regard (ICRP Publication 64, 1993). The ICRP states in this report:

(3) The purpose of this report is to elaborate upon the principles and objectives ofICRP recommendations as they relate to potential exposure; explain the basic concepts, j

terminology and methodologies associated with the application of the recommendations; i

and provide general guidance on its practical application. The report is intended to j

l provide a basisfor the preparation ofmore detailed guidance related to specific practices.

Hopefully, this ICRP publication will be followed up soon by the detailed guidance needed for its properimplementation by the nuclear industry.

With respect to the evaluation of the efficacy of respiratory protection in light of the-e requirement to maintain the total effective dose equivalent ALARA, the following publications may oc userui:

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Quealy, Patrick D., " Efficiency of a Worker Not Wearing a Respirator Versus Wearing A l

Respirator," Northwestern University Masters Project Advisor, Herman Cember, Spring l

1993.

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Greenwood, R. A. and O'Dou, T. J., " Dose Expansion from Using Respirators,"

i Radiation Protection Management, Volume 9, Number 4, August,1992.

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Johnson, A. T., Grove, C. M., and Weiss, R. A., " Respirator Performance Rating for Nontemperate Environments," Am. Ind. Hyg. Assoc. Journal,53(9), pp. 548-555,1992.

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Early, J. N., " Profession,al Spide: Disregarding The Respiratory Protection Realities of the Revised 10 CFR 20," Radiation Protection Management, Volume 9, Number 4, August 1992.

2.3 Comparison of PECo's and the NRC's Evaluation of the Potential Internal Radiat;on l

Exposure Associated with the January 27,1993 TIP Tube Contamination Incident Inherent in the NRC cited violations for the TIP tube contamination incident is the NRC 1

l conclusion that there was potential for workers performing the task, without benefit of l

l respiratory protective equipment, to sustain a significant intake of radioactive material.

Presumably, by a "significant intake" the NRC means one above the quanerly intake limits. On page 3 of the NOTICE OF VIOLATION,it is stated by the NRC:

In the NRC inspection report transmined to you by leuer dated February 26,1993, we recogni:e that you performed a thorough investigation on the January 27 event. However, afterfurther review during the inspection on March 17 ofyour evaluation methodology of the potentialfor overexposure of workers, we have determined that your evaluation failed to identify the magnitude of the radiologicalproblem. This was because your calculation only took into account a small amount of the total radioactive material on the subpile room floor or a small amount of the total radioactive material on the grating and on the walls.

While it is our conchision that your overall event investigation was strong, it is also our conclusion that your evaluation of the potentialfor a substantial exposure in excess of the regulatory requirements was weak.

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This section provides a comparison and evaluation of the assumptions and methodologies in the NRC and PECo reports used to evaluate the potential exposure associated with the January 27 TIP tube contamination incident. The problem hinges on (1) the appropriate estimate, after the fact, of the total activity potentially available or actually released to the air environment of the exposed workers and (2) the evaluation of a number of possible scenarios, again after the fact, leading to exposure of individuals via various pathways.

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0 231 Activity Conversion Factors per Unit lon Chamber Response, NRC Estimate of Total Activity Released and Present on Subpile Floor, Quarterly intake Limit (QiL) and Corresponding MPC for Mixture ofRadionuclides The NRC inspector made estimates of the removable activity per unit area of the subpile floor from a correlation between the licensee's gamma scan results and the dose rate reading of a smear over 100 cm~ of the floor to estimate the quantity of removable radioactive contamination of each radionuclide on the subpile room floor. I agree with the methodology used by the NRC inspector as being a reasonable estimate of the removable activity of each radionuclide on the subpile floor as shown in Table 1 on page 6 of the NRC Supplement to Report Nos.

50-352/93-04 and 50-353/93-04. I also agree with the inspector's " correction factor", which is

-2 calculated here to three significant digits as 1.45x10"#

Ci cm per mrad h to convert ion chamber instrument readings in units of mrad h~1 of smears over 100 cm to an estimate of removable surface activity in units of pCi cm'~. It is also reasonable to use a dose rate to total activity conversion factor of 1.45x10~' yCi per mrad h~1 to conven an instrument reading of a i

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smear or of a Q-tip swipe of a TIP tube to estimate the total activity on such samples. From the average instrument reading of 64 vad h"I and the instrument conversion factor, the average specific activity S f rem vabb contamination on the subpile floor is calculated:

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-4 (64 mrad h~1) (1. 45x10 pCi cm' per mrad h~ );

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=

g S

= 3.23x10 ~

pCi ;m~

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+

w'.ich gives a total activity A on the subpile floor estimated to have a total area of 315 ft or 5

9 2.93x10 cm -

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-2) (2. 93x100 2

cm ) = 2,720 pCi a 2,700 pCi, A=

(9.28x10 yCi cm 1

which is the value shown in the NRC report.

The quarterly intake limits shown in Table 1 for the identified principle radionuclides are l

calculated from equation 1 in Appendix A using the more restrictive insoluble MPC values in the extant 10 CFR 20 regulation. I agree with this choice. Thus, the total activity A estimated to be f

present on the subpile floor is shown correctly on page 6 and in Table 1 as about 2,700 Ci or 11.63 times the quarterly intake limit (QIL), which henceforth will be rounded and designated as A = 11.6 QIL. Thus, the quarterly intake limit or QIL for the mixture in activity units is O

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calculated from the quotient of the total activity A of 2,700 pCi by 11.6 QIL:

QIL = (2,700 pCi) / (11. 6) = 233 pCi.

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l The applicable MPC for the mixture of radionuclides is calculated from the quotient of 8

3 this QIL value and the volume 6.3x10 cm of air inhaled in 1/4 year by Standard Man:

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Ci cm 3.70x10 (233 yCi) / (6.3x10 cm )

MPC

=

=

t on page 15 of the PECo report gives an activity conversion factor of 0.0145 pCi cm-1 per mrad h'I, which is in agreement with the factor of 1.45x10-2 Ci per mrad h-I shown above, except for the incorrect inclusion of a em'I unit in the PECo conversion factor.

This extra incorrect unit, however, was not included in any of the calculations shown in attachments 1 - 6, conesponding to the 6 scenarios used in the PECo report to evaluate the l

internal exposure potential from the TIP tube contamination incident. With respect to the i

conversion factor shown in attachment 7 for obtaining activity estimates of Tc-99m plus Mo-99 l

from ion chamber readings of smears, I do not believe that such a conversion factor is very useful because the ion chamber response is for the total activity present and the contribution of Tc-99m.

i because of its much larger QIL, is insignificant compared to Mo-99. Although Mo-99 is the ines: agniE=: ruicr.alide prsent in the inixture of radionuclides found in the gamma scan of O

the smear sample, from the standpoint of its amount relative to its QIL, other radionuclides are l

shown to make a signincant contribution in the NRC Table 1.

232 NRC Estimates ofInternal Exposure Potential i

i On page 7 of the NRC report supplement it is stated with respect to the total removable I

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activity on the subpile floor of 11.6 QIL:

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The inspec:or estimated that suf]icient activity was present, if entirely taken into the body, i

to result in an intake of radioactive material of at least 11 times that permined by 10 CFR 20.

1 Although this statement is correct,it is entirely unrealistic to assume that 100% of this activity on the floor would be inhaled by a single individual. A maximum fraction for these radionuclides, all of which are quite refractory and non-volatile, might be assumed equal to lx10-6 This value is given in the NRC Regulatory Guide 8.25 mentioned in section 2.2 above. Application of this O

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Q value of 1x10-6 as a conservative intake fraction would make the potential exposure equal to an intake of 1,1x10~ QIL, which is completely insignificant with respect to the regulatory limit or j

any reference level in 10 CFR 20. This maximum intake estimate would correspond to an I

exposure of only 0.0055 MPC-hours. This alternative realistic after the fact assessment, based upon actual experience, thus would show the internal exposure potential to be essentially nil.

i Use of respiratory protective devices, which cause an increase in the exposure time of workers, l

i would be counter productive and would cause an increase in the external dose and TEDE. Such use would be contrary to the ALARA requirements in the revised 10 CFR 20 regulation.

1 In another scenario shown on page 7 of the NRC report supplement,10% of the total 3

activity on the subpile floor is imagined to be released into an air volume of 64 ft where it is calculated to "resuh m a short term gross airborne radioactivity concentration of about 1.5 x10'# pCi/cm in the area." It is then stated that "this would result in a short term

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3 radioactivity concentration value of about 330 times the allowable continuous occupational concentration value specified in 10 CFR 20. Appendix B." Based on the applicable MPC of shown in sec 'n 2.31 above. the actual ratio is (1.5x10-#)/(3.7x10-7) or 405]

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-3 3.7x10 uCi cm times the applicable concentration limit for the mixture. The inspector's conclusion is an inappropriate application of the MPC values, which are correctly interpreted and applied as concentration guides useful in the design and control of the working environment. They were never intended to be used as a demucation between acceptable and unacceptable at any instant in time, and the ICRP has stated this in their ICRP Publication 2, paragraph 52e, page xxxi, that

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allowed variation of intake with.n a period of 13 weeks provided the total intake was controlled j

to what would be equivalent to 1/4 of an annual intake limit (AIL), i.e.,1/4 AIL, which is the j

QiL in tne extant 10 Lf R 20 regulation.

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Besides this release to the air of 10% being rather large and unrealistic when compared to the recommendation of a maximum inhaled fraction of lx10-6 in the NRC Regulatory l

Guide 8.25, a short term radioactivity concentration need not cause much of an exposure if ventilation or other removal mechanisms cause this concentration to decrease rapidly with time.

j The contamination released from the TIP tubes would not be expected to consist of very fine aerosols that might stay susr.cnded in the air for long times. In fact, it is implied from the I

monitoring results that the contamination was of such physical characteristics that it quickly fell through the grating to the subpile floor below. Air sampling data failed to indicate significant aircorne radioactivity. The workers, who wore face shields that might have provided some protection from such short term concentrations in their immediate vicinity, did not receive any sienificant intake from the contamination released from the TIP tubes. Clearly, this evidence O

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would indicate that this release of contamination had little potential to create an airborne hazard.

It would be more realistic to classify this contamination incident simply as a spill, which had very little if any potential of causing significant airbome concentrations.

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2.33 Exposure Potentialfrom Release into a Ventilated Volume

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To funher show the unrealistic nature of this last NRC scenario, the following model is used to depict the input of an initial activity A(0) to a ventilated volume V where it mixes uniformly, is removed by various pathways as shown and defined below including deposition on walls, the ceiling, and the floor, and is resuspended into the air from inner surfaces defining the air volume V. Although this is a rather simple model, it could in principle be modified to take into consideration more realistic situations. The model also may be used to predict potential exposures in any space whose inner surfaces become contaminated by radionuclides whose resuspension and settling parameters can be estimated. Uniform surface deposition and resuspension is depicted for simplicity here for a single radionuclide. Deposition rates on surfaces and resuspension rates from surfaces are described for simplicity by first order rate constants, which can be converted to more conventional parameters from the surface to volume ratio for the ventilated space:

K A

kA /"

A(0)

V 1'A N

E A

o E

A s

s Parameters shown in the above model are defined below:

A(0) - initial activity input of a single radionuclide into air of volume V.

A - activity suspended in air at time t after initial input.

A - activity on surfaces at time t.

s V - air volume of space, em'.

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3 'I F - ventilation flow rate, em s K - ventilation removal rate constant = FN for uniform mixing, s'I.

y 1 - decay constant of radionuclide, s'I.

K - the sum of all other removal rate constants other than decay or ventilation, including g

diffusion deposition on walls, gravitational settling, etc., s'I k - K + 1 + K = rate constant describing total removal of radionuclide from air, s'I.

y g

K - rate constant describing instantaneous fraction of surface contamination resuspended into the air per unit time, s'I For simplicity in the evaluation of the last NRC scenario, consider the only removal from 3

3 min'I the ventilated volume V of 64 ft to be ventilation at a ventilation flow rate F of 64 ft

~I which yields a ventilation removal rate constant of 1 min or total removal rate constant k of 3 min ~I, which 60 h-1. Reference Man's breathing rate F is 20 L min ~I or (20)/(28.3) ft b

represents a fraction F N r 0.663 h'I of the volume per unit time. This removal pathway is b

neglected compared to the ventilation pathway. The total intake I by an exposed worker who l

remains in the volume until all of the activity A(0) is removed is calculated from the integral of I

the concentration C(t) and breathing rate F 2 b

A(0)

~.

e",

(1)

C (t)

=

V which when multiplied by F and integrated yields the intake I after substituting p;uame:er b

values, including the extremely unrealistic value of A(0) set equal to 10% of the total contamination of11.6 QIL used in the NRC scenario:

-1 (F /V) 0.663 h x

(1.16 O N,

(2)

I = f C (t) F dt

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AM

=

=

b

_l 0

k 60 h which gives:

I = 0.0128 OIL.,

which corresponds to an exposure E:

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O s = co.12e artl<soo arc-nours/ art > = s.4 xec-nours.

Although this value of 6.4 hiPC-hours exceeds the recording reference level of 2 hiPC-hours in a day specified in the extant 10 CFR 20 regulation,it is not a very significant exposure. Even so,it i

is not deemed to be a realistic scenario either considering the fact that the contamin.ation did not 3

remain airborne very long, the actual room volume was 4,869 ft rather than the 64 ft used in l

the NRC scenario, and a much smaller fraction of the activity would be expected to be released.

l With even more unduly conservative assumptions, an exposure exceeding the 40 hiPC-hour control action level might be calculated. If all other methods to limit a calculated potential exposure exceeding 40 hiPC-hours had been exhausted, then it would require respiratory protection under the extant regulation even though it certainly would cause a significant increase i

in the external dose and TEDE received by workers.

l If the actual room volume of 4,869 ft is used instead along with an unrealistic 3

4 conservative ventilation rate constant K of I h-I, which corresponds to a mean residence time in y

the air of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, then an intake of 0.0101 QIL is obtained. This corresponds to an exposure (i 5.05 hiPC-hours, which again is considerable less than the regulatory limit of 500 hiPC-hours l

l and the extant 10 CFR 20 control action level of 40 hiPC-hours. Again, even this trivial Q

exposure is unduly conservative by many orders of magnitude for the reasons stated here and in j

the paragraph abov..

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i 234 Exposure Potentialfrom Release into a Local Ventilation System l

. m-It is obvious that any local ventilation present in the January 27,1993 TIP tube incident j

I was not always effective in preventing release of radioactive material to the general area. In the i

February 11, 1993 TIP tube operations, the NRC inspector expressed concern about contamination and representative air samples for workers. On page 29 of the NRC February 26, 1993 letter to hir. D.51. Smith, this concern was expressed by several observations by the NRC l

1 inspector, including:

a

- Workers periodically placed their heads and upper bodies inside the TIP drive box beveen the source of contamination and a portable ventilation system. (Levels of radioactive contamination in the TIP drive box ranged up to 14 mradthr per 100 cm'. See Figure 5) i O

l 14 i

E 1

t O

Although this observation on February 11,1993 by the NRC inspector is not related directly to the January 27,1993 TIP tube contamination event, it does represent a potential exposure that l

should be evaluated. It is shown in the evaluation below that the maximum fraction of the '

l activity released into the ventilation system that is inhaled is simply the ratio of the worker's breathing rate by the total ventilation flow rate when the worker always breathes from the exhaust stream.

To evaluate tne potential exposure in such situations, the following scenario is proposed. If more realistic parameter values can be shown to apply to this or other similar situations, then the calculation shown below should be done using parameter values applicable to the actual situation, including the fraction of time the worker is actually breathing air from the exhaust stream. It is conservatively assumed that a single worker is continuously exposed in the air t

stream of a portable ventilation system that processes the entire activity A of 2,700 pCi or j

11.6 QIL, which was estimated by the NRC inspector as removable contamination on the subpile j

floor in the January 27,1993 TIP tube contamination event. Although this activity A was not j

associated with the NRC inspector's observation of February 11,1993, this activity is used here to demonstrate the potential exposure that would have been associated with a release of this i

amount of activity. The activity A is assumed to enter and mix uniformly across a 2 ft x 2 ft O

cross section of the TIP drive box, and the air stream velocity perpendicular to this area is l

-1 3 min ~I l

assumed to be 100 ft min This yields a local ventilation flow rate F of 400 ft or 1.13x10 L min ~I It is assumed that the worker continuously breathes at a flow rate Fb I

20 L min ~I from the exhaust of this local ventilation with no respiratory protection until all of the

)

activity has passed through the exhaust system. If it is assumed that the worker's presence ~or l

t breathing does not change the total ventilation flow rate, then the worker's total intake 1 is simply calculated:

4 I=

(F /F) A = (20/1.13x10 ) (11. 6 QIL) = 0.0205 QIL.

b This corresponds to an exposure E:

E=

(0. 0205 QIL) (5 00 MFC-hours /QIL) = 10.3 MPC-hours.

i I

Despite the obvious undue conservatism employed in this scenario, this exposure is very

]

small indeed. The worker is expected to inha!e a fraction F /F or 1.77x10-3 of the total activity b

released into the local ventilation system. This compares to the value of lx10-6 recommended in Regulatory Guide 8.25 for the maximum fraction of loose radioactive material expected to be 15 i

I l

l Q

inhaled by a worker processing such material. Although the calculated exposure exceeds the recording reference level of 10 MPC-hours in a week and would require tracking for the j

remainder of the quarter, it is considerably less than the regulatory limit of 500 MPC-hours as well as the control action level of 40 MPC-hours in a week. Thus, respiratory protection would i

not be required even under this control action level of the extant regulation. The extemal radiation exposure received by the hypothetical worker would no doubt result in a much larger dose than the committed effective dose equivalent from this internal exposure. So that the TEDE l

is maintained ALARA, respiratory protecthn should not be used in such situations.

235 Conclusions ofNRCInspector I

The NRC inspector concludes as shown on page 8 of the NRC report supplement:

f Based on the above review, the inspector concluded that no intakes of airborne radioactive material occurred, but that a significant quantiry of radioactive contamination

'l was unknowingly released into the work area under the Unit 2 reactor vessel. Based on s

the loss of control of this ha:ard and the amount of activity released into the workers' i

immediate airspace, there was a potential for an exposure in excess of regulatory l

O requirements. Ilowever, in reviewing (other) the otherfactors associated with thejob, as described above, a substantialpotentialfor such an exposure did not exist.

Although I agree with many of the concems expressed by the NRC inspector regarding the evaluation and control of internal exposures and although I agree with the final conclusion, I strongly disagree about his perception of the radiological health and regulatory significance of i

i the contamination released in this TIP tube incident. I do not consider a spill of activity that is i

equivalent to 11.6 times the quanerly intake limit for inhalation very significant. It is my belief that there never was a potentialfor exposure in excess of the regulatory requiremer.ts. This has been demonstrated by my more realistic scenados and evaluations shown in other sections above.

Although more realistic than the scenarios used by the NRC inspector, my scenarios are still unduly conservative by several orders of magnitude based upon the maximum intake fraction of c

ixw ~ recommenceu m ant neguiatory uuice o._2. tertauny, respiratory prote.:non snouiu I

not have been employed in this operation. The NRC alleged failure of PECo to use the

" knowledge of the potential hazard from the May 31 event" in planning and disseminating j

information needed to control alleged potential intemal exposure hazards realized in the January

)

O 16 1

i

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27 event was, in fact, not this failure. The failure of PECo was to not appreciate the lack of a f

significant internal radiation exposure potential despite the observed contamination levels associated with the TIP tube work. If PECo were to evaluate the potential for internal exposures employing the unrealistic scenarios of the NRC inspector and if they had used the contamination information from the May 31 event in this way to estimate potential exposures in the January 27 TIP tube work, they would have required respiratory protection. This would have caused workers a significant increase in their external doses and their TEDE. It is my belief that PECo's failure to communicate the alleged significance of contamination experienced in the May 31-event, fortunately, led to ALARA in the January 27 event.

In order to maintain the TEDE of workers in our commercial nuclear plants ALARA, j

respiratory protection need not, indeed, should not be used in most. situations where it is currently being used. Internal exposures can be easily evaluated and controlled provided appropriate monitoring procedures ( e.g., CAMS that alann before workers receive significant exposures and r

i personal air samples (PAS) that provide timely and accurate assessments of their exposures) are l

used along with local ventilation systems or oser engineering controls to prevent releases to the ambient air and that of individual workers. Once workers. management, and NRC inspectors receive adequate training that provides an appreciation of the fact that extemal radiation hazards

}

in nuclear power plants far exceed the intemal radiation hazards, workers will be willing and will be allowed to enter contaminated areas without the use of respiratory protective equipment. This situation must finally be realized for the proper implementation of the requirement in the revised i

10 CFR 20 regulation for maintaining the total effective dose equivalent ALARA.

l l

2 2.36 Evaluation ofPECo's Scenarios Used to Estimate the Potential Exposure of Workers Alicugh ! agrec.uS de 'lan.w 3 and icn $= nc si;n ficant internal :gcF=

potential existed either during the May 31 TIP tube contamination event or during the January 27 event, I do not believe that the scenarios in the 6 cases used by the licensee to evaluate the I

potential internal radiation exposures are realistic. In most cases, despite the fact that some argument can be made that more conservative assumptions could have been employed in some of the scenarios, the scenarios employed by the licensee to evaluate the potential exposure were unduly conservative by many orders of magnitude. This undue conservatism results from, among other things, the assumption of release fractions that are several orders of magnitude larger than values realized in practice. This concerns me when it is realized that different -

i circumstances and even more unduly conservative assumptions might lead to the incorrect

, O 17 1

.,n.-

--r

Q conclusion that respiratory protection should be required in future operations of a similar nature.

For example, the calculation shown for case i leads to an exposure rate estimate of

- 49 MPC-hours per hour of exposure, which exceeds the control action level of 40 MPC-hours in a week. Undue concern and this control action level in the extant regulation might cause the

.l licensee to require respiratory protection, which would cause the workers to receive greater

}

external doses and a TEDE that would fail to meet the ALARA requirements of the revised j

4 pCi cm-3 j

10 CFR 20 regulation. The licensee used a conservative Mo-99 MPC value of 2x10 0 pCicm-3 derived in section 2.31 above for the l

instead of the appropriate value of 3.7x10 3

mixture of radionuclides present. The use of a 1 ft exposure volume is meaningless when it is realized that Reference Man almost breathes at a rate of about I ft min ~I (i.e., (20/28.3) ft3 min ^) and that the total activity released is only (0.28 pCi)(1 QIL/233 pCi) or 1.20x10-3 QIL, f

which corresponds to a maximum total exposure of only 0.601 MPC-hours. Instead of l

estimating exposure rates in units of MPC-hours per hour of exposure, the licensee should have'-

l estimated intakes and total exposures to assess the internal exposure significance of the TIP tube t

contamination event. This comment and recommendation applies to all six cases.

l Despite the problem mentioned in section 2.31 on the units of the conversion factor, the calculations in general were done correctly for the assumptions made. In the attachment to case l

2 7

number 4, a mistake,1,858 cm per ft,in the conversion of ft to cm was found. The correct conversion factor is 929 cm per ft~. This caused a factor of two overestimate in the total activity level, which caused an overestimate of the activity concentration and exposure rate.

1

3. CONCLUSIONS AND RECOMMENDATIONS l

Based upon mv more realistic, still undulv conservative scenarios. as well as guidance provided in the Nuclear Regulatory Commission Regulatory Guide 8.25, Air Sampling in the Workplace, it is concluded:

l

1. Radioactive contamination associated with the TIP tube operations of May 31,1992 and j

january _/, ma at umenex. Nuclear Generaung stauon never posed any sigmficant potential l

for intemal radiation exposures above the regulatory limit, the control action level, the recordinc level, or the monitoring level requirements in the extant 10 CFR 20 regulation.

l O

1 18 1

C

2. The total removable contamination level of 11.6 times the quarterly inhalation intake limit, which is 2.9 times an equivalent annual intake limit (AIL), estimated to be present on the subpile floor and the removable contamination level of 28 yCi, which is 0.12 QIL or 0.03 AIL, estimated

]

.to be present in a TIP tube, are several orders of magnitude below the activity level that would.

j even require air sampling under the guidance provided in Regulatory Guide 8.25, which states.

When quantities handled in any year are less than 10,000 times the AL1, air sampling is l

generally not needed.

It is to be noted that the AIL and ALI have similar radiological health and regulatory i

significance,

3. Use of respiratory protective equipment in work on the TIP tubes would likely cause workers to receive extemal doses and a total effective dose equivalent (TEDE) greater than values that j

would be received if no respiratory protection were to be used. Requiring respiratory protection i

for this work would violate the requirement in the revised regulation to maintain the TEDE l

ALARA.

i O

4. Based upon the above conclusions, the NRC cited violations: (a) failure to make adequate surveys to detemnne that indiwdads were act exposed to airborne concentranons exceeding the limits specified in 10 CFR 20.103, and (b) failure to adequately inform workers as to the presence of high levels of radioactive contamination or of means to minimize their exposure to -

1 such contamination are unfounded by either the facts and the NRC recommendations provided in-jl Regulatory Guide 8.25.

4 Despite these conclusions, it is recommended that greater use be made of personal air samplers (PAS) in job situations like that involving work on the TIP tubes where a worker can j

create aircome radioactivity in his immediate air environment through his own work activities.

This recommendation is being made so that hard data are obtained for convincing workers, management, and NRC inspectors tnat not using respiratory protecnon is otten, in tact, al AnA.

It also is being made because the use of the PAS procedure provides the most sensitive, accurate, and timely method for the evaluation and control of the working environment, which ultimately provides for the protection of workers. The efficacy of PAS devices and bioassay procedures for the evaluation and control ofintemal radiation exposures is shown in Appendix D for one of the O

19

i J

h most difficult to measure transuranic radionuclides, Pu-239. It is shown that PAS devices are f

particularly sensitive and useful in the evaluation of exposures to this tnnsuranic radionuclide.

The sensitivity and accuracy of a PAS device for fission and activation products is even better.

At a sampling rate of only 2 L min ~I the PAS exposure assessment procedure is capable of detecting exposures of less than 0.01 DAC-hours from Co-60. Regulatory Guide 8.25 and NUREG-1400, which provide excellent guidance on air sampling in the workplace, should be consulted. However, I am recommending, for the reason stated above, much greater use for air sampling than can be justified by this excellent guidance. Other specific recommendations for the evaluation and control of exposures to combined internal a.d extemal radiation sources ALARA are summarized:

2 L

Establish a training program for management and staff that includes: (1) the relative magnitude and hazards associated with internal and external radiation sources, (2) the purposes, scope, and requirements for maintaining the TEDE ALARA in the revised 10 CFR 20 regulation, (3) the use of engineering controls and procedures to maintain the TEDE from external and internal sources ALARA, (4) monitoring requirements, including the use of personal air i

samplers and bioassay, and (5) the efficacy for the use of respiratory protective devices.

O 2.

From work histories at Limerick and other nuclear generating stations, document the i

expected external exposure rates and the pete tial irternal exposure rates to detenine the efficacy of using respiratory protection for each major job activity having measurable or j

statistically significant internal exposures.

~

l

?

3.

Based upon an acceptable estimate of the increased (or decreased) work time for each job activity with respiratory protection, calculate the projected extemal, internal committed effective i

dose equivalent, and TEDE expected to result from exposures with and without respiratorv protection. Unless more specific information justifies otherwise. use the maximum intake fraction of lx10~0 referenced in Regulatory Guide 8.25 to convert estimated available loose l

activity to a maximum potential intake. Make a decision to use respiratory protection on the basis of maintaining TEDE ALARA only after all reasonable efforts have been made to first use

)

engineering controls and procedures to limit exposures to both internal and external radiation sources.

0 20 e

m, m

,+r---

w-e

.. ~

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4.

Document the actual work times with and without respiratory protection and the actual internal and external doses for alljob activities having statistically significant intemal exposures.

I Significant internal exposures should be estimated through the use of PAS devices, and filter samples should be analyzed for fission and activation products. For especially plants that have j

had a history of failed fuel, some of the PAS and area filter samples and other samples of l

contamination also should be analyzed for transuranics through gross alpha counting or by other methods such as alpha spectroscopy. Jobs expected to have significant but unknown amounts of f

internal exposure should be monitored by CAMS that have alarm set points established to prevent exposures at 2 DAC-hours if at all possible without causing too many false alarms (See l

Regulatory Guide 8.25.).

I 5.

As needed and justified by regulatory and other considerations, establish routine and operational bioassay programs in conjunction with the air sampling and air monitoring program.

l Establish control action levels from air sampling results, e.g., an estimated exposure of an f

individual worker above 40 DAC-hours, that trigger confirmatory bioassay procedures capable of f

detecting such an exposure. Routine procedures should be chosen that have the frequency, i

sensitivity, and accuracy for confirming that workers have, in fact, not received any significant l

)

exposures to internal radiation sources.

6.

From records of the internal and external exposures received on jobs with and without l

respiratory protection, estimate the savings in the collective TEDE for the population of workers.

l Include this information in the training sessions.

These recommendations are intended to help assure that workers' exposures to combined i

l external and intemal radiation sources are being maintained ALARA as required in the revised

{

10 CFR 20 regulation that becomes effective on Januaq 1,1994. The recommendations when l

j properly implemented also are intended to provide timely detection of significant exposures so i

that corrective actions can be taken to reduce and/or eliminate exposures that are not ALARA.

{

1 To overcome the undue concerns regarding intemal exposures, concerns which, in fact, have j

caused workers to receive larger extemal doses and a TEDE that was not ALARA, the proper implementation of the first recommendation is extremely important.

i 4

O 21

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APPENDIX A COMPARISON OF EXTANT tWD REVISED 10 CFR 20 REQUIREMENTS l

t FOR THE EVALUATION AND CONTROL OFINTERNAL RADIATION EXPOSURES IN CONJUNCTION WITH EXTERNAL RADIATION EXPOSURES i

O AT NUCLEAR POWER PLANTS IN THE UNITED STATES ALARA I

r t.

8 e

O

<l I

l I

Q.-

SUMMARY

OF MAJOR DIFFERENCES l

The requirements for the evaluation and control of external and internal exposures under j

the extant and revised 10 CFR 20 regulations differ in several aspects. A comparison of the requirements for the evaluation and control ofinternal 'udiation exposures shows them to differ in significant ways that affect their practical application in the control of the total effective dose equivalent (TEDE) from combined sources of external and internal radiation as low as reasonably achievable (ALARA). Major differences relating to internal radiation exposure requirements are summarized as follows.

The extant regulation provides quarterly intake limits separate from and independent of the I

external limits specified in a 13 week control period while the revised regulation limits the I

combined doses from internal and extemal sources to certain values from exposures within a'one year control period. Limits within the extant regulation are based upon recommendations in ICRP Publication 2, while the limits in the revised regulation are based upon recommendations in ICRP Publication 26 and ICRP Publication 30. Those portions of the extant and revised 10 CFR 20 regulations that most directly r-late to the evaluation and control of internal radiation exposun:s are quoted in this appendix, which also includes pertinent comments relative to their meaning and practical application.

Comparison of the Limits in the Extant and Revised 10 CFR 20 Regulations The extant regulation limits the intake of radionuclides to certain quarterly intake limits (QILs) independent of any external dose received by workers. The QIL is 1/4 of what would-correspond to an annual intake limit (AIL) for exposure of Standard Man, which is based upon a committed dose to the critical organ corresponding to the maximum permissible dose (MPD) recommended by the ICRP. Each QlL corresponds to an exposure of 500 MPC-hours of Standard Man and not the 520 MPC-hours commonly applied by NRC inspectors and the nuclear industry (See comments below.). The revised regulation limits the total effective dose equivalent (TEDE) to the whole body and dose equivalent to any organ or tissue from combined extemal and internal sources to respectively 5 rems and 50 rems to the actual exposed workers in any year of practice. Application of these revised limits on doses received by the actual exposed workers can lead to practical difficulties that put undue emphasis on the control of individual workers rather than on the working environment where regulatory requirements belong. Considerably flexibility is given for the acceptable methods for demonstrating compnance with these dose 23

f C.

limits in the revised regulation (See comments below.).

Comparison ofRequirementsfor Monitoring and Assessment of Exposures I

The extant regulation requires that exposures to airborne internal radiation sources be monitored, assessed, and recorded when workers enter an airborne radioactivity area and whea their exposure in any day exceeds 2 MPC-hours or in any week 10 MPC-hours. The revised regulation requires monitoring, assessments, and recording of intemal radiation doses when it is likely that the intake in any year codd exceed 10% of the applicable annual limit on intake (A1.1). Experience in the U.S. nuclear power industry shows, in fact, that workers are not likely i

to e:ceed 10% of the ALI. Provided that transuranic radionuclides do not make a significant contricution to the mix of radionuclides, exit portal moni: ors at nuclear power stations are in 3

most cases capable of confirming thb

• ct. Significant exposures of workers above the extant quanerly intake limit have not occurree Jespite the fact that accidental exposures in situations invoking high area surface contamination levels have sometimes been easily detected from the external contamination monitoring of workers, air sampling results, and/or through bioassays.

I Comparison of Control Action. Investigation. and ALARA Requirements O

The extant regulation requires the use of respiratory protective equipment if the projected exposure in any seven consecutive days is expected to exceed 40 MPC-hours and all other means j

have already been taken to limit such an exposure. If an exposure is found after the fact to exceed this control action level of 40 MPC-hours, then an investigation is required to prevent a J

recurrence and to document the circumstances involved, among other things. Respiratory

{

protection is often used in the nuclear power industry to meet the control action level requirement of 40 MPC-hours in the extant regulation. Respiratory protection has been taken to be mandatory, if all other means to limit projected exposures greater than 40 MPC-hours have been exhausted, even if the external dose will be significantly increased thereby causing a larger total effective dose equivalent (TEDE) than that projected without the use of respiratory protection.

Although the revised regulation does not specify similar control action and investigation levels. it.

does require that the total effective dose equivalent (TEDE) from combined exposures be maintained ALARA. Thus,if the use of respiratorf protective equipment or other procedures to limit the internal exposure are shown to cause workers to receive a hrger TEDE than the value anticipated when auch equipment /and or procedures are not used, then such equipment and O

24

- ~ - -

, - - +

n

i l

)

l Q

procedures should not be used. Such use would not be ALARA, and it might be interpreted as a violation of the revised 10 CFR 20 regulation (See comments below.).

a The formal requirement in the revised 10 CFR 20 regulation to maintain the TEDE from combined exposures to internal and external sources ALARA is intended to overcome the ALARA deficiencies in de extant 10 CFR 20 regulation. The proper implementation of this new ALARA requirement v. iP. not be easy. It will require appropnate evaluations of potential i

exposures to internal and external radiation sources. To maintain a proper balance of the requirements for the evaluation and control of external versus internal radiation exposures, training programs for all staff, including management of nuclear facilities will be required in t

overcome the many misconceptions and the undue concern about internal radiation exposures.

The Nuclear Regulatory Commission will have to implement simiar training programs for j

inspectors and other NRC staff responsible for implementing the revised 10 CFR 20 regulation, including regulatory guides on the proper evaluation of potential exposures for helping to assure that the total egcctive dose equivalent (TEDE) is maintained,in fact, ALARA. It is obvious that honest mistakes will be made, and that some of these minakes, no doubt, will be caused by the undue concems and j:rceptions about internal radiation exposures.

j These significant differences in the current and revised regulations influence the

{

detern,ination of the reculatorv sienificance of sudace and airborne contamination levels as well O

as methods regarding the evaluation and control of potential radiation exposures to exterral and internal radiation sources.

i l

I E' '?RPTS AND COMMENTS ON THE APPLICABILITY OF THE

~

l EXTANT AND REVISED 10 CFR 20 REGULATIONS 1.

Requirements for the Evaluation and Control of Internal Exposures in the Extant 10 CFR 20 Regulation.

The extant occupational limit for internal exposures is given in S 20.103 Exposure of

{

individuals to concentrations of radioactive materials in air in restricted areas, which is partially quoted:

(a)(1) No licensee shall possess, use, or transfer licensed material in such a manner as to permit any individual in a restricted area to inhale a quantity of radioactive material in O

1 25 1

0 any period of one calendar quarter greater than the quantity that would result from inhalation for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per weekfor 13 weeks at uniform concentrations of radioactive j

material in air specified in Appendix B, Table 1. Column 1.1.2,3,4.

j i

Footnote 3 states among other things:

Multiply the concentration values specified in Appendix B, Table 1, Column 1, by i

0 6.3x10 mi to obtain the quarterly quantir; timit.

Thus, the quarterly intake limit (QIL) in units of pCi is to be calculated from the MPC value in Table 1 specified in units of yCi mL-I for each radionu' elide:

8 QIL = (6.3x10 mL) MPC.

(1) 1 i

8 9

It is to be noted that the volume of 6.3x10 mL is 1/4 of 2.5x10 mL, the volume of air inhaled l

7 by Standard Man in 1 year of occupational exposure, i.e., (1x10 mL per work day)x(250 work days per year). Thus, equation 1 assumes that a 2 wef. vacation is distributed in equal parts over each quarter so that the quarterly intake limit corresponds to an exposure of 500 MPC-homs and not the 520 MPC-hours that would be inferred fmm paragraph (a)(1) quoted above. When intake l

{

of several radionuclides :akes place, then the mixture rule must be applied to demonstrate compliance, i.e., the intake of each radionuclide relative to its QIL summed over all radionuclides -

must not exceed unity. Similar mixtme rules apply to the revised regulation. It is to be emphasized that this ICRP Publication 2 b

? quarterly intake limit (QIL) given by equation-1 '

applies to Standard Man and not the

.:ul er..wd worker, whose actual intake could be f

different when exposed to the s une concu_

.,t.

Standard Man. It is an intake limit for l

Standard Man and not an exposure or dose limit for the worker. This limit provides proper i

emphasis on the evaluation and control of working environments and a practical way of limiting i

the exposures, intakes, and doses received by workers. When Standard Man is exposed to a 4

concentration of 1 MPC for 2,000 occupational hours in a year, he will have a total exposure E of 2,000 MPC-hours, an intake I of what would be equivalent to an annual intake limit (AIL), and a j

committed dose equivalent to the critical organ equal to the stated marimum permissible dose

. (MPD) recommended by the ICRP for combined exposures from external and internal radiation sources in any given year of practice. The extant regulatory limit is 1/4 of the AIL value, which 9

3 I

can be calculated from the product of the MPC and the volume of 2.5x10 cm of air inhaled by

. O i

26 3

O Standard Man during 1 year of occupational exposure. This ICRP Publication 2 derived AIL is to be distinguished from the ICRP Publication 30 derived annuallimit on intake or ALI value,

{

which is the derived limit applicable to the revised 10 CFR 20 regulation.

I Although the extant regulation does not require a specific ALARA program as in the revised regulation, it does give specific reference levels for recording, control action, and investigation below the regulatory limit stated above as well as other requirements relating to the evaluation and control ofinternal radiation exposures. These requirements relate to an airborne radioactivity area, which is defined in 5 20.203(d)(1)(ii) and quoted:

(d) Airborne radioactivity areas. (1) As used in the regulations in this part " airborne radioactivity area" means (i) any room, enclosure, or operating area in which airborne radioactive materials composed wholly or partly of licensed material, exist in concentrations in excess of the amounts specified in Appendix B, Table I, Column 1 of this part; or (ii) any room, enclosure. or operating area in which airborne radioactive material exists in concentrations which averaged over the number of hours in any week during which individuals are in the area, exceed 25% of the amounts specified in Appendix B.

Table 1, Column 1 of this part.

O The definition given for an airborne radioactivity area means that requirements stated in other paragraphs of 5 20.103 that are conditional on the existence of an airborne radioactivity area need only be implemented if the level of airborne radioactivity exceeds or potentially could l

exceed 25% of the MPC. This conditional nature of the requirements is particularly obvious in l

the part relating to requirements for the recording of exposures, i.e., the recording reference f

levels for internal radiation exposures below which exposures need not be recorded or tracked.

These recording reference levels are given in paragraph (a)(3) of S 20.103, which is partly quoted below. These and other paragraphs in s 20.103 most pertinent to dis report and j

recommendations are summarized and quoted as follows:

I (a)(3) For the purpose of determining compliance with the requirements of'this section the licensee shall use suitable measurements of the concentratim of radioactive materials in air for detecting and evaluating airborne radioactivity in restricted areas and in addition, as appropriate, shall use measurements of radioactivity in the body, measurements of radioactivity excretedfrom the body, or any combinations of such measurements as may be necessary for timely detection and assessment of individual O

27 i

l i

Q intakes of radioactivity by exposed individuals.. When assessment of a particular l

individual's intake of radioactive material is necessary, intakes less than those that would

')

resultfrom inhalationfor 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in any one day or 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> in any one week at umform

}

concentrations specified in Appendix B, Table I, Column 1 need not be included in such 1

assessment, provided thatfor any assessment in excess of these amounts the entire amount.

is included.

(b)(1) The licensee shall, as a precautionary procedure, use process or other engineering controls, to the extent practicable, to limit concentrations of radioactive j

materials in air to levels below those which delimit an airborne radioactivity area as J

definedin 6 20.203(d)(1)(ii).

i (b)(2) When it is impracticable to apply process or other engineering contro!c to limit l

concentrations of radioactive material in air below those defined in S 20.203(d)(1)(ii),

other precautionary precedures, such as increased surveillance, limitation of working times, or provision of respiratory protective equipment, shall be used to maintain intake of l

radioactive material by any individual within any period of seven consecutive days asfar i

below that intake of radioactive material which would result from inhalation of such materialfor 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at the uniform concentrations specified in Appendix B Table 1, i

Column 1 as is reasonably achievable. Whenever the intake of radioactive material by an individual exceeds this 40-hour control measure, the licensee shall make such evaluations l

a and take such actions as are necessary to assure against a recurrence. The licensee shall maintain records of such occurrences, evaluations, and actions taken in a clear and readily idennfiableform suitablefor summary review and evaluation.

(c) When respiratory protective equipment is used to limit the inhalation of airborne radiocc:!ve material pursuant to paragraph (b)(2) of this section, the licensee shall use l

equipment that is certified or had cernfication extended by the National Instinue for

?

Occupational Saferv and Healthafine Saferv and Health Administration (NIOSHIMSHA).

+

The licensee may make allowance for this use of respiratory protective equipment in

~\\

estimating exposures ofindividuals :o this materialprovided that:

(1) The licensee selects respiratory protective equipment that provides a protection factor greater than the multiple by which peak concentrations of airborne radioactive materials

\\

in the working area are expected to exceed the values specified in Appendix B Table 1, O

)

28

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1

l Column 1 of thispart.

The definition of an airborne radioactivity area and the recording reference levels stated in paragraph (a)(3) above mean that monitoring and assessment of internal exposures from inhalation need not be required if the airborne concentration is less than 25% of the applicable value for any mixture of radionuclides. Paragraph (b)(2) specifies a control action level and investigation level corresponding to an exposure of 40 MPC-hours in any seven consecutive j

days. If all other methods for limiting exposures have been implemented, then respiratory protection equipment must be used to prevent this control actior. level exposure of 40 MPC-hours from being exceeded. This must be done even if such use of respiratory protective equipment increases the exposure time so that the total whole body effective dose equivalent from external and internal sources is greater when this equipment is used than when it is not t sed. Paragraph i

(c)(1) allows the application of protection factors in calculating exposures to workers using j

4 protective equipment provided other conditions are met. This includes, as indicated, the choice j

l of equipment that reduces the concentration inhaled to the MPC value. This mus; be done even 1

if the use of such equipment might cause longer exposure times to complete a task and larger l

t total doses to be received.

i O

2.

aeruirements for the evaiuation and Controi of internai exposures in the aevised 10 CFR 20 Regulation Effective January 1,1994.

(

r Similar requirements for measurements as those quoted above for the extant regulation I

apply to the revised regulation, which are given in 5 20.1204 Determination of internal exposure. Somewhat different from the definition in the extant regulation, an airborne radioactivity area in the revised regulation is defined in terms of the ICRP Publication 30 derived internal radiation protection limits:

i Airborne radioactivity area means a room, enclosure, or area in which airborne radioactive materials, composed wholly or partly of licensed materials, exist in concentrations-(1) In excess of the derived air concentrations (DACs) specified in appendix B, to i

5 5 20.1001 - 20.2401, or (2) To such a degree that an individualpresent in the area without respiratory protective equipment could exceed during the hours an individual is present in a week, an intake of i

O l

29 l

'i

Q Column 1 of thispart.

The definition of an airborne radioactivity area and the recording reference levels stated in t

paragraph (a)(3) above mean that monitoring and assessment of internal exposures from inhalation need not be required if the airbome concentration is less than 25% of the applicable value for any mixture of radionuclides. Paragraph (b)(2) specifies a control action level and investigation level corresponding to an exposure of 40 MPC-hours in any seven consecutive days. If all other methods for limiting exposures have been imp'emented, then respiratory protection equipment must be used to prevent this control action level exposure of 40 MPC-hours from being exceeded. This must be done even if such use of respiratory protective equipment increases the exposure time so that the total whole body effective dose equivalent from external and intemal sources is greater when this equipment is used than when it is not used. Paragraph (c)(1) allows the application of protection factors in calculating exposures to workers using protective equipment provided other conditions are met. This includes, as indicated, the choice of equipment that reduces the concentration inhaled to the MPC value. This must be done even if the use of such equipment might cause longer exposure times to complete a task and larger 1

total doses to be received.

O Requirements for the Evaluation and Control of Internal Exposures in the Revised 2.

10 CFR 20 Regulation Effective January 1,1994.

Similar requirements for measurements as those quoted above for the extant regulation apply to the revised regulation, which are given in 6 20.1204 Determination of internal-exposure. Somewhat different from the definition la mc c.uam 4crauvu, an airborre radioactivity area in the revised regulation is defined in terms of the ICRP Publication 30 derived intemal radiation protection limits:

f Airborne radioactivity area means a rcom. enclosure, or area in which airborne k

radioactive materials, composed wholly or partly of licensed materials, exist in concentrations-i (1) In excess of the derived air concentrations (DACs) specified in appendix B, to 5 5 20.1001 - 20.2401, or (2) To such a degree that an individual present in the area witht:ut respiratory protective equipment could exceed during the hours an individual is present in a week, an intake of

+

O 29

-s.---

i O

o.ss otthe an-i umit on intake <Au> or i2 Dsc-hours.

i The condition stated in (2) above corresponds to an airborne concentration that is 30% of the DAC, which compares to 25% of the MPC in the extant regulation.

l The limits in the revised regulation are given in S 20.1201 Occupational dose limits for adults, which are understood to result from combined exposures to external and internal sources in any year of practice:

S 20.1201 Occupational dose limits for adults.

l (a) The licensee shall control the occupational dose to individual adults, except for planned special exposures under S 20.1206, to thefollowing dose limits.

(1) An annuallimit, which is the more limiting of-(i) The total effective dose equivalent being equal to 5 rems (0.05 Sv); or (ii) The sum of the deep-dose equivalent and the committed dose equivalent to any organ or tissue other than the lens of the eyes being equal to 50 rems (0.5 Sv).

(2) The annuallimits to the lens of the eye, to the skin, and to the extremities, which are:

?

(i) An eye dose equivalent of15 rems (0.15 Sv), and (ii) A shallow-dose equivalent of50 rems (0.50 Sv) to the skin or to any extremity.

For the design, evaluation, and control of working environments as well as for the evaluation of internal radiation exposures of individuals, other derived limits are calculated from the dose limits for combined exposures. These include the Annual Limits on Intake (ALIs) for t

inhalation and ingestion and the Derived Air Concentrations (DACs). The DAC is calculated" j

fmm the quotient of the ALI and the volume of 2.4x10 cm, which is the volume of air inhaled f

9 3

by Reference Man in an occupational year, i.e., (20,000 cm min'I)(60 min h-I)(2,000 h y'I).f 3

When the ALI is based upon the nonstochastic dose limit to any organ or tissue specified in (1)(ii) above, the committed dose to Reference Man over the 50 year period following the intake is expected to be the limit of 50 rem for that organ or tissue. When the ALI is based upon the stochastic dose limit specified in (1)(i) above for the total effective dose equivalent (TEDE), the committed effective dose equivalent (CEDE) to Reference Man over the.50 year period _

following the intake is expected to be the limit of 5 rem for the whole body. When the ALI is -

i based upon the nonstochastic limit of 50 rem to any organ or tissue, Table 1 in Appendix B of the revised regulation, also gives the stochastic based intake limit in parentheses below the designation of the critical organ or tissue determining the ALI or nonstochastic based annual 30

~

i l

O iimit oa iataxe.

Conditions requiring monitoring are given in S 20.1502 which is partially quoted:

S 20.1502 Conditions requiring individual monitoring of external and internal occupational dose.

Each licensee shall monitor exposures to radiation and radioactive material at levels sufficient to demonstrate comphance with the occupational dose limits of this part. As a

\\

i l

j minimum-(a) Each licensee shall monitor occupational exposure to radiation and shall supply and require the use ofindividual monitoring devices by-(1) Adults likely to receive, in 1 yearfrom sources external to the body, a dose in excess of10 percent of the limits in S 20.1201(a),..

l (b) Each licensee shall monitor (see 5 20.1204) the occupationalintake of radioactive material by and assess the committed effective dose equivalent to-j (1) Adults likely to receive, in 1 year, an intake in excess of 10 percent of the applicable j

ALi(s) in table 1, Columns 1 and 2, of appendix B to S S 20.1001 - 20.2401; and.

O This section thus provides a monitoring anc recording reference level of 10% of the ALIs.

If it is not likely for workers to receive intakes in a year of occupational exposure that would exceed this 10% of the ALI reference level, then monitoring and assessment of their internal l

radiation exposures is not required. This would correspond to a total exposure of 200 DAC-hours from internal sources in any year of practice when inhalation is deemed to be the' only significant exposure pathway.

A great deal more flexibility for the acceptable methodology for the determination of internal exposure is given in the revised 10 CFR 20 regulation in S 20.1204 than in the extant regulation. Although air monitoring data and bioassay data may be interpreted in terms of intakes, exposures, and doses that would be received by Pcference Man.through the use of the -

derived ALIs and DACs, other methods can be employa nat would provide an estimate of the actualintakes and doses of the exposed workers based upon the characteristics of the radioactive material, its deposition in the respiratory tract, and its behavior in the exposed workers.

Paragraphs giving this additional flexibility are quoted partially:

O 31

(

O 5 20 22o4 oerermiaatioa oriatera=> exposure-(c) When specific information on the physical and biochemical properties of the radionuclides taken into the body or the behavior of the materialin an individualis known, the liceruee may-(1) Use that information to calculate the committed effective dose equivalent, and, if used, the licensee shall document that information in the individual's record; and (2) Upon prior approval of the Commission, adjust the DAC or AU values to reflect the actual physical and chemical characteristics of airborne radioactive material (e.g., aerosol.

j size distribution or densitv); and (3) Separately assess the contribution of fractional intakes of Class D. W, or Y compounds of a given radionuclide (see appendix B to S 6 20.1001 - 20.2401) to the committed effective dose equivalent.

1 (d) If the licensee chooses to assess intakes of Class Y material using the measurements given in S 20.1204(a)(2) or (3), the licensee may delay the recording and reporting of the assessmentsfor periods up 7 months, unless otherwise required by G $ 202202 or20.2203, in order to pennit the licensee to make additionalmeasurements basic to the assessments.

(h)(1) In order to calculate the committed effective dose equivalent, the licensee may assume that the inhalation of one AU, or an exposure of 2,000 DAC-lwurs, results in a committed efective dose equivalent of 5 rems i0.05 Sv)for radionuclides that have their AUs or DACs based on the committed efective dose equivalent.

(2) When the AU (and the associated DAC) is determined by the nonstochastic organ dose limit of 50 rems (0.5 Sv), the intake of radionuclides that would result in a committed efective dose equivalent of5 rems (0.05 Sv) (the stochastic AU) is listed in parentheses in table 1 of appendix B to 55 20.1001 - 20.2401, in this case the licensee may, as a '

simplifying assumption, use the stochastic AUs to determine committed efective dose equivalent. However, if the licensee uses the stochastic ALis, the licensee must also demonstrate that the limit in 5 20.1201(a)(1)(ii) is met.

Althnach I wree that 'he most accurate assessment of internal dose after a known and significant intake takes place should be recorded in an individual's record and although I agree

$= $: men ndinic.'.1.B band : pen p=v= =d deca==ud==nh c Se phy icd =d -

biochemical characteristics of radionuclides should be used to assess the regulatory and internal radiation dose significance, I do not believe that it is either appropriate or practical to provide this 32

O fiexibility in the evaiuation and controi ofinternai radiation exposures. secause the actuai intake and dose depends on the individual characteristics of workers, their actual doses will differ when

[

exposed to the same working environment. It is not practical for a regulatory standard to specify i

l a limit on the assessed dose to individual workers. It is not practical for a regulatory standard to

(

specify acceptable working environments that depend on the characteristics of individual workers. A regulatory standard cannot regulate a worker's breathing pattern and metabolism i

before any exposure takes place. The regulatory standard should specify a minimum acceptable working environment in terms of the exposure of Reference Man using his derived ALIs and l

i DACs (unless other values can be justified by documented research). The evaluation and control of working environments in terms of permissible exposures established for Reference Man is

+

what ultimately protects workers, not the assessment of their doses after an exposure and intake j

takes place.

1 The total efective dose equivalent (TEDE) limit of 5 rems given in (a)(1Xi) is defined in the revised regulation:

(TEDE) means the sum of the deep-dose equivalent (for external exposures) and the i

committed efective dose equivalent (for internal exposures).

The revised regulation requires that the TEDE and other doses be maintained as low as reasonably achievable (ALARA). This ALARA requirement is stated in paragraph (b) under i

s 20.1101 Radiation protection programs:

(b) The licensee shall use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as reasonably achievable {ALARA).

This ALARA requirement means that respiratory protection equipment should not be employed under the revised regulation if it can be demonstrated that the TEDE from combined external and internal exposures would be less without respiratory protection, regardless of the magnitude of.

the projected intemal exposure. This is funher clarified in the revised regulation, which puts a condition for ALARA in S 20,1702 Use of other controls, which is quoted entirely:

i O

1 I

33

1 1

i When it is not practical to apply process or other engineering controls to control the Q

concentration of radioactive material in air to values below those that define an airborne radioactivity area, the licensee shall, consistent with maintaining the total effective dose l

equivalea.t ALARA, increase monitoring and limit intakes by one or more of thefollowing

'j means:.

f (a) Controlofaccess; l

(b) Limitation ofexposure times:

(c) Use of respiratory protection equipment; or (d) Other controls.

t r

a f

Further clarification for maintaining the TEDE of workers ALARA is given under paragraph (b)(1) under 5 20.1703 Use ofindividual respiratory protection equipment, which is quoted partially with added words shown in parenthesis to clarify the intent of this statement:

I If the selection of a respiratory protection device with a protection factor greater than (by i

tnefactor) the peak concentration (exceeds the DAC) is inconsistent with the goal specified i

in S 20.1702 of keeping the total qfective dose equivalent ALARA, the licensee may select respiratory protection equipment with a lower protection factor only if such selection l

would result in keeping the total efective dose equivalent ALARA.

SUMMARY

AND CONCLUSION i

i The control action level of 40 MPC-hours in the extant 10 CFR 20 regulation and undue concern for internal radiation exposures have caused excessive use of respiratory protection in the nuclear power industry that has caused workers to receive a TEDE that was not ALARA.

The revised regulation attempts to overcome this ALARA deficiency by requiring that the TEDE I

from combined internal and extemal exposures be ma'mtained ALARA. The extant regulation properly emphasizes the evaluation and control of the working environment in terms of exposure i

of Standard Man. The revised regulation improperly uses dose limits for the actual exposed workers, which puts undue emphasis on the control of individual workers rather than control of the working environment where the regulatory requirements belong. For practical and other reasons relating to the protection of workers, the regulatory standard should specify a minimum acceptable working environment in terms of exp sure of Reference Man.

..O 34

5 i

-h-

'l

't i

i r

i i

APPENDIX B t

i EVALUATION AND CONTROL OF INTERNAL RADIATION EXPOSURES ON THE BASIS OF COMM17TED DOSE EQUIVALENT O

-s,

'. h I

l i

O

~

35 t

v

y t

?

l EVALUATION AND CONTROL OF INTERNAL RADIATION EXPOSURES l

ON THE BASIS OF COMMITTED DOSE EQUIVALENT 8

hv. Kenneth W. Skrable I

INTRODUCTION

(

1.1 Purposes for the Assessment and Recording ofInternal Radiation Exposures

,he purposes :Or tne assessmenC anc reCerC:ng c:.

nternal rac at10n exposures anc deses are in many ways muCn

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4 1.2 Perceptions of Internal Radiation and Associatei Risks P*

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. 1 i

i its effective half-life, and the doses may be received over many j

years after an actual intake.

When a deposition is confirmed by a bioassay result, the second fact often elicits an emotional response and undue concern by exposed workers, management, and l

l others responsible for the radiation protection program.

A I

worker's concern about a confirmed deposition often will not be l

justified by the magnitude of the risk associated with the dose being received in the current year or the dose committed to future years.

Misconceptions and undue concern about internal radiation and possible litigation have caused:

(1) the use of controls for limiting or preventing internal doses that have i

caused larger external and total doses to be received by workers, (2) an unbalance of the controls on workers and those on the working environment, (3) an unbalance of the monitoring programs for individual workers and those used for the assessment and control of the working environment, and (4) the use of internal dose assessment methods and recording policies that have been inconsistent with the recommended ccmmitted dose limits for exposures of individuals in any year of practice, j

Because of these problems and the fact that item 4 has caused i

what still appears to be an unresolved issue within the radiation l

prctection ccmmunity, at least within the United States, it is Q

important that the purposes for internal radiation dose e

assessments be viewed first in terms of the general concepts of radiation dose and biological effects of concern and then finally i

in terms of how these assessments are used in an overall radiation protection program.

This approach for summarizing the purposes will show that the primary purpose f.:

internal radiation assessments is to assure that workers are receiving adequate protection from internal radiation sources and that this primary purpose may be achieved in many ways.

i 1.3 Assessments ofInternal Radiation Exposures Assessments of the adequacy of the control of a given working' environment can be expressed in terms of (1) measured

.in : :n

.2.=--,

a.q::.=:,

~. :.

2n-

^

(4) committed dose equivalents expected frcm exposures of Reference Man

• to that working environment.

Alternatively, the Reference Man is a person with the characteristics defined in ICRP Publication 23 as modified in ICRP Publication 30 (ICRP, 1975; ICRP, 1977).

O _

assessment of the adequacy of controls can be expressed in terms of either (1) expected annual dose equivalents or (2) expected committed dose equivalents for the actual workers when they are exposed to the given working environment.

It is important to understand the differences and practical difficulties associated' with the various quantities being assessed and controlled.

It is particularly important to appreciate the fact that the term exposure means the subjection of a person to either an

.)

external or an internal radiation field and not the actual' dose received.

Exposures to internal and external radiation fields I

are measured in various units.

For external gamma radiation, exposure is measured in the popular unit of Roentgens or the SI

-1 unit of coulombs kg For internal sources of airborne radicactivity, exposures are measured as a product of the

.2 concentration and exposure time in the SI units of Eq s m or in l

the more practical and popular units of MPC-hours or DAC-hours l

for the demonstration of conformance to respectively the ICRP l

Publication 2 or ICRP Publication 30 recommendations of the International Commission on Radiological Protection (ICRP, 1960 and ICRP, 1977).

For either external or internal radiation

~t exposures, the dose distributicn throughcut the body may vary in a complicated way, and the risk of any resulting stochastic

_ O effects such as cancer may be protracted over many years 1

following the year that the exposures and doses are actually received.

For internal sources, the resulting internal dose and its distribution in the body may vary in a complicated way with time.

The actual dose received may be protracted over many years after the year in which the exposure is received.

i a

DISCUSSION

2. General Concepts of Radiation Dose t

The relationship between the radiation dose received by an organ or tissue and the possible biolcgical effects are very complex.

Variations in the. biological effects have~been notec' I

with the type of radiation; the volume or mass of tissue irradiated; the time distribution of the dose and the influence of biological repair mechanisms; the importance of the irradiated tissue or organ to the overall health of the individual; and'the

age, sex, genetic makeup, and the general health of the o -l

Q irradiated individual.

For internal radiation exposures, uncertainties in the doses and hence biological effects are compounded by other difficulties arising f rom uncertainties in the metabolic models.

For the purpose of establishing practical radiation protection limits and guides, many simplifying

)

assumptions are necessary.

2.1 Committed Dose Equivalent l

7 To account for the total dose equivalent and committed risk" i

associated with the intake of a radionuclide, a committed dose equivalent is used.

The committed dose equivalent to a given target organ or tissue is the expected dose equival'ent averaged throughout that organ or tissue in the 50 year period following the intake of the radionuclide.

For historical perspective and to show the connection of the previous to the current t

recommendations of the International Commission on Radiological Protection (ICRP) and the National Council on Radiation i

Protection and Measurements (NCRP), the relationships between the primary committed dose equivalent limits and derived limits are i

summarized as follows.

The maximum permissible concentration (MPC) of a radionuclide O

erecifiea for e critice1 oraen in 1 car rubiication 2 end in NCaP i

It is the l

Report 22 is a

derived concentratiop limit.

concentration that gives Standard Man during one year of occupational exposure (i.e.,

2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />):

(1) a total exposure of 2,000 MPC-hours, (2) an intake corresponding to what is equivalent to an Annual Intake Limit (AIL) and (3) a committed j

dose equivalent equal to the Maximum Permissible Dose (MPD) specified for the critical organ (ICRP, 1960 and NCRP, 1959).

Recommendations in NCRP Report 22 and ICRP Publication 2 include limits respectively on exposures and intakes consistent with the same committed dose equivalent limits used at that time (NCRP l

Report 22, p. 19, statement after equation 2 and ICRP Publication 2,

p. xxxi, paragraphs (52e) and (52g)).

The acronym AIL is used here for an ICRP Publication 2 derived Annual Intake Limit to Committed risk is the potential or future risk associated with either external or internal radiation exposures.

3 Standard Man is a person with the characteristics defined at the Chalk River Conference in September, 1949 as modified in Tables 6 through 11 in ICRP Publication 2 (ICRP, 1960)

_4_

i i

f distinguish it from the acronym ALI used in ICRP Publication 30 for the - Annual Limit on Intake.

Although numerical values-are j

given for the ALIs in ICRP Publication 30, numerical values actually are not given'in ICRP Publication 2 for the AILS.

The of j

AILS for inhalation intakes can be calculated from the pgoduct g

the reported MPC values and the volume 2.5 x 10 cm of air j

inhaled by Standard Man during one year of occupational exposure.

TheAILsforingestionintakescanbecalculatedfromtheproguct of the reported MPC values for water and the volume 2.75 x 10 mL

+

of water ingested by Standard Man during one year of occupational exposure.

Committed dose equivalent limits are provided in the current ICRP and NCRP recommendations.

They also are calculated over a 50 year period following an intake.

The limits represent the committed dose equivalents that would be received by Reference Man (RM), a person with the characteristics defined in the report of the ICRP Task Group on Ref erence Man (ICRP Publication 23),

when he is exposed at the new Eerived Air Concentration (DAC) limits for 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (ICRP, 1975).

Thus, the DAC is a derived concentration limit that gives Reference Man during one year of occupational exposure: (1) a total exposure of 2,000 DAC-hours, j

(2) an intake corresponding to the Annual Limit on Intake (ALI),

j C

and (3) a committed dose equivalent equal to either (a) the non-stochastic limit of 0.5 Sv (50 rem) for any target' organ or tissue or (b) the stochastic effective dose equivalent limit of 0.05 Sv (5 rem) to the whole body depending'on which of these two limits is more restrictive.

Committed dose equivalent conversion factors for various target organs and tissues, ALI values, and l

DAC values are provided in ICRP Publication.30 and the Federal l

Guidance Report No. 11 of the Environmental Protection Agency l

(EPA) (ICRP, 1977 and EPA, 1988) r 2.2 Annual Dose Equivalent t

Conceptually, an annual dose equivalent, as opposed to a i

committed dose equivalent, is the actual dose equivalent received by various organs and tissues in the body during one year.

This annual dose equivalent includes that resulting from occupational i

. exposures to internal and external scurces in the year under

. consideration as well as internal dose received in this same year

[

from the radionuclide burdens resulting from exposures in i

previous years.

Although the words " actual dose equivalent" are-

.O

-s-

used to distinguish this quantity from the ICRP standard 50 year committed dose equivalent, the actual annual dose equivalent, whether from internal or external sources, is difficult if not impossible to determine exactly.

This is particularly true for the internal component, especially in situations where radioactive progeny born in the body after the intake of a parent contribute significantly to the " actual" dose equivalent.

The actual annual dose equivalent received is also distinct from an annual committed dose equivalent, 'also called previously an annual dose equivalent commitment, which is a committed dose equivalent calculated over a one year time period af ter an intake of radioactive material rather than the standard 50 year period consistently used by the ICRP to calculate committed dose.

This special definition of an annual committed dose equivalent is both unfortunate and confusing.

The use of the i

term committed dose would normally mean the ICRP standard 50 year integrated dose following an intake of radioactive material, and one might falsely conclude that an " annual committed dose" or

" annual dose commitment" might be thought to mean the ICRP standard 50 year integrated dose due to " annual exposures",

i.e.,

from exposures in a given year of practice.

Currently, intake to dose equivalent conversion factors only exist for the full 50 Q

year committed dose equivalent.

Although annual committed dose equivalent conversion factors can in principle be derived from appropriate models, many assumptions would be necessary, and the task would not be easy to compile all of those needed for the demonstration of compliance with annual committed dose equivalent limits should such limits ever be used in a regulatory standard.

It should be noted also that annual committed dose equivalenE conversion factors, should they be calculated, would not be applicable to the calculation of the

" actual annual dose equivalents" resulting from all intakes in a current year or in j

previous years.

This is because an intake may occur at any time during a year, and the dose equivalent during the remainder of the year from that intake is less than the so called " annual dose commitment".

For radionuclides having effective half-lives in the body much less than one year, the entire 50 year committed dose equivalent is essentially received during the year of the exposure.

The committed and annual dose equivalent are then essentially equal to one another.

For radionuclides having effective half-lives in the body much greater than one year, the O

t

-*ww

1 l

O committea aose eaeive1ent =ev he protrectea over menv veers following the year of the exposure.- In this case, the annual dose equivalent in the year of the exposure and every year thereafter is much less than the 50 year committed dose equivalent.

The sum of all the annual dose equivalents expected to be received by Reference Man over the 50 year period after an intake, of course, equals the committed dose equivalent for that intake.

2.3 Organfrissue Dose Equivalent i

f For the intake of some radionuclides, a given organ or tissue 4

may have a specific activity much larger than others during the time the radionuclide is retained in the body.

Examples are the i

radiciodines that concentrate in the thyroid and the actinides I

that concentrate on bone surfaces.

As a result, the dose equivalent delivered to such organs and tissues is much greater than that received by other organs and tissues in the body.

For radionuclides with short effective half-lives, the dose equivalent may be delivered over a short period of time.

In situations involving large accidental exposures, the main biological effect of concern will often be an acute or O

non-stocheetic effect in which the dose eausvetent exceede some threshold value for the impairment of the function of that organ or tissue.

Dose estimates.over time periods of concern for such

effects, as opposed to committed dose estimates, will be important guidance for medical intervention.

-w 2.4 Accumulated Dose Equivalent The accumulated dose equivalent from external and internal sources to a given time or age in a worker's occupational history can refer to either (1) the estimated total dose equivalent actually accrued or (2) the sum of the external and internal

{

committed dose equivalent accumulated at that time or age.

Dose f

equivalents in either case can refer to individual organ or 1

2--.

,,,,4,,,

o,,,, 4,, e.

,., g whole body.

O \\

i 2.s Time Distribution of Dose Equivalent O

The effective half-life of a radionuclide in the body can greatly influence the dose distribution over time.

To demonstrate the general concepts regarding the distribution of f

the internal dose in time and the practicality of the committed and annual dose control

systems, consider an exposure of l

Reference Man under the following extreme situations.

It is l

assumed that Reference Man has a chronic internal exposure over a j

50 year period.

This chronic exposure results in an intake of 1 f

stochastic based annual limit on intake (S-ALI) per year of a radionuclide either having (a) an effective half-life very short j

compared to 1 year or (b) an effective half-life very long compared to 50 years.

See Fig. 2.5 a and Fig. 2.5 b below.

If exposure in case (a) is continued for 50 years, then I

Reference Man will receive 0.05 Sv per year and an accumulated effective dose equivalent UE f 2.5 Sv over the 50 year period with no significant dose being received after the end of that 50 t

year exposure period.

This is deemed to be equivalent to the risk associated with receiving an effective dose equivalent of O.05 Sv per year for 50 years from external sources.

I In case (b), the assumption that the effective half-life is O

much greater than the 50 year exposure period with the additional assumption that the relative activity distribution of the radionuclide in the body does not change with time means that the radionuclide burden and effective dose equivalent rate both increase linearly with time.

The limiting body burden at the end of the 50 year chronic exposure would be what is equivalent to

+

the ICRP Publication 2 Maximum Permissible Body Burden ( MPBB )',"

I which would be causina an effective dose ecuivalent rate of 0.05

~1

~

-1

~

Sv y only after 50 years of exposure at l S-ALI y When exposure in case (b) to the radionuclide with the long j

effective half-life is continued for 50 years, Reference Man will I

have a committed effective dose equivalent of 0.05 Sv per year and will accumulate a committed effective dose equivalent of-

-l 2.5 Sv, which for simplicity is assumed to have the same risk as j

in case (a).

This committed effective dose equivalent of 2.5 Sv comes from the integration of the effective dose equivalent rate resulting from an intake of 1 S-ALI in each year of practice over

]

O.

1

)

0.05 E

Sv y' end of exposure t

t t

i

/

t t

f I

t 0

10 20 30 40 50 60 70 80 90 100 TIME SINCE ONSET OF CONTINUOUS EXPOSURE, years Fig. 2.5 a. Average effective dose equivalent rate since onset of continuous exposure at 1 stochastic base ALI per year of a radienuclide with an effective half-life much less than 1 year.

The H actually received during the 50 year exposure period E

equals the committed H of 2.5 Sv.

g 0.05 l

/'I j

i

/

I (1)

I (3)

/

I O

H, I

/

I l

l

-1

/

Sv y 1

/

i (2)

I

/

(4) i i

/

1 I end of exposure i

t I

f t/

t t

t t

t 0

10 20 30 40 50 60 70 80 90

.100 TIME SINCE ONSET OF CONTINUOUS EXPOSURE, years Fig. 2.5 b. Average effective dose equivalent rate'since enset of continuous exposure at 1 stochastic based ALI per year of a.

radionuclide with an effective half-life much-greater than-50 actual H received during the 50 y 1.25 Sv years.

Area 2

=

=

y exposure period.. Areas 2+3 = 2.5 Sv = committed H from intake y

of 50 S-ALI during 50 year exposure period.

Areas 2+344

=

3.75 Sv = actual H received during 100 year period since onset y

~

1.25 Sv = additional H allowed under an of. exposure.

Area 1

=

y~

annual dose control system.

O

_ g.

t t

a 50 year period.

This 50 year integration following each incremental intake extends in time to the slanted dotted line in Fig. 2.5 b.

The integrated area representing the accumulated committed effective dose equivalent of 2.5 Sv is represented by i

the sum of the areas 2 and 3.

i The actual effective dose equivalent accumulaued over the 50 year exposure, however, is only 1.25 Sv (shown as area 2).

After the end of the 50 year exposure, internal dose will continue to be received at the rate of 0.05 Sv per year.

Over the 75 year period from the onset of exposure, the accumulated H is 2.5 Sv, e

1.25 Sv during the 50 year exposure period and 1.25 Sv in the 25 l

years after the end of exposure.

Beyond this time, the actual H i

100'yearperich received will be greater than 2.5 Sv.

Over the since the onset of the 50 year exposure period, the actual accumulated H.

is 3.75 Sv

the sum of areas 2,

3, and 4).

c Although the risk per Sv may be different when the dose is l

E received at an older age, there is not sufficient scientific evidence to make any distinction between the risks associated with the two exposure cases cited here.

In addition, the implied risks apply only to a population of workers with Ref erence Man i

characteristics and not to exposed workers, whose actual doses and individual risks are likely to be different from one another~

_ O ema trom aetereoce xea-

^1twouan the eccumu1etea a over the too s

year period since the onset of exposure in case (b) is 1.25 Sv creater than that in case (a),

most of the effective dcse equivalent will not likely be received and the risk implied by this additional 1.25 Sv aill nct 1:.kely be expressed during the rrmaining lifetime.

2.6 Implications of Annual and Committed Dose Control Systems Practical implications regarding the hypothetical exposure i

cases discussed in section 2.5 above when (1) an annual dose i

assessment and control procedure is used or (2) when the ICRP l

committed dose assessment and control procedure is used are summarized as follows.

The annual and committed dose limits are both assumed to be equal to the 0.05 Sv effective dose equivalen-I limit for the whole body.

-Mwever, the annual limit applies ~

the actual annual effective cose equivalent H received during y

any year of practice, accounting for the external and internal j

~

dose from exposures in that year and from the current j

burden of radionuclides from exposures in previous years.

The

.O.-

committed dose limit applies to the sum of the external and internal committed effective dose equivalent arising only from exposures and intakes in the year of consideration.

In case (a) involving exposure to a radionuclide with a very short effective half-life, the annual and committed dose control systems require that intakes be controlled within*1 S-ALI per year, yielding the same effective dose equivalent H over the 50 E

year exposure period.

The annual H received in each year is E

equal to the committed H from exposures and intakes in that E

year.

i In case (b) involving exposure to a radionuclide with a very long effective half-life, the actual annual H received'following E

an intake of 1 S-ALI at the beginning of the first year of exposure is only 1/50 of 0.05 Sv or 1 mSv, but the 50 year committed effective dose equivalent H is equal to the ICRP limit E

of 0.05 Sv (i.e.,

0.001 Sv y times 50 years).

Under a strict I

annual dose control system, the worker would be allowed an I

additional 0.049 Sv from external sources during this first year of practice in addition to this intake of 1 S-ALI.

Under the ICRP committed dose control system, the worker would not be l

allowed any further exposure to either internal or external l

sources because he has already received an exposure and Q

associated committed risk corresponding to the limit of 0.05 Sv, which is put in the record in the year of the intake of the 1 S-ALI.

e Under the ICRP committed dose system in which the committed l

dose is controlled and put in the record in the year of the intake, the worker starts each year of practice with the same allowed exposure as any other worker.

For example, if the worke3 l

had an intake 1 S-ALI per year for 25 years as in case (b), then

{

his internal burden would be giving an effective WB dose equivalent rate of 0.025 Sv y

, which would continue the rest of

-1 the worker's life without any further exposure.

In the twenty-sixth year the worker would be allowed another S-ALI or 0.05 Sv from external sources or any combination of external and internal exposures resulting in a total external and internal committed ef fective dose equivalent of 0.05 Sv.

If the worker were to receive only external radiation exposure in the twenty-sixth year corresponding to an effective dose equivalent of 0.05 Sv, his actual total effective dose equivalent in this year would be 0.075 Sv (7.5 rem), which includes 0.025 Sv of internal radiation that was already put in his record as part of O.

f

'~'

i the committed dose equivalent from previous intakes.

Although the hypothetical exposure noted above for the twenty-sixth year would violate the annual dose equivalent limits stated in the current NCRP Report 91 and the annual dose

" limiting values for assessed dose to individual workers" specified in the Environmental Protection Agency (EPA) " Radiation Protection Guidance to Federal Agencies for Occupational Exposure," it would not violate current ICRP recommendations (NCRP, 1987 and EPA, 1987).

Under a. strict annual dose control system, this worker would be allowed to receive only 0. 025 ' Sv (2.5 rem) from external sources during the twenty-sixth year.

i

Thus, workers with different internal burdens would have different allowed exposures under an annual dose control program.

Employers would have to establish different exposure control 5

programs for each worker even if the workers received their burden of radionuclides from exposures under a

different

[

employer.

An annual dose control system emphasizes the control of individual workers rather than the working environment.

Such controls imposed upon workers put undue emphasis and concern for their internal burdens, which will create an atmosphere for s

litigation whether or not they contract some form of cancer.

j It is to be emphasized that the ICRP has consistently

{

Q recommended a committed dose limitation system, recognizing that i

the actual annual dose received in a given year of practice could exceed its numerical limits for combined exposures to internal 4

t and external radiation sources when exposures were being controlled within its recommended limits.

Because internal doses i

can be protracted over many years, the actual doses received in-a i

f given year of practice from combined exposures at the ICRT stated limits can be less than, equal to, or greater than these limits.

j In other words, the ICRP has never recommended a Sacrosanct

[

annual dcse limit ft cc.d ined expcsures to internal and external sources.

The only requirement ever recommended by the ICRP is i

4 that the sum of the external and internal committed dose resulting from exposures in any year of practice be maintained l

belcw its stated limits.

The ICRP uses this ccmmitted dose control system for practical and other reasons.

• nder an annual dcse control system, a worker's allcwed external exposure is limited by his current internal burden of radicnuclides.

Despite this constraint, it is to be ncted that an intake of 1 S-ALI per year for 50 years of a radionuclide-having a long effective half-life also allows this worker vnder j

O

.2 -

i t

i t

i i

i h

an annual dose control system to receive in addition an external i

dose equivalent in each year of practice that totals over the 50 l

year exposure period to 1.25 Sv ( area 1,

Fig. 2.5 b).

This l

additional dose equivalent would not be allowed under the ICRP e

I committed dose control system.

Under a strict annual dose control

system, a young worker not having any burden of j

radionuclides from previous exposures could have a single acute intake of 50 S-ALI of a radionuclide with a long effective j

half-life.

This is because the annual dose equivalent limit l

would not be exceeded.

Yet, this same annual dose control system would not allow this worker any further exposure to either internal or external sources because he would receive an annual

-l dose equivalent equal to the limit the rest of his life from the previous 50 S-ALI intake.

t Except for some industries where chronic exposures, in fact, i

4 do occur, most significant internal exposures result from accidents.

However, the practical considerations concerning the use of the alternative dose evaluation and control programs discussed above would still apply.

The ICRP committed dose e

assessment and control system has the following desirable features for the establishment of a practical internal radiation i

dose assessment and control program:

l O

It provides a simple and practical way of assessing and i

controlling the total committed risk Irom exposures in any year of practice.

i It places the responsibility for any intake,

dose, and i

associated risk under the appropriate employer.

l 1

1 It has minimum impact on the allowed exposures and livelihoed of exposed workers.

l 1

4 It responds to the concerns of workers that internal 3

expc sur es may result in radionuclide burdens that renair

{

with them for the rest of their lives.

j It poses appropriate and technically feasible requirements

}

1 en employers for the monitoring and centrol of the exposures cf worke-f i

O 1.

l

}

v e-

+ - ~. -

m

Y t

f It provides assurance that workers are receiving adequate l

protection ~ f rom internal radiation sources when action, l

investigation, recording, and other reference levels are established with respect to the committed dose limits.

{

It is consistent with the committed dose limits and derived quantities used in the design, evaluation and control of-I working environments.

j It places proper emphasis on the control of the working i

environment rather than individual workers.

l i

i

3. Demonstration of Compliance l

The primary purpose for assessments of internal radiation doses is to assure that workers are being adequately protected I

from internal radiatien sources.

Such assessments also may be required to demonstrate compliance with regulatory limits.

[

Monitoring and assessment procedures should be chosen that have I

sufficient sensitivity and ac; curacy for this primary purpose.

Assessments may require methods involving the monitoring of the working environment, e.g.,

breathing =cne and area air samples,

[

O eaa the = aitori=9 =i iaaivieue1 workers throuan nioassev procedures.

For demonstrating that the working environment is

.{

being adequately con rolled, it is preferable to interpret i

measurements in terms of estimated intakes and exposures of Reference Man.

3.1 Individual Committed Dose Equivalent Limits 1

The ICRP currently recommends that exposures in any year of -

j practice be limited t: values corresponding to certain ccmbined j

external and internal committed dose equivalents that (a) limir 1

the risks of stochastic effects and (b) prevent non-stochastic f

effer s.

j i

3.1.1 Non. stochastic Effects and Dose Equivalent Limits Acute effects in which the dose equivalent is received over-a l

relatively short time interval and in sufficient magnitude to j

impair the function of the irradiated organ or tissue are classified as non-stcchastic effects.

Such effects generally O

! t i

,y--

,y

-,-s-.-#

-m c

require threshold doses in excess of several to tens of Sieverts, with the severity depending upon the total dose equivalent received.

Although the thresholds, because of biological repair, no doubt are influenced by the time distribution of the doses, single limits are recommended for exposures' of workers to combined sources of internal and external radiation in any year I

of practice.

These non-stochastic limits are 0.15 Sv for the lens of the eyes and 0.50 Sv for any other organ or tissue.

j These limits are considered to be sufficiently below the thresholds to prevent such effects.

The internal component of this dose equivalent is a ccmmitted dose equivalent calculated over a 50 year time interval following each intake in the year of

)

practice.

Because of biological repair during the long time interval necessary to receive the full complement of the f

committed dose equivalent,

his non-stochastic limit may be unduly restrictive for a radionuclide having a long effective j

half-life in the body, such as Pu-239, 3.1.2 Stochastic Effects and Dose Equivalent Limit j

i l

The long term effects such as cancer and hereditary disease are those for which the probability of the effect, rather than l C its severity is regarded as a function of dose equivalent without f

i threshold.

Because of the pr:babilistic nature of such effects and their hit or miss nature, su-h effects are called stochastic effects.

They may have latent periocs of several to tens of i

j years after the radiation is received before they are observed j

within an irradiated populaticn.

Whether the dose equivalent i.s l

received from internal or external

sources, the risk c:

i stochastic effects thus may extend over many years after the year i

in which the dose s recei/sd.

To limit the tctal risk cf 1

stochastic effects to an a:ceptable level, an effective dose equivalent limit of 0.05 Sv has been reccmmended by the ICRP for i

occupational exposure of individuals to combined anternal and j

external sources in any year cd practice.

The internal component is a committed effective dose equivalent calculated over a 50 year time interval following each intake in the year of practice.

For the purpose of establ;shing this practical radiation protection limit and other quantities derived frcm this limit, the effective dose equivalent limit of 0.05 Sv, whether from internal or external sources, is deemed to have the same risk as a uniform ;rrad aticn of the who_e body to a dose equivalent of LO.

I 1

1 i

0.05 Sv.

The effective dose equivalent H is to be calculated e

~

from the sum of the weighted dose cquivalents of all significantly irradiated tissues:

g = [T wH

( '1}

H 7 T' i

where:

I T-weighting factor for target organ or tissue T representing w

the ratio of its stochastic risk to the total risk when tre whole body is irradiated uniformly, and l

the dose equivalent to target organ or tissue T and which H

=

T for internal sources is understood to be a

50 year integrated value following each intake.

The NCRP currentlv sugaests a ncminal lifetime somatic cancer

-2

-1 mortality risk of 10 Sv of effective dose equivalent for use in radiation protection for adults (NCRP Report 91, p.

23).

The ICRP combined external and internal committed effective dose equivalent of 0.05 Sv may be considered as a limit on the total committed risk from exposures in any year of practice of about 5 x 10 '.

3.1.3 Application of Committed Dose Equivalent Limits by the ICRP and the NCRP f

In the absence of external radiaticn, exposures above the stated committed dose equivalent limits are considered by the ICRP to be overexposures that warrant ccrrective action whether or not the worker receives an actual dose equivalent in the year of exposure above the stated limits and whether or not the exposed worker is likely to receive the full c o=i t t ed dose equivalent during his remaining lifetime.

Although the NCRP agrees that such exposures warrant corrective actions, the NCRP 1

restricts the application of the committed dose equivalent limits (NCRP Report 84, pp. 13, 38, and 39 and NCEP Report 91, pp. 28, 32, and 48).

The NCRP considers the use of the committed dose equivalent and its associated limits useful for radiation protection planning and for the demonstration of compliance with those

plans, and recommends its use for such purposes.

i However, in the retrospective evaluation of an individual's exposure status, the NCRP recommends that estimates of the actual absorbed dose equivalent be made and ccmpared to its annual

],

..,u.-

m

.w w

]

limits.

The annual dose equivalent. limits in NCRP Report 91 are-numerically equal to the current ICRP combined external and internal committed dose equivalent limits for stochastic and non-stochastic effects summarized in sections 3.1.1 and 3.1.2.

When compared to the ICRP committed dose equivalent limits, the annual limits in NCRP Report 91 allow for much greater exposures and committed dose equivalents from radionuclides having long effective half-lives in the body.

A further discussion of this matter is provided in section 2.5.

Although the NCRP states that a committed effective dose equivalent system should specifically not be used as a measure of an individual worker's exposure status (NCRP Report 84, p.13, l

p.15, and p.33), practical and other reasons favor the u a of the ICRP committed dose limits for both the prospective control of the working environment and tne retrospective evaluation of actual exposures (NCRP,1985).

To use committed dose limits for the design and control of working environments and then annual l

dose limits for the retrospective evaluation of the actual exposures of workers is inconsistent.

Such a practice fails to meet the radiation protection goal of maintaining exposures in any year of practice to an acceptable level of committed risk.

The ICRP confirmed its policy regarding committed dose (ICRP, O

1984):

"At the Stockholm meeting the Commission reviewed those aspects of its policy underlying the use of committed dose equivalent.

The Commission confirms ~that its policy is to limit the risk committed by each year of operation, no credit being taken for earlier years if these have lower committed risks ~5t for future years in the expectation of improved conditions of exposure.

This objective is achieved by che use of annual limits on intake calculated f rom the committed dose equivalent, using a 50 year integrating period.

The commission recognizes that there are practical

[

difficulties in using monitoring results to estimate annual i

intakes of some materials, notably plutonium, but it believes I

that these difficulties can be overcome and that their exictence dces not invalidate the above conclusions."

O - c

3.2 Work Place Limits Control of a working environment within appropriate exposure limits is what protects workers from receiving internal radiation doses in the first place.

A regulatory standard should establish i

minimally acceptable working environments in terms of exposure limits for Reference Man; otherwise, practical problems arise when regulatory limits are expressed in terms of the " actual" doses received by exposed workers.

For example, two workers could be exposed to the same working l

environment that is being controlled within the derived air l

concentrations (DACs) established for Reference Man.

Neither

-l worker has Reference Man's breathing pattern, breathing rate, metabolism, or physical size and weight.

One worker's intake and

a committed dose equivalent could be above the stated limits while j

the other worker's values could be below the stated limits even when concentrations are being controlled within the DACs.

As another example, a radiopharmaceutical facility could hire i

wcrkers who don't have thyroid glands to avoid regulatory violations, to minimize the costs of controls, or in recognition that such workers would be at a lower risk than those having thyroid glands.

Although the last reason for such an employment

{

strategy is laudable, such flexibility in a regulatory standard.

[

can lead to a compromise of radiation protection as well as practical difficulties regarding the livelihood of differen*

i l

workers.

Although the most accurate estimate of a worker's dos.e equivalent should go into his record, it is not practical or even l

possible for a regulatory standard to control the actual dose equivalents received by workers exposed to either external or internal radiation sources.

The actual effective dose equivaients or cm wor,cers c: u w erent s1:ea anc weign;s naving the same exposures to an external radiation field can be quite l

different.

A regulatory standard cannot " regulate" a worker's size, shape, and metabolism.

An acceptable working environment can't be defined separately for each worker depending upon his 7

characteristics or current internal burden of radionuclides.

l This is not practical.

This does not mean that ef. forts should not be expended in characterizing the hazards of the working environment in as much detail as required.

Such details may include the size distribution tf radioactive aerosols, other physical and biochemical characteristics of the radioactive h

i r

__ ~

t s

material to which work.er s are exposed, and its behavior in j

exposed workers.

When justified and properly documented, such 1

information can and should be used to derive ALIs and DACs applicable to the specific working environment.

To assure that facilities are designed and controlled within the basic radiation protection committed dose equivalent limits for stochastic and non-stochastic effects, other quantities are l

derived from these limits.

These then can be applied in the-l design and control of the working environment and in the j

evaluation of exposures of individual workers.

The derived limits assume internal exposure in any year of practice of j

Reference Man (RM) such that an intake and deposition will cause l

committed dose equivalents equal to either (1) the non-stochastic limit of 0.5 Sv to a critical target tissue or organ or (2) the j

stochastic effective dose equivalent limit of 0.05 Sv.

These j

derived limits are summarized as follows.

i 3.2.1 Annual Limits on Intake (ALI)

-f i

[

The more limiting intake derived frcm the stochastic or non-stochastic committed dose equivalent limit in sections 3.1.1 and 3.1.2 above is designated as the Annual Limit on Intake

]

(ALI).

Intake limits are given in ICRP Publication 30 for both inhalation and ingestion of raci.onuclides in various chemical forms.

Because of the anatomical and physiological differences 7

between Reference Man and an exposed worker, the actual 50 year l

integrated committed dose equivalent ultimately received by a worker having an intake of 1 ALI can ce less than or greater than the stated limits.

Any additional cancer risk associated witF the incremental portion of the dose equivalent in excess of the 1

stated limits frcm exposures in any year of practice is not i

considered tr be significant r rpared t: the ;ncertainties ir the actual risk of the exposed worker or the risks the worker receives from other agents.

Because the limits have been t

established well below the threshold for non-stochastic effects, i

an actual 50 year integrated dose equivalent somewhat above the l

statec 21m t wouza not resu.t in any su;n e::ects.

O i

l 3.2.1.1 Stochastic Annual Limit on Intake (S-ALI) and Total Committed Effective Dose Equivalent I-IE

(

Whether or not the ALI is based upon the non-stochastic limit of 0.5 Sv to some critical organ or tissue or the stochastic committed effective dose equivalent limit of 0.05 Sv, the ICRP provides a value of the Stochastic Annual Limit on Intake (S-ALI) in ICRP Publication 30 for all radionuclides and chemical compound forms that are listed.

When intakes of mixtures of j

radionuclides occur, the applicable S-ALI values can be used to estimate the total committed effective dose equivalent H g:

Hg = 0.05 Sv [ (I /S-ALIg),

(3.2) g where:

I.

= the intake estimate for radionuclide i, and i

1 i

S-ALI. - the stochastic based ALI for radionuclide i.

1 I

3.2.1.2 Non-stochastic Annual Limit on Intake (N-ALI) and Total Committed Dose Equivalent II to Target Organ T

[

T r

When intakes of one or more radionuclides might cause a significant committed dose to some critical target organ cr j

O tissue T, then the Ncn-stochastic Annual Limit on Intake (N-ALI) listed for each radionuclide can be used along with the intake EU estirated fer each i radienuclide tc obtain an estimate of the committed dose equivalent H,to this critical target T:

H,=

0.5 SV [ (I:/N-ALIg).

(3.M 2

i g

This equation will provide the total ctmmitted dose equivalent to a critical target T provided that each radionuclide has a N-ALI

• alu listad in IIEP ?ubl.,.ati

  n 20 f:

tha: carger.

If the AL:

for a radionuclide is based upon the S-ALI, a N-ALI value is not given in ICRP Publication 30.

Alternatively, the committed dose conversion factors

<H,,/I>.

in units of Sv per Eq of intake that f

^

are listed in ICRP Publication 30 for all significantly irradiated target organs and tissues T and each radionuclide i can be used to estimate the total committed dose equivalent for a particular target organ or tissue:

r H,,= [ I;.

4

< H.,,/ I >.

(3.4) s O

i

i 3.2.2 Derived Air Concentration (DAC) and Exposure E (DAC-h) j i

The ALI values for the inhalation of radionuclides in the form of radioactive particulates assume an activity median f

aerodynamic diameter (AMAD) of 1 m and a geometric standard deviation of 4.5 or less for the aerosol.

A total of 63% of the inhaled activity of this aerosol is expected to deposit in the three regions of the respiratory tract.

Values are listed for l

various chemical compound forms of the radionuclide, which are

'l

~

designated as Class D, Cisss W, or Class Y and which correspond j

to half-clearance times respectively of 0.5 days, 50 days, and i

500 days from certain compartments within the pulmonary region of the respiratory tract.

The ALI value. for inhalation when divided i

a by the volume of 2,400 m' of air inhaled by RM during one year of l

occupational exposure (calculated from a breathing rate of 3

-1 1.2 m h 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week for 50 weeks per year) yields the Derived Air Concentration (DAC).

This derived quantity has the same application as the previous Maximum Permissible Concentration (MPC) given in ICRP Publication 2 and NCRP Report 22.

The quotient of the airborne concentration C

of a

radionuclide by its DAC value when multiplied by the exposure time t in units cf hours gives the exposure E in units of DAC-h:

l Q

E=

(C/DAC)t.

(3.5) l

[

When RM is exposed to the DAC under occupational exposur:

conditions for 1-year (i.e.,

2,000 occupational hours), he will have an exposure of 2,000 DAC-h and an intake of 1 ALI.

Thi-s intake represents a committed dose equivalent for Reference Man equal to the basic limit, i.e.,

either the 0.5 Sv non-stochastic l

limit in section 3.I.1 or the 0.05 Sv effective dose equivalent i

limit in section 3.1.2.

[

When the intake I of each radionuclide can be estimated, then the total exposure E in DAC-hours can be estimated from the sum j

of the ratios of the intakes and ALIs.

E=2,000DAC-h{(I;/ALI;).

(3.6) r i

Some radionuclides may have DAC values based upon the 0.05 Sv stochastic based limit and others on the 0.5 Sv non-stochastic limit; therefore, the exposure E may not he directly related to l

O

' ?

the committed effective' dose equivalent to the whole body.

4. Assessment of Risk and Management of Overexposed Workers Because of the latent period associated with stochastic effects, the risks from either-erternal or internal radiation doses are committed over many years after the year in which the I

l exposures are actually received.

Ideally, a limit stated j

directly in terms of the risk of stochastic effects would be the l

most desirable and appropriate radiation protection limit.

Practical and other problems make it difficult if not impossible to assess the actual risks associated with exposures of workers to all radionuclides of concern.

Instead, the effective dose equivalent received from exposures to external and internal sources in any year of practice is the quantity used for the l

primary radiation protection limits.

The Internal component is a committed effective dose equivalent, which again is to be calculated for all significantly irradiated organs and tissues over a 50 year integrating period.

Thus, the committed effective dose equivalent is being used as a surrogate of the potential stochastic risk, which is taken to be about 1 in 100 per Sv of effective dose equivalent regardless of its distribution in time.

This simplification and use of committed effective dose Q

equivalent is made for the purpose of establishing practical radiation protection limits that account in this way for the total committed risk frcm exposures ir. any fear of practice.

(

Eecause of the latent perloa asscciated with stocnastic effects, the risks from either external or internal radiation doses are committed cver many years after the year in which the exposures are actually received.

A younc worker who has received an acute intake of 50 ALI cf i

a radionuclide with a long effective half-life has been overexposed even though an annual dose limit may not be exceeded.

[

t In such an exposure, the committed dose and committed risk is 50 times greater than the recommended limit.

Annual dose assessments may help to put this risk in proper perspective with

'=- 7f c x = - =2 rMi ? ti-r ecei-c 4 =- -Me r =tc? liri-4-ci not, however, change the fact that the worker has received a I

serious accidental cverexposure corresponding to the maximum value that culd

'r e all:wed :nly cver an entire occupati:nal lifetime.

If this is not appreciated by responsible persons, f

then apprcpriate corrective actions may not be taken.

These l

i O

' t C

i i

t

L 4

actions should correct the situation in the working environment that led to such a high exposure, and they need not include j

necessarily constraints on the future exposure of the overexposed i

worker.

Although the worker's exposure related lifetime cancer i

t mortality risk (2.5% for Reference Man) in the above example may have been significantly elevated, the degree of its elevation in f

comparison to the lifetime risk from all other non-radiation sources (about 20%)

is still relatively small.

Despite all i

efforts to prevent them, accidents happen, they will continue to

happen, and the fact that they have happened can never be changed.

The worker in this hypothetical accident would not have i

any immediate acute effects and would nct likely have any long term effects such as cancer.

His value as a skilled radiation worker would not be diminished by the accident unless some-artificial annual or lifetime dose limit were to require l

respectively the limitation or termination of his further exposure to radiation.

Even though the worker would have some j

increased chance of cancer, this fact cannot be changed.

Each Sv of offective dose equivalent is deemed to have the same mortality

(

risk as every other Sv, which corresponds to about a 1% chance of cancer death over the remaining lifetime.

The exclusion of this worker from gainful employment as an occupational exposed worker Q

cannot be justified, especially when it is realized that alternative employment could involve greater risks to the worker.

[

Such exclusions only exaggerate the risks associated with accidental overexposures.

With respect-to significant overexposures to either external l

or internal radiation sources, the final decision on the future' exposure of affected workers should be decided on the basis of discussions and agreements made between employers and the

[

overexposed workers.

i

5. Dose Assessments Needed for MedicalIntervention

?

When exposures are being controlled within the basi-ccmmitted dose equivalent limitt then assessments of the l

coinmit t ed dose equivalents can be made on the basis of the I

estimated intakes and intake to dose equivalent ccnversian factors established for Reference Man.

When accidental intakes above 1 ALI occur, then all available information should be considered in the estimation of the com:nitted doses and doses O

. t C

5

- - + -

1

+

I l

1 over-other periods of time to determine the need for medical l

O intervention.

The implementation of this recommendation becomes.

more important the higher the estimated intake.

It normally l

would be considered to be mandatory above an estimated intake of about 5 ALI, which would correspond to a committed effective dose equivalent of 0.25 Sv (25 rem) or a committed dose equivalent of 2.5 Sv (250 rem) to a critical target organ or tissue depending on whether the ALI is based upon the 0.05 Sv stochastic effective dose equivalent limit for the whole body or the 0.5 Sv non-stochastic limit to a critical target.

In the case of overexposures to radionuclides with short ef f ective half-lives in the body, the immediate concern is for non-stochastic effects and whether or not medical intervention might be used to reduce the projected internal radiation dose.

E Annual or dose assessments over even shorter time intervals might l

be helpful in making decisions on medical procedures.

In-the

'I case of overexposures to radionuclides with long effective i

half-lives in which the internal doses are protracted over long periods of time that allow for biological repair, non-stochastic l

effects may not occur even if the projected committed dose equivalents greatly exceed the 0.5 Sv ICRP stated limit.

Annual dose equivalent assessments may be helpful in such cases to Q

demcastrate that they are expected to be belcw the threshold for non-stochastic effects.

A young worker receiving such an exposure may ultimately accrue the entire cc= nit t ed stochastic risk while an older worker nay receive only a portion of this j

risk.

Depending upon the level of the overexposure and the age of the exposed worker, assessment of dose equivalents over time intervals other than the standard 50 year period may be useful for again making a decision on medical intervention for reducing the future dose equivalent and its associated risk.

The purposes noted above for making annual dose equivalent assessments apply only to cases where exposures significantly above the recommended limits may have occurred.

They do not justify broad requirement for annual dose equivalent assessments when exposures are being controlled within the basic l

limits, nor do they necessarily imply that a worker's future i

exposure shculd be controlled belcw an annual dose limit.

J l

1 i

i 4

0 J

. =

i i

l

6. Dose Assessments for Litigation j

It is to be emphasized that assessment of internal radiation doses is an 'after the fact' procedure that generally provides little direct protection of workers.

If a worker has not received the full complement of the committed dose equivalent i

that has been put into the dose record, then at least part of this dose equivalent could not have caused an alleged health effect claimed by the worker.

This is an argument for annual as i

~

opposed to committed dose assessments.

The litigation purpose for annual dose assessments may help an employer win a suit made I

by a worker for an alleged health effect, but it does not protect workers.

If the exposures involved in litigation cases are greatly in excess of the applicable limits, then there may be l

Very little chance for the employer to win such a suit.

Certainly such a purpose does not justify application of a broad requirement for annual-dose equivalent assessments of all workers.

It can be justified only in actual litigation cases or i

perhaps when overexposures are known to have occurred.

If working environments are adequately controlled, then there will r

[

be no need for making annual dose equivalent assessments.

For the purpose of litigation, it is important for an employer to be Q

able to demonstrate that the worker, in fact, was adequately t

protected frcm radiation sources.

This will require the employer to show that an adequate radiation protection program was in place and functioning during the time of the worker's employment.

Adequate records are a key element in litigation cases.

e

7. Radiation Protection Programs i

Depending en the magnitude of external and internal radiation

{

sources an

.o potencla_ f

exp;sures, e emen s :ha; may ;c required for an adequate radiation protection program may include l

all of those in Table 7 below.

l A review and evaluation of the elements listed in Table 7 for the safe and efficient operatien of nuclear facilities will show item 1 influences the implementation and evaluation of every that cther element.

Therefore, it is very i:rportant that the pu moses j

for the primary radiation protection limits in item 1 be clearly stated and f.w.: the lini:s -hemselces le expressed in : e: ms of appropriate and clearly understood quantities.

O.

- +,

w

7 Q-Table 7. Radiation Protection Program Elements i

1.

Adequate primary radiation protection limits.

l t

2.

Qualified staff.

j 3.

Identification of potential external and internal exposure-t pathways and engineering controls to maintain exposures as low as reasonably achievable (ALARA).

4.

Adequate design, construction, operation, and maintenance.

l S.

Methods and procedures to maintain the combined dose from internal and external exposures ALARA.

[

6.

Monitoring and survey programs for the facility and surrounding environment.

7.

Assessments of radiation exposures and doses.

8.

Action, investigation, recording, and other reference levels.

9.

Emergency plans and procedures.

10. Adequate records.

j

11. Quality assurance and audit programs.

The purposes of the radiation protection limits are (1) to prevent non-stochastic effects and (2) to limit the lifetime risk of stochastic effects from exposures to internal and external radiation sources in any year of practice.

It is to be noted f

that the limits under item 1 apply to a one year exposure time Q

interval and that they are to be applied to the sum of the appropriate quantity from combined sources of external and internal radiation sources.

The one year time interval for the f

summation of combined exposures is deemed to meet the stated purposes without putting undue constraints on the operation of nuclear facilities or on the livelihood of exposed workers.

The l

radiation protection limits needed to achieve purposes 1 and 2 are expressed in terms of certain dose equivalents.

The internal compenent of the dose equivalent limit is understood to be a committed dose equivalent to a target organ or tissue or a committeo effective dose equivalent to the whole body calculated over a 50 year period following each incremental l

intake in any year of practice.

In the absence of external j

radiation, the committed dose equivalent limits are interpreted j

in termc :f :thsr ferived ~uincitier needed f:r the derip evaluation, and control of the working environment and the assessment cf the ex;csures of individual w:rkers.

Relati:nships between these derived quantities and the primary limits are I

s =.arized in se; m.. :.2 alcvs.

O 1 I

[

l Provided that exposures are maintained below certain limits, O

they will be below the threshold for.non-stochastic effects and the stated purpose number 1 above will be achieved.

These limits have been established in terms of a dose equivalent limit of 0.15 Sv for the lens of eyes and 0.5 Sv for any other organ or tissue.

To achieve purpose 2, a stocnastic effective dose equivalent limit of 0.05 Sv for the sum of the external and the internal committed effective dose equivalent is used.

This primary radiation protection limit corresponds to a total committed cancer mortality risk from exposures in any year of practice of about 5x10 '.

The use of this limit for both the prospective

~

control and retrospective evaluation of actual exposures provides a consistent approach that minimizes the impact on the livelihood of workers and the requirements imposed on employers.

In addition, the committed dose system provides a simple way of communicating internal radiation risks to workers having chronic or accidental exposures.

It provides a direct response to their concerns regarding the fact that the radioactive material may be retained for the rest of their lives.

With respect to elements 6, 7,

and B in the list of elements for a

radiation protection-program, measurements must be O

• • • =***** i" te=="

=="P^==' *== *=i=^=7

==="i='*d equivalent limits or to other appropriate quantities derived from these limits, including the reference levels listed under element number 8.

If, for example, acticn, investigation, and recording reference levels were to be incorrectly exprassed in terms of say 10% of annual dose limits made squal to t',e CRP committed dose limits, then significant exposures would ic al towed, would not b'e investigated, and would not be recorded.or

4. hose radionuclides having long ef fective half-lives in the i.Ny.

If an internal radiatirn assessnen:

r

sfurs is

'; 1 e t

detect with sufficient confidence a particular refere e level, then another more sensitive and accurate procedure siruld be employed.

For some airborne radionuclides in the

orm of radioactive particulates, routine bioassay procedures ~ay not have sufficient sensitivity and accuracy to detect anc measure exposures of workers known to be receiving chronic and perhaps variable inhalation intakes.

Breathing tone air samples then should he used to monitor their exposures much in the same way that personnel dosimeters are used to estimate doses of workers chronically exposed to external radiaticn (Caldwell, 1972).

O O

CONCLUSION The summation of the external and internal committed effective dose equivalent not yet actually received from s

exposures in a given year of practice provides a simple way of accounting for and controlling the total committed risk and the demonstration of conformance with the current ICRP recommended l

limits.

Although this procedure requires certain simplifying i

2 assumptions, its simplicity and practicality far outweigh any disadvantages.

Its use places proper emphasis on the control of 4

the working environment as opposed to the control of individual i

workers when annual dose limits are used in place of the committed dose limits.

In addition, the use of committed dose limits for both the prospective centrol and retrospective evaluation of actual exposures provides a consistent approach that minimizes the impact on the livelihood of workers and the l

requirements imposed on employers.

i f

?

4 O

i 4

s J

1 l

o j

i O

- :s -

l

REFERENCES

Caldwell, R.,

1972, " Evaluation of Radiation Exposures,"

Health Physics Operational Monitoring, Volume 1,

p.

590 (New York:

Gordon and Breach).

Environmental Protection Agency,

1987,

" Radiation Protection Guidance to Federal Agencies for Occupational Exposure,"

Federal

Register, Volume 52, No.

17,

Tuesday, January 27,

1987, Presidential Documents,

p. 2830.

Environmental Protection Agency,

1988,

" Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion l

Factors for Inhalation, Submersion, and Ingestion," Federal i

Guidance Report No.

11 (Office of Radiation Programs, U.S.

Environmental Protection Agency, Washington, DC 20460).

International Commission on Radiological Protection,

1960,

" Permissible Dose for Internal Radiation, "ICRP Publication 2 (Pergamon Press: New York).

International Commission on Radiological Protection,

1975,

" Reference Man:

Anatomical, Physiological and Metabolic l

Q Characteristics," (Pergamon Press:

New York).

International Commission on Radiological Protection,

1977,

" Limits for Intakes of Radicnuclides by Workers," Parts 1, 2,

and 3 and supplements (Pergamon Press:

New York).

j International Commission on Radiological Protection,

1984,

" Statement from the 1984 Stockholm Meeting of the ICRP," Annals f

of the ICRP 14, No. 2 p.

i (Pergamon Press:

New York).

i National Council on Radiation Protection and Measurements, 1959,

" Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and Water for Occupational

]

j

._2.a_

7...

7 7_;;.

_5 Handbook 69)

(National Council on Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 800, Bethesda, Maryland j

20814-30Cr'-

O 4

National _ Council on Radiation Protection and Measurements, 1985,

" General Concepts for Dosimetry 'of Internally Deposited Radionuclides,"

NCRP Report No.

84 (National Council on-Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 800, Bethesda, Maryland 20814-3095).

National Council on Radiation Protection and Measurements, 1987,

)

" Recommendations on Limits for Exposure to Ionizing Radiation,"

i NCRP Report No. 91 (National Council on Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 800, Bethesda, Maryland 20814-3095) i l

i i

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F h

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4 :.

i APPENDIX C i

i-L t

LIMITS FOR RADIOACTIVE SURFACE CONTAMINATION

'l t

d d,

I i

' G I

t n

'l l.

l l

1

ii i

8 1-l i

i i

i' e

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E' 1

4 4lG f

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36 i

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l

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Reprinted with permission by the U.S. DEPARTMENT OF HEALTH, EDUCATION, AND WELFARE Public Health Service From Handbook of Laboratory Safety The Chemical Rubber Co., Cleveland, Ohio, 1967 Limits for Radioactive Surface Contamination i

11. F. Klein and G.1). Schmid:

This chapter is a compilation of recommended guides for radioactise surface con-I tammation and limits used by U.S. (iosernment Organizations, prnate organi/ations.

and gosernments of other countries, Surface contamination limits given for use in oe-cupational areas and in non-controlled areas are applicable to skm. clothing, equipment and other surfaces; e g., floors and benches.

The control of surface contamination has been a common procedure of radiation control programs since the early dap of the atomic industry. These procedures include both the direct measurement of total surface contamination using portable sursey meters, and smear or wipe tests for sampling the "remosab!e" surface contamination.

Smears hase proven erTective in detecting and measurmg the extent of radioisotope spills and contamination tracked into a clean area. The smear samples are usually counted in sensittse laboratory instruments and proside measure' ment of " removable" contamination, which may be related to the extent of powble mternal exposure (due to mgestion. inhalation, wound contaminanon, and so forth).

Standards for the permissible amounts of radioactive surface contamination have not been established by governmental athority in the United States. How es er, surface contammanon limits base been adopted by most nuclear organizations on an indnidual basis. These limits have not necessarily been derned on the basis of the associated personnel hazards but rather rr.ay be low lesels obtainable without economic hardship.

dp The general philosophy adopted for radiation protection in this country has been to l

limit surface contamination to "the lowest practicable lesel." In some cases, the salues adopted appear to be extremely consersative, howeser, the facility has been able "to the with" these s alues. Surface con: amination limits h.ne also been set at a low lesel because of a desire to hmit (prevent) the contamination of sensitive areas and materials l

(such as in low lesel counting rooms and cross contamination of esperiments).

l Because of the wide range of contammanon lesel3 adopted, and the lack of data relatmg to personne! hazards from surface contaminanon, a resicw has been under taken. From the hierature and a quesuonnaire sent to sarious organizations a compila-tion has been made of the surface contammation limits used in the United States and in other countries. The linuis for use in occupational radiation areas and release to n n-controlled areas (unrestricted or gencial public use) are summarized in Table I and 2 regectis ely.

The contamination categories used in Tables I and 2 in olve an arbitrary usugn.

ment of the actual data w hich meluded hmits for more than 30.spceitic categorie, Er. c generally similar values are used for many of these limits, the categories in Tabic 1 and 3 v.. n

.n, c..,,.,.4 n n n :. a cor o,

. n -. ~. n. - n, a.. ~

Basic Guides-contammation area bench top and floors restricted area actise area tools. equipment and vehicles (a higher direct d y lesel is frequentiv allowed)

Chan.4 reas-unrestricted areas inactive area i

cold laboratory non-process area O

Protective Clorhme - anticontamination clot hing protective shoes respirators (no removable contammanon allowed) 283 i

--_ __m._

on-i

P 2

a-

.m a

284 Radiztion Hazards Material to be Released-scrap equipment vehic!cs for release Skin and Perwnal Clothing-skin af whole hady skin of hands (direct A-y levels higher by factor of 5 frequently allowed) j personal clothes and shoes Under " Type of Facility or Reference" the following classification was made:

Referenced Guides-includes selected limits referenced in the open literature and suggested by individuals or adopted by advisory group or governmental agencies.

Major U. S. Nuclear Centers -an arbitrary assignment was made between the major nuclear facility and the smaller facdity classdied as an isotope user.

U.X twrope l' sert meludes the smaller facilities with a more limited use of radia-tion.

\\

Recommendations of Other Countries-this includes limits adopted or used in-formally by organizations in other countries (generally gosernmental-see ref-crence 13).

The following comments and interpretations of the limits were frequently mad f

the ur.

t l The limits are to be used as guides and in practice professional judgement t

should be used by the radiological physicist to determine the acceptahdity of the actual con tamination.

2. The salues are for the most hatardous isotopes and a relaxation factor of 10 (factor of 10 increase) may be allowed for other isotopes. The most hatardou isotopes (high toxicity) are generally taken as those in Class I of the IAEA s

classification" w hich includes, for instance. *Sr,2 'Ra, *Pu and Am.

2 i

O

3. Wherelimits are gisen only for remosable contamination, the acceptability of i

the tetai ceetaminat en is as,e, sed en the easis ef e ternai exnesure ie nersee-I ne!.

It was obsened that the contamination limits for the same category used by dif-feren: organizations vary by factors of up to 1000. There are basically two different limits used.(1) that recommended by Dunster" and Barnes' and used by most orgamzations, and (2) the " lowest practicable limit.". generally used in the United i

States. The value recommended by Dunster for readily removable contamination is 22.000 dpm/100cm for alpha emitters whde most organizations in the United States 2

use values of 30-300 dpm '100cm for removahh M.

.u..

m wu...an, oray a sea c.ncs a here work in contaminated ar a Suited in a descetable body burden. The article by insenbud Blatz and H e s have re-that significant radium nurdens arry' reported were oh.ened only when the contammation was greater than about 100.000 Becher showed a correlation betwdpm alpha activity per 100 cm; of surface.

5 Schultz and een the amount of surface contammation. the air-borne concentrations, and the urinary excretion of uranium. A contamination level of 40.000 dpm/l00cm; was correlated with the masimum permissible air for uramum TV " '"~

cone mo w

..o -no nao med m a radium con-tammatea residence was reported by Evans.* Radium was process d i until 1941.

n the basement Suneys of the residence in 1964 showed extendte alpha contamination e

lese!s in :he range of 10.000 to 100.000 dpm/100cm; with peak lesels greater th n 1.000,000 dpm/100cm. The results of radium body burden measurements were n 2

a tive with the esception of those ;yrsons who participated m the radium nr negalise fmdinp included indMduals 4 ho lisesi m the nouse as yotmr chddrcn.

oceW ng. The O

h.egae w eug,-

+45ush 46 N ** "

u.

,<-em-se

.-6*

6 4

r 7.3 Limits for Radiozetive Surface Contamination 285 TABl.E I Surface Contamination Guides for Occupational Expowre Alpha idpm /100 cm )

Beta (dem /100 cm )

Type oIFacihty ar keference Total (Fmde Removable Total t t uicdF.t Removable BASIC GUIDES Saenger',

300 30 0.25 mrem /hr troo Dunster*

22.000 220.000 UK Mimstry of Health' 22J100 220,tKK)

M ajor U. S. Nuclear Centers" 300 - 2.000(5) 10 540 f 6) 1.0 mrem /hr (3) 80-2,200 (6)

N D. (2)

_25 mrem /hr (1)

N D. (2)

U.S lsotope Users" XIO 2.000 (4) 10 100 (5) 0.75 mremjhr (1) 200-1.txx)(4)

N D. (!)

0.1 mrem /hr (3)

N.D. (1 )

Eccommendations or 200- 22.000 (7) 22.000 (1) 1.000-220,000 (5) 220,000(1) l Other Countnes" 22.000(4) l i

CLEAN AREAS General D) namics' 7

'Barnes' 0.1 mrem /hr 130 2.200 Dunster 660.000 2.200 71000

UK Mmistry of Health' 2.200 22.000

. Major U. S Nuclear Centers" 50 - 1.000 (8) 4 54 (e)

.05. 3 mrem /hr (7) 50-220(6)

N. D (Il N.D.O) 0 i mrem /hr (5)

N. D. (4 N. D. 18)

(

U. S. hotope U>ers" 2.0 (2) 40(0

\\

N. D. (O N D. (l)

N. D. (1) -

N, D. (1)

Recommendations of 40-2 2% r3) 200-22.000 (3)

Other Couninca" N. D. ( O N D. (O ND (O N. D. (O PROTECTIVE CLOTillNG (SHOES):

Saenger' 150 75 mrem /hr General Dy namics' 500 cpm 1.5 mrem /hr Major U. S Nuclear Centers" 100 5.00(h8) 4 300 (2)

.l.2.5 mrem /hr (10) 80-1.000 (5)

N. D (t)

N. D. (2) 1.0 mrem /hr Q) 1.000 (3 lo S hotope Users"

$00 dpm 100 dpm U

.,~~...t.

w...aso us JUD eu 4 I n,

Other Countries * *

.~i.,

r I

l l

N. D. - no de;ectable Jctivif y

• ts han noe m til sases ' wen enubie to ucactmme #f.: eu,Je appised to a total or hie.1 umiamariatum s iewr et>

the faulnics awume shas tirZ of the t.esal.ssinny is removahic ab, smc ri

+11 han heen a..ameJ that dtrcel.M dows races sarimesis rcporicJ as mr. hr mrad /hr

.and mrep,,hr are equivaicist to mrrm hr

• • The s alue m p.irentheses es the number of 1.icihtses mefuded m the rere of tcporting the specific *alues

% ben reported seminar alues are uwd for f c+rators ewept ihat no removat>le amens a allowed I

~

l

Radiation Hazards 286 L

T AIll.E 2 Surface Contamination Guiden for Release to Non-controlled Areas 2

Alpha idnm/l(Am # ) _

14cta 6dpm, imm f

}

Tmalit ner H emo.absc Total d acJ)*.9 Rcmo. ble i

Tspe af f acihis or Referese

%KINI AND PI-RSON Al. CLOTHING f

i 150 N D.

Or> mrem /hr N.D.

500 ND 05 mrem /hr N.D.

! Saenger' 50 cpm N D.

10 mrem /hr N.D.

ILAst tit 35

f tioneral Dy n.imws' 2.. f:70 66.0t N)

! Barr.cd 2.200 22JNm) f lDunster 2.200 22,0 m l'K Minntry of Health *

(00 1.500 p )

5 (2) Ofr2.0 mrem /hr in)

% tIi; 100 On N D it>)

0 t mrcm/hr 0)

N D (6d Major O $ Nutteat Centers" I

1 N D. (lI l,

100 cpm (2) 200 500 (2)

U.S. Isotope Users **

11.000-22.000 (2) i.100-2.200 (2)

Recommendations of Other Countries" M ATERIAL TO BE RELEASED 300 30

.05 mrem /hr 200 Isacreer' N D.

N D.

N.D.

N D.

LASLthM*

100 cpm 6.7

.I mr m/hr DO General Dynamies' 100 N D.

01 mrem /hr N D.

l

{NBS HB-92" N.D 5.000 ND

!N 7.2" 100-2.000 ( 4 5 500 (2) l-3 mrem /hr On no-t00 (2) l Major U S. Nuclear Centers" N D (1)

N.D. (4)

N.D. til N,D. (3)

}

l

_~,

i

^l l

Ocncrally the N D -no detectabic activity

• lt has not sn allcaes been pouable to determme if a guide apphed to total or fucd wraaminounn.

is remo=4bic tby sme.30 h

lem facibnes suume than 10", of the 10:41 attmt,'It be been uumed that d! rect Ja deses raici sermudy reported as n

fi t

• • Tbc value in paremhesis n the number of fauhocs m6luded in the range or reporung inc speo c *a uca.

to mrem; hr trocased bw a fattor af 5 rcflceling the permiss6hie hm;t, fre.psemit II or the skin of Ifit fie ds the total J t n

cattemH) dose o( TS rem /yt.

s CONCl.l'SION From the limited data sailable it appears that the surface contamination limits in use in the United State > are conservatise and have been etTectne m minimizing radiatm esposure. The data further indicates that surface contammation levek higher than hmits generally used in the Unded States should not present ne l

author G. D. Schmidt.

I REFERENCES The Derivation of M.saemum Per-

1. Dunster. H.J..Contaminatiori of Surfaces by Radeoactive Materials:

missible Lcvck. 4 tomes.( Auguu 1955)

2. Dunster.H J,5urfacc Contummatum Measurements n art Indca of Comrol of Radioactive Maters

(

/truith l'As su t. I'.4 4. pp 3H W. (19e621 m

w6 e

e

- am p

m s.

4

f 73 i

1.imits for Radioactive Surface Contaminati 3 Barnes. D. E., " Basic Crneria m the Control of A 1

287 O(d clear Instattations; Report of Sympouum Organaedn and Surface Contamination i

(May 1950).

Eaenbud. M.. Blatz. H. and Berry, E. V., How Irnpo tat the Danish Atomic 4

u-12, No. 8, (1954L

3. Schultz, N. B., and Becher, A. F.. -Correlation of Ur ant is Surface Contamina j

Concentrations. and Urmary Escretann Rates '

Ifralsh Phrsics. Vol. 9.pp 90) 409 (1963tramum

6. Evans. Robley D., Long. term EKects of Resid duum and Mesorkortum Possonmg and Donmetry a d ience m a Highly Contammat pp.15-24. (May 1964).

n Ra-

7. Saenger. E. L., Medical Aspects of Radiati n A nstrumentarwn Techmques an 4ppford Radwacti

vas, industroal li n gsennis, ( February 5 % 31 Unsted Knngdom Msnestry of Health. Code of Pra thccidenb. flandhun o

8 entsund Heahh Physica. General th nam <cs I/andboo&Radsatwns artsangf i

9 6.

agamst lanzung to. LASL-1835. General Handhovi for Radion ber 1958) em Munaaring. U.5 Atorruc En 1l. % ?.2. Radianon Proarction in Auclear React cryy Commmun. (Novem-(July 19636

12. Handbook 92. Safe #andhnt of Radmarnw Mor Fuelfabe ssociation.

cau of Standards.(March 1964).

13. International Atomic Energy Agency No 1 S farrrials. U.S. Department o ur-

.. a e Handlang of RadwesoioresJ1962).

9 t

h 9

O

_m g

ee

i 1

)

i OL l

I I

i

'I APPENDIX D t

EXAMPLE CALCULATIONS FOR THE EVALUATION OF Tile EFFICACY OF INTERNAL RADIATION EXPOSURE ASSESSMENT PROCEDURES i

O i

I i

i r

i

'{

s i

i

?

-i O

i 37 T

o C

EXAMPLE CALCULATIONS FOR TIIE EVALUATION OF TIIE EFFICACY OF INTERNAL ASSESSMENT PROCEDURES INTRODUCTION Timely detection of exposures by personal air samplers (PAS) or by other types of breathing zone air (BZA) monitoring systems can be used to initiate and improve the sensitivity of any follow-up bioassay procedures. Determinations of the radiological, physical, and biochemical properties of radioactivity collected on filter samples representative of intake by workers can improve the estimates of intakes and doses obtained from follow-up bicassay t

procedures. For example, measurements of the activity distribution of radionuclides on representative air filter samples along with measurements of a gamma emitting radionuclide easi!" detected by chest counts can be used to estimate the intake and activity of another radionuclide, such as Pu-239, not detectable by chest counting with sufficient sensitivity and accuracy. The efficacy of various procedures for the assessment of exposures to airborne concentrations of Pu-239 is evaluated in terms of their physically significant levels (PSLs) and O

minimum detectable amounts (MDAs). Values for radiation protection quantities for each assessment procedure are obtained by dividing each of the statistical quantities by the intake retention fraction (IRF), i.e, the expected measurement for the procedure relative to a unit intake by Reference Man.

f EXAMPLE CALCULATIONS OF THE EFFICACY OF PERSONAL AIR SAMPLING (PAS)

AND BIOASSAY PROCEDURES FOR DETECTING EXPOSURES TO Pu-239

1. Assessments ofInternal Exposures to Pu-239 by a Personal Air Sampling Procedure For a personal air filter sample having a 100% retention, the expected activity on the filter re:me :o Reference Ma's incke, ;.e. de ;at.& retentionfaction er IRF v&e ::ppli=ble :c 2e

)

filter sample,is simply the sampling flow rate F divided by Reference Man's breathing rate 3

F i20 L min For example, an IRF value of 0.1 would apply to a PAS filter for a RM sam 9 as io-rate or 2 ' mia ' and filter retention of 100%. The following examples using this O

ti r

IRF value of 0.1 illustrate the calculations required for the assessment of exposures to Class W l

i O

and Class Y Pu-239 using the following PAS sampling procedure. For Class W Pu-239, the stochastic based annual limit on intake (S-ALI) is 400 B'q (24,000 dpm), and the non-stochastic based annual limit on intake (N-ALI) is 200 Bq (12,000 dpm) based upon the 50 rem committed l

dose equivalent limit to bone surfaces, which determines the DAC of 0.08 Bq m-3

. For Class Y Pu-239, the stochastic based annual limit on intake (S-ALI) is 600 Bq (36,000 dpm), and the i

non-stochastic based annual limit on intake (N-ALI) is 500 Bq (30,000 dpm) based upon the 50

-3 rem committed dose equivalent limit to bone surfaces, which determines the DAC of 0.2 Bq m,

Alpha counting of the PAS filter sample on a gas flow proportional counter having an alpha t

particle and gross detection efficiency E of 0.3 c d for l'a-239 is used to estimate the activity on f 60 minutes. The the filter. The PAS filter is counted over a sample counting interval Ts4 background counting rate R is determined for a blank filter to be 0.05 cpm determined over a b

counting interval T f 60 minutes.

b The enclosed Figure 1 shows the relationship between the critical level L r decision tool e

used to decide when statistically significant activity is present on a PAS filter sample and the lower limit of detection L r that net counting rate having a probability (1-#) of being detected d

above L. The enclosed Figure 2 shows the relationship between the primary and derived c

internal radiation protection limits.

The decision tool or critical level L in net epm and physically signi6 cant level PSL in c

O dpm are calculated for a false positive probability a of 0.05 ( normal deviate k = 1.645). Thus,1 out of 20 blank samples would be expected to yield a net counting rate above the critical level or decision tool L when no activity is actually present in these samples. The minimum detectable c

activity MDA, which corresponds to a net counting rate equal to the lower limit of detection L '

d is that PAS filter activity that would have to be present to have a probability (1-4) of 0.95 of obtaining a net counting rate above the decision tool L. Thus, the MDA derived quantities have c

a 95% chance of being detected and 5% chance of not being detected. The critica!!evel L '

c physically significant level PSL, lower limit of detection L, minimum detectable activity MDA, d

and quantities derived from the PSL and MDA are calculated assuming Poisson counting statistics and a normal distribution for the net counting rates.1.1 Class W Pu-239: ALI = 200 Bq: DAC = 0.08 Bq m*E:S-All = 400 Bq:1RF = 0.lfor BZA:

The critical level L or net counting rate having a 5% chance of being exceeded when no activity e

is present is calculated:

3 R

b b

) 0 ~ ~A k (

L

=ko

+

=

C 0

m T

• s+b b

Q 2

CRlTICAL LEVEL L, AND LOWER I2MIT OF DETECTION Lg OFA COUNTINGPROCEDURE t

prob. = a = p = 0.05, f(R )

when k = 1.645.

3 a = prob. of R

1. b when <R > =s 0. c' s

1. = prob. of R, b

• c'

<L when <R >

• s d

a O

L b

c d

s NET COUNTING RATE =

R

=Rs+b ~ Rb, where:

s R

= NET COUNTING RATE, cpm, s

Rs+b = GROSS COUNTING RATE OF SAMPLE, cpm, and R3 = EACKGROUND COUNTING RATE, cpm.

R3 R

II#s+b)2 b)2))1/2 g

b

)1/2-where:

I#

o

(

=

+

s T

T s+b b

s.= THEORETICAL POISSON PROPAGATED ERROR IN R, cpm, o

s Ts+b = SAMPLE COUNTING INTERVAL, minutes, and T3 = BACKGROUND COUNTING INTERVAL, minutes.

Ab b

)1/~7 L

=0+ko

=k (

+

c g

T T

s+b 3

k Ld=ko

+ko b

g d*

c' Tg

(

i Figure L Relationship Between Critical level L and Lower Limit of Detection L

• e d

3

l ICRP PUBl2 CATION 30 RELATIONSHIPS FOR REFERENCE MAN EXPOSURE TO CONCENTRATION OF 1 DAC FOR j

2,000 HOURS 4

EXPOSURE OF 2,000 DAC-h 1

INTAKE OF 1 ALI AND 4

IF ALI IS SALI IF ALI IS NALI (50y)

Hy(50y)

Hm*OF

~OF OR 0.05 Sv 0.5 Sv (5 rem)

(50 rem)

WHOLE BODY CRITICAL TARGET T O

SALI NALI EXAMPLE: A PAS filter sample indicates intakes I of 400 Bq of Co-60, which has a S ALI ALI of IE6 Bq. and 10 Bq of Pu-239, which has a NALI of 500 Bq and a S ALI of 600 Bq.

C I 4 E=

{[

) ( 2,000 DAC-h), or

^

i ALI 400 Bq' 10 Bq

) (2,000 DAC-h) = 40 DAC-h.

E=

(

+

1E6 Bq 500 Bq i

H, = ([

) ( 5,000 mrem), or i

SALI 400 ?;

lo ?..

) (5,000 mrem) = 90 mrem.

.l E=

(

+

1E6 Bq 600 Bq Figure 2. Relationships Between Primary and Derived Limits for Internal Radiation Protection.

O 4

u

i 0.05 0.05 5

O 0.0672 com.

t

=xe

= 1.645 (

+

)

=

c 60 60' The physically significant level (PSL) in activity units of dpm is obtained from the quotient of Lc by the counting efficiency E:

L 0.0672 cpm c

0.224 dpm.

PSL =

=

=

_1 E

0.3 e d The physically significant intake (PSI) is calculated for the IRF value c PAS filter:

PSL 0.224 dpm y

= 2.24 dpm ( 1.87x10 ALI).

PSI

=

=

IRF 0.1 The physically significant exposure (PSE) is calculated:

2,000 DAC-h 2,000 DAC-h 0.373 DAC-h.

PSE = PSI (

) = 2.24 dpm (

)

=

ALI 12,000 dpm O

The physically significant committed effective dose equivalent (PSH ) is calculated:

E 5,000 mrem 5,000 mrem 0.467 mrem, PSH, = PSI (

) = 2.24 dpm (

)

=

S-ALI 24,000 dpm which has a 5% chance of being exceeded when no exposure actually occuned.

The lower limit of detection L r net counting rate having a 95% chance of being detected d

above the L is calculated:

c k~

(1.645)~

0.179 cpm.

(0.0672 cpm) 4=

L

+

=

d T

60 min

~

g3 The minimum detectable amount (MDA) or PAS filter activity having a 95% chance of being detected above the L,,is calculated:

L 0.179 cpm d

0.598 dpm.

MDA =

=

=

E 0.3 c d

• O 5

i Q

The minimum detectable intake (MDI) having a 95% chance of being detected is calculated for an intake retention fraction IRF of 0.1 applicable to the BZA filter:

l i

MDA 0.598 dpm

_4 5.98 dpm ( 4.98x10 ALI).

MDI =

=

=

-IRF 0.1 1

The minimum det~mble exposure (MDE) having a 95% chance of being detected above the L l

c is calculated:

i 4

2,000 DAC-h 2,000 DAC-h 0.997 DAC-h.

5.98 dpm (

)

MDE = MDI (

)

=

=

ALI 12,000 dpm The minimum detectable committed effective dose equivalent (MDH ) having a 95% chance of

[

E being detected is calculated.

5,000 mrem 5,000 mrem 1.25 mrem, 5.98 dpm (

)

MDH

= MDI (

)

=

=

S-ALI 24,000 dpm i

which also is the committed effective dose equivalent that is expected to go undetected 5% of the i

time when activity is present on the PAS filter at the MDA.

f For an exposure time T of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (480 min), the minimum detectable concemration j

(MDC) having a 95% chance of being detected above the L is calculated:

c MDA MDA 0.598 dpm

_4

_1 -

j

6.23x10 dpm L MDC

=

=

V TF (480 min) (2 L min _1) s

-3 MDC = 2.81x10

' yCi cm '

1.04x10

• Eq m

= 0.130 DAC.

-l'

~

=

l The MDC also can be calculated:

5 MDI MDA/IRF (0 598 dpm) / (0.1)

_4

= 6.23x10 dpm L ^.

[

MDC =

=

=

TF TF (480 min) (20 L min _y)

RM RM This altemative expression may be used to calculate the MDC of a bionssay procedure.

1.2 Class Y Pu-239: ALI = 500 Bq: DAC = 0.2 Bq m' ; S-All = 600 Bq:IRF = 0.1for BZA:

l Because of its higher limits, the derived quantities for Class Y Pu-239 are even lower than the values calculated for Class W Pu-239 in section 1.1 above:

6 l

y The physically significant-intake (PSI) is calculated for-the IRF value of 0.1 applicable to the PAS filter as for Class W Pu-239:

l PSL 0.224 dpm

-5

= 2.24 dpm ( 7.47x10 ALI).

PSI =

=

IRF 0.1 The physically significant exposure (PSE) is calculated:

2,000 DAC-h 2,000 DAC-h 0.149 DAC-h.

PSE = PSI (

) = 2.24 dpm (

)

=

ALI 30,000 dpm The physically significant committed effective dose equivalent (PSH ) is calculated:

E 5,000 mrem 5,000 mrem 0.311 mrem, PSH,= PSI (

) = 2.24 dpm (

)

=

S-ALI 36,000 dpm which has a 5% chance of being exceeded when no exposure actually occurred.

The lower limit of detection L nd minimum detectable activity MDA have the same values as d

those calculated for Class W Pu-239. but the derived quantities are different occause of the higherlimits for Class Y Pu-239:

The minimum detectable intake (MDI) having a 95% chance of being detected is calculated for an intake retention fraction IRF of 0.1 applicable to the PAS filter as before for Class W Pu-239:

MDA 0.598 dpm

_4

= 5.98 dpm ( 1.99x10 ALI).

MDI =

=

IRF 0.1 The minimum detectable exposure (MDE) having a 95% chance of being detected above the Le is calculated:

2,000 DAC-h 2,000.DAC-h MDE = MDI (

) = 5.98 dpm (

)= 0.399 DAC-h.

ALI 30,000 dpm The minimum detectable committed effective dose equivalent (MDH ) having a 95% chance of E

being detected above the L is calculated:

c 5,000 mrem 5,000 mrem 0.831 mrem, Q.

MDI (

) = 5.98 dpm (

)

MDH

=

=

p S-ALI 36,000 dpm

~

7 i

i

r O

which also is the committed effective dose equivalent that is expected to go undetected 5% of the time when activity is present on the PAS filter at the MDA.

For an exposure time T of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (480 min), the minimum detectable concentration (MDC) having a 95% chance of being detected above the L is calculated as before:

c MDA MDA 0.598 dpm

-4

-1

6.23x10 dpm L MDC

=

=

V TF (480 min) (2 L min _y) s

-13

-2

-3 MDC = 2.81x10 yCi cm-

= 1.04x10 Eq m

= 0.052 DAC.

Conclusionfor the Eficacy of a PAS Internal Exposure Assessment Procedure:

l The PSL derived quantities consisting of the physically significant intake, exposure, and committed effective dose equivalent demonstrate that a breathing zone air sample has all the sensitivity needed to avoid false positive results that wculd warrant concern. Likewise, the MD 4 derived quantities consisting of the minimum detectable intake, exposure, and committed f

effective dose equivalent provide very high assurance that workers are receiving adequate protection from exposures involving inhalation intakes of Pu-239. The possibility of frequent monitoring and screening of filter samples provides for timely detection and accurate assessments of exposures.

EXAh!PLE C.ALCULATIONS OF THE EFFICACY OF BIOASSAY PROCEDURES FOR DETECTING EXPOSURES TO Pu-239 l

t A bioassay sampling procedure may be considered as an air sample of the breathmg 7one of a worker when inhalation is the primary exposure pathway. The efficacy of bicassay j

procedures for Pu-239 in meeting radiation protection requirements is summarized below for annual monitoring by chest counting, trinalysis, and fecal sampling. Obviously annual monitoring does not provide for timely detection of those exposures that might warrant corrective j

actions or protection of plutonium workers by initiation of chelation procedures. When such procedures are used as routine bicassay procedures, they should have the sensitivity and accuracy

[

necessary to demonstrate that workers have, in fact, received adequate protection over the past year from all exposures to plutonium. As shown by the calculations below, none of the O

procedures summarized below, except perhaps annual fecal monitoring for Class Y Pu 239, meet s

. this radiation protection requirement.

8 l

\\

O

1. Assessments ofInternal Exposures to Pu-239 by Annital Chest Counting:

The acceptable minimum detectable activity (MDA) of Pu-239 by chest counting is given as 110 nCi (4,070 Bq) in the Intemal Dosimetry Technical Basis Manual for the Savannah River

-l Site (report number WSRC-IM-90-139 dated 12/20/90). For the standard mixture of plutonium l

isotopes, the MDA varies from about 1/2 to 3/4 of this MDA value or perhaps as low as 55 nCi.

(2,035 Bq). This value is used below to calculate the MDI, MDE, MDH, and MDC for Class E

W compoands and Class Y compounds of Pu-239 from an annual chess count using the indicated IRF values shown. The intake retention fraction (IRF) values were calculated for the respective

~

1 compound classes using our INDOS computer code and intake retention functions for the l

Jungs. Assumptions used in the calculation include a 1 micrometer AMAD aerosol, the l

respiratt ry tract model in ICRP Publication 30, and an annual frequency of monitoring, i.e., the l

I IRF values were calculated at a measurement time 365 days after a single acute intake.

1.1 Class WPu-239: MDA = 2.035 Bq andIRF = 1.36x10' for the lungs:

j Other applicable parameter values used in the calculations are:

l ALI = N-ALI = 200 Bq, and DAC = 0.08 Bq m-

.O S-ALI = 400 Bq:

-1

and g = 20 L min F

I T = 8 h = 480 min = exposure interval for calculating the MDC.

t The minimum detectable intake (MDI) is calculated for the IRF value of 1.36x10-3 applicable to I

the annual chest count:

i MDA 2,035 Bq i

6

(

= 1.50x10 Bq (7,480 ALI).

MDI =

=

IRF 0.00136 i

I The minimum detectable exposure (MDE) havine a 95% chance of beine detected is calculated I

i tor a enest coum.

l t

}

1 INDOS,Inte:nalDosimetry Computer Programs,1986, Skrable En erpris es,

)

6 Ruthellen Road, Chelmsford, MA 01824.

O sxrau1e, et ai.1988. intate Retention Functions ans rheir Apviications to sioassay ans i

the Estimation ofInternal Radiation Doses, Health Physics 55, 933 - 950.

.\\

l i

' O 2,000 DAC-h 2.000 DAC-h 6

1.5x10, DAC-h.

MDE = MDI(

) = 1.5x10 Bq(

)

=

.ALI 200 Bq j

The minimum detectable committed effective dose equivalent (MDH ) having a 95% chance of E

being detected is calculated:

}

5 rem 5 rem i

0

.x0 Bq (

) = 19,000 rem, l

MDH

=

E=

S-ALI 400 Bq which also is the committed effective dose equivalent expected to be missed 5% of the time.

j r

I For an exposure time T of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (480 min), the minimum detectable concentration (MDC) having a 95% chance of being detected is calculated:

6 MDI 1.5x10 Bq 5

4 1.56x10 Bq m MDC =

=

=

TF (480 min) (0.02 m min ~1) 3 I

g which is 2.0x10 times the DAC of 0.08 Bq m-3 for Class W Pu-239.

l 6

1.2 ciass y eu 239: uva = 2.oss ar. ans zar = o.io4 tor zac iungs:

i O

Ccher applicable parameter values used in the calculations are:

l A*_I = N-ALI = 500 ?q, and DAC = 0.2 Eq m ";

r

~1 S-ALI = 600 Bq;.Fg = 20 L min

and T = 8 h = 480 min = exposure interval for calculating the MDC.

f t

The minimum detectable intake (MDI)is calculated for the IRF value of 0.104 applicable to the j

annual chest count:

MDA 2,035 Bq 4

= 2.0x10 Bq ( 39 ALI).

MDI =

=

IRF 0.104 l

l The minimum detectable exposure (MDE) having a 95% chance of being detected is calculated j

for a chest count:

l 2,000 DAC-h 2,000 DAC-h i

4 4

8.0x10 DAC-h.

j MDE = MDI(

) = 2.0x10 Bq(

)

=

ALI 500 Bq 10

]

4

,y

-m

t

.l l

i The minimum detectable committed effective dose equivalent (MDH ) having a 95% chance of E

being detected is calculated:

l 5 rem 5 rem I

.0x1d Bq (

) = 1% mm, j

MDHE=

(

)

=

S-ALI 600 Bq 1

which also is the committed effective dose equivalent expected to be missed 5% of the time.

i i

For an exposure time T of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (480 min), the minimum detectable concentration (MDC)

)

having a 95% chance of being detected is calculated:

4 MDI 2.0x10 Bq 2.1x10 Bq m

-l MDC =

=

=

TF (480 min) (0.02 m min ~1) j d

g 4

-3 which is 1.0x10 times the DAC of 0.2 Bq m for Class Y Pu-239.

l Conclusionfor the Efficacy of an Annual Chest Count as an Internal Assessment Procedure:

i Although the MDA derived quantities for Class Y Pu-239 are much lower than those for O

Class W Pu-239, chest counting in either case is not sufficiently sensitive and accurate for meeting the needs of an internal radiation protection program. Although more frequent f

monitoring improves the IRF and sensitivity for detecting exposures to Class W Pu-239, such improvement is not adequate. Very little improvement from more frequent monitoring occurs for Class Y Pu-239 because the IRF value for chest counting after a few weeks does not change

]

much during the year following an acute intake. The PSL derived quantities were not calculated; they would be approximately, but no larger, than 0.5 the MDA derived quantities shown above.

False positive chest counting resu!ts 2re expected to occur 5% of the time. At the physically

[

significant level, such false positive results correspond approximately to a committed effective q

dose equivalent of about 10,000 rem for Class W Pu-239 and 80 rem for Class Y Pu-239 for j

annual chest counts. Although subsequent counts can be used to lower the false positive rate, i

such a procedure should not be used to conclude that a significant exposure has not occurred, i

especially when positive results from a BZA or other air filter sample have already shown that an exposure may have occurred. These calculations show that annual chest counting is not an appropriate routine bioassay procedure because very significant exposures, intakes, and doses can easily go undetected. Chest counting for plutonium certainly does not me-* ** --~4-~~

f of routine bioassay, which is thefinal quality controlprocedure used to assure that workers are infact receiving adequate protectionfrom internal radiation exposures.

I1 I

'O

2. Assessments ofInternal Exposures to Pu-239 by Annual Urinalysis:

The minimum detectable activity (MDA) of Pu-239 by urinalysis is given as 0.02 pC; L'I (0.001 Bq for a 1.4 L,24 h urine sample) in the Internal Dosimetry Technical Basis Manual for the Savannah River Site (report number WSRC-IM-90-139 dated 12/20/90). This value is used below to calculate the MDI, MDE, MDH, and MDC for inhalation intakes of Class W

)

E compounds and Class Y compounds of Pu-239 for an annual,24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> urine sample using the IRF l

values shown below. The intake retention fraction (IRF) values were calculated for the l

respective compound classes using the INDOS computer code and intake retention functions 3

generated from the plutonium urinary excretion function reponed by Jones. Assumptions used in the calculation include a 1 micrometer AMAD aerosol, the respiratory and GI tract models in ICRP Publication 30, and an annual frequency of monitoring, i.e., the IRF values were calculated l

for a urine sample taken at 365 days after a single acute intake.

l i

2.1 Class W Pu-239: MDA = 0.001 Bq andIRF = 4.88x10 for annual urinalsis:

t Other applicable parameter values used in the calculations are:

O ALI = N-ALI = 200 Bq. and DAC = 0.08 Bq m l

-3 t

S-ALI = 400 Bq; FRM = 20 L min ~I; and 1

T = 8 h = 480 min = exposure interval for calculating the MDC.

l i

The minimum detectable intake (MDI) is calculated for the IRF value of 4.88x10-6 applicable to the urine sample:

MDA 0.001 Bq

= 205 Sq (1.03 ALI).

j MDI =

=

2RF 554:j

~

l The minimum detectable exposure (MDE) having a 95?c chance of being detected is calculated i

for the urine sample:

3

3ones, S.R.,

1985, Derivation and Validation of a Urinary Excretion Functionfor O

Plutonium Applicable Over Tens of Years Post Uptake, Radiation Protection Dos meny 11.

i 12 l

1 I

J

i O

2,000 DAC-h 2,000.DAC-h 1

MDE = MDI(

) = 205 Bq(

) = 2,050 DAC-h.

ALI 200 Bq The minimum detectable committed effective dose equivalent (MDH ) having a 95% chance of E

being detected is calculated:

b 5 rem 5 rem E*

(

MDH 9

S-ALI 400 Bq which also is the committed effective dose equivalent expected to be missed 5% of the time.

l l

For an exposure time T of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (480 min), the minimum detectable concentration (MDC) having a 95% chance of being detected is calculated:

MDI 205 Bq

-3 21 Eq m MDC =

=

=

TF (480 min) (0.02 m min-1) l 3

g t

f

-3 which is 260 times the DAC of 0.08 Bq m for Class W Pu-239.

'0 2.2 Class Y Pu.239: MDA = 0.001 Bq, and1RF = 8.02x10 for annual urinalysis:

Cther applicable parameter values used in the calculations are:

ALI = N-ALI = 500 Bq, and DAC = n.2 Sq.m

{

-3 6

a.d i

g = 20 L min 5-ALI = 600 Eq; F T = 8 h = 480 min = exposure interval for calculating the MDC.

The minimum detectable intake (MDI) is calcula:ed for the IRF value of 8.02x10 applicable to the mine sample:

MDA 0.001 Bq 3

= 1.25x10 Bq ( 2.5 ALI).

MDI =

=

IRF 8.02x10 I

f The minimum detectable exposure (MDE) having a 95% chan:e of being detected is calculated for the unne sample:

2,000 DAC-h 2,000 DAC-h 1.25x10 Eq(

) = 5.,000 DAC-h.

l MDE = MDI(

)

=

i O

^L1 soo 82 13

~!

O The mi#imum actectadie committea errective do e equivateat tuon > aavies a 95* chance or I

e being detected is calculated:

5 rem 5 rem 9 (

}

)

MDHE*

S-ALI 600 Bq which also is the committed effective dose equivalent expected to be missed 5% of the time.

For an exposure time T of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (480 min), the minimum detectable concentration (MDC) having a 95% chance of being detected is calculated:

3 f

MDI 1.25x10 Bq

-3 130 Eq m MDC =

3

.=

=

TF (480 min)(0.02 m min")

l g

-3 which is 650 times the DAC of 0.2 Bq m for Class Y Pu-239.

l Conclusionfor the Efficacy ofAnnual Urinalysis as an Internal Assessment Procedure:

t Minimum detectable derived quantities calculated for Class Y Pu-239 for urinalysis are O

somewhat higher than those calculated for Class W Pu-239, which is just the opposite of the comparison obtained for chest counting. The higher derived minimum detectable quantities j

calculated for Class Y Pu-239 for urinalysis result from the smaller IRF value. Although i

)

urinalysis is a much more sensitive procedure than chest counting, urinalvsis for Pu-239 is not i

sufficiently sensitive as routine bioassay procedure. The PSL derived quantities were not 4

calculated; they would be no larger than 0.5 of the MDA derived quantities shown above. False I

i positive urinalysis results are expected to occur 5% of the time, At the physically significant level, such false positive results correspond to a commined effective dose equivalent of about I.3 l

t rem for Class W Pu-239 and 5 rem for Class Y Pu-239 for annual urinalysis. Although

.j subsequent counts of a sample can be used to lower the false positive rate. such a procedure should not be used to conclude that a significant exposure has not occurred, especially when positive results from a BZA or other air filter. sample have already shown that an exposure may j

have occurred. These calculations show that annual urinalysis is not an appropriate routine bioassay procedure because exposures, intakes, and doses above the committed dose limits can l

l easily go undetected.

i

. O 14

=_.

1 Q

3. Assessments ofInternal Exposures to Pu-239 by Annual Fecal Analysis:

l The minimum detectable activity (MDA) of Pu-239 by fecal sampling and analysis is given as 0.1 pC; per sample (0.0037 Bq) in the Internal Dosimetry Technical Basis Manual for the Savannah River Site (report number WSRC-IM-90-139 dated 12/20/90). This value is used below to calculate the MDI, MDE, MDH, and MDC for inhalation intakes of Class W j

E compounds and Class Y compounds of Pu-239 for an annual,24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> fecal sample using the IRF values shown below. Again, the intake retention fraction (IRF) values were calculated for the -

l respective compound classes using the INDOS computer code and intake retention functions

)

generated from the plutonium urinary excretion function reponed by Jones. Although this excretion function by Jones is only applicable to urinalysis, it can be used to derive a pseudo j

systemic retention function for estimating systemic excretion by the fecal pathway. Most fecal excretion 1 year after an intake comes from direct clearance of companments in the respiratory tract, and the error introduced by using the Jones function is small, particularly for Class Y Pu-239. Other assumptions used in the calculations include a 1 micrometer AMAD aerosol, the respiratory and GI tract models in ICRP Publication 30, and an annual frequency of monitoring, i.e., the IRF values were calculated at a time of measurement 365 days after a single acute intake.

0

' Q 3.1 Class WPu-239: MDA = 0.0037 Bq andIRF = 1.3x10 for an annualfecalsample:

l Other applicable parameter values used in the calculations are:

ALI = N-ALI = 200 Bq, and DAC = 0.08 Bq m-3,

-1 j

g = 20 L min

and S-ALI = 400 Eq; F T = 8 h = 480 min = exposure interval for calculating the MDC.

The minimum detectable intake (MDI) is calculated for the IRF value of 1.3x10-5 applicable to the fecal sample:

MDA 0.0037 Bq 4

=

= 280 Bq (1.4 ALI).

MDI =

IRF 1.3x10" The minimum detectable exposure (MDE) having a 95% chance of being detected is calculated for the fecal sample:

2,000 DAC-h 2,000 DAC

• 2,800 DAC-h.

)

MDE = MDI(

) = 280 Sq(

)

=

ALI 200 Sq j

O The minimum detectable committed effective dose equivalent (MDH ) having a 95% chance of E

15

'l

t i

]

being detected is calculated:

l 5 rem 5 rem MDHE = MDI (

) = 280 Bq (

) = 3.5 rem, S-ALI 400 Bq which also is the committed effective dose equivalent expected to be missed 5% of the time.

1 For an exposure time T of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (480 min), the minimum detectable concentration (MDC) having a 95% chance of being detected is calculated:

MDI 280 Bq

-3 29 Bq m MDC =

=

=

TF (480 min) (0.02 m min-1) 3 g

which is 360 times the DAC of 0.08 Bq m-3 for Class W Pu-239.

4.2 Class Y Pu-239: MDA = 0.0037 Bq, andIRF = 8.5x10'# or an annualfecalsample:

j f

Other applicable parameter values used in the calculations are:

ALI = N-ALI = 500 Bq, and DAC = 0.2 Bq m'

-1 S-ALI = 600 Bq; Fg = 20 L min

and T = 8 h = 480 min = exposure interval for calculating the MDC.

1 The minimum detectable intake (MDI) is calculated for the IRF value of 8.5x10-5 applicable to i

the fecal sample:

~

MDA 0.0037 Sq

= 44 Bq ( 0.088 ALI).

i MDI =

=

IRF 8.5x10 "

i i

The minimum detectable exposure (MDE) having a 95% chance of being detected is calculated for the fecal sample:

2,000 DAC-h

_2, 0 0 0 D A C - h --

5 MDB = MDI(

) = 44 Bq(

) = 180 DAC-h.

i ALI 500 Bq l

t l

The minimum detectable committed effective dose equivalent (MDH ) having a 95% chance of E

being detected is calculated:

5 rem 5 rem h

MDH

= MDI (

) = 44 Bq (

) = 0.37 rem, y

S-ALI 600 Bq

~

16 I

~

6 I

which also is the committed effective dose equivalent expected to be missed 5% of the time.

f i

For an exposure time T of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (480 min), the minimum detectable concentration (MDC) i i

having a 95% chance of being detected is calculated:

-i MDI 44 Bq

-3 4.6 Bq m MDC =

=

=

3 TF (480 min) (0.02 m min ')

(

g

-3 which is 23 times the DAC of 0.2 Bq m for Class Y Pu-239.

t Conclusionfor the Efficacy ofAnnual Fecal Sampling as an Internal Assessment Procedure:

Minimum detectable derived quantities calculated for Class Y Pu-239 for an annual fecal i

sample are much lower than those calculated for Class W Pu-^239, which is just the opposite of the comparison obtained for urinalysis. The lower derived minimum detectable quantities calculated for Class Y Pu-239 for fecal analysis result from the larger IRF value associated with j

the longer clearance half-life of compartment g in the ICRP Publication 30 respiratory tract I

model. All other respiratory tract compartments, even after a period of only 2 weeks following I

O an intake, make an insignificant contribution to fecal excretion. Compartment g initially contains j

i 10% of the activity associated with an intake of 1 micrometer AMAD aerosols. The clearance half-life of compartment g is given as 50 days for Class W compounds and 500 days for Class Y i

compounds. Therefore,it almost acts a linear integration of a worker's exposure to Class Y compounds for an annual frequency of monitoring. Obviously, the actual clearance of radioactive aerosols deposited in the respiratory tract of a worker could differ greatly from the models assumed for Reference Man. The models for Reference Man, if no better information is available and properly documented, do provide a basis for the evaluation of the etticacy of a

.j bioassay procedure of this fecal monitoring procedure.

4 1

Although annual fecal monitoring for Class W compounds is not sufficiently sensitive as a routine bioassay procedure, it can be used to detect Class Y compounds below the applicable committed dose limits. To avoid detecting more recent low level intakes, routine fecal samples should be taken after a two week vacation.

When breathing zone air sampling screening results, area air monitoring results, or other information indicates that workers might have had significant exposures to either compound j

class of Pu-239, then fecal sampling and analysis should be initiated immediately to confirm and O'

make a better estimate of the actual intakes of workers. All fecal excretion for at least a week j

and preferably two weeks following the time of the intake should be collected and analyzed.

17

e O

The eS' derivea 9eaatities were aot caiceiatea: they woutd de o iarser thaa o s or the MDA derived quantities shown above. False positive fecal analysis results are expected to occur 5% of the time. At the physically significant level, such false positive results correspond to a committed effective dose equivalent of about 1.8 rem for Class W Pu-239 and only 0.18 rem for Class Y Pu-239 for an annual fecal sample. Although subsequent counts of a sample can be used to lower the false positive rate, such a procedure should not be used to conclude that a significant exposure has not occurred, especially when positive results from a BZA sample indicate that an exposure has occurred. To reject a positive result from a BZA sample, the sensitivity and accuracy of the fecal analysis procedure would have to be improved to a level comparable to that of the BZA assessment. These calculations show that annual fecal sampling and analysis following a two week vacation period may be an appropriate routine bioassay procedure for Class Y compounds of Pu-239, especially if the MDA can be lowered for example to 1/5 of the current MDA.

CONCLUSION Area air monitoring programs, personal air monitoring programs. such as those provided by PAS and other types of BZA samplers, and bioassay programs are not mutually exclusive.

All of these programs may be required for an adequate internal radiation protection program.

Continuous air monitors (CAMS) provide waming that can prevent significant exposures.

Alarms can be used to initiate follow-up special bioassay procedures. Personal air samplers may be required to monitor and assess the exposures of workers who may have chronic intakes or significant potential for large accidental intakes that are difficult to detect at the level required for the protection program. Such use of BZA samplers may be particularly important when concentrations can vary greatly in space and time, e.g., when a worker's activities might cause releases to the working environment that might not be detected by CAMS. The use of personal air samplers (PAS) provides the most sensitive and most accurate measure of the exposure of a worker to airborne radioactive aerosols.

They provide a direct measure of the adequacy of control of working envircnments and the exposure of workers to those environments in a timely way that improves the overall internal radiation protection program, including both the routine and special bioassay programs.

O 18