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29£OQ2RQ-RN-2-4-10j> 

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Internal SFUND Box Number 35 OF 283 

Initials/Date: LDS / 3-8-2007 

/^EsyNO 
P.2 

I 

NUREG-1140 

A Regulatory Analysis on 

Emergency Preparedness for 

Fuel Cycle and Other 

Radioactive Material Licensees 

Draft Report for Comment 

Manuscript Completed: May 1985 

Date Published: June 1985 .;, 

S. A. McGuire 

Division of Risk Analysis and Operations 

Office of Nuclear Regulatory Research 

U.S.IMuclear Regulatory Commission 

Washington, D.C. 20555 

.

 %

.~ y 

/ 
P.3 

EXECUTIVE SUMMARY 

The question this Regulatory Analysis sought to answer is: Should the. . 

NRC impose additional emergency preparedness requirements on certain fuel cycle 

and other radioactive material licensees for dealing with accidents that might, 

have offsite releases of radioactive material? To answer the question, we 

analyzed potential accidents for 15 types of fuel cycle and other radioactive 

material licensees. -'..;•'_ 

The most potentially hazardous accident, by a large margin,-was determined 

to be the sudden rupture of a heated-multi-ton cylinder of UF6. Acute fatalities 

may be possible in the -immediate vicinity of the_ release point. Acute permanent -

injuries may be possible for many hundreds of meters, and clinically observable 

transient injury of unknown long term consequences may be possible for distances 

up to a few miles. These effects would be caused by the chemical toxicity of 

the UF6 and the products resulting from its reaction with moisture. Accompanying 

* 

radiation doses would not be of significance. 

The most potentially hazardous accident due to radiation exposure was" 

determined to be a large fire at certain facilities handling large quantities 

of alpha-emitting radionuclides: (i.e., Po-210, Pu-238, Pu-239, Am-241, Cm-242, 

Cm-244) or radioiodines (1-125 and 1-131). However, acute fatalities or 

injuries to people offsite due to accidentfa1 releases of these materials do 

not seem plausible. A few other radionuclides are of lesser importance (H-3, 

P-32, Sr-90, and Cs-137>. If plutonium were to be handled at fuel cycle 

facilities, a fire~in such a facility may a-lso be of significance. 

The only other significant accident was identified as a long-term 

pulsating criticality at fuel cycle facilities handling high^enriched uranium 

or plutonium in aqueous solutions. 

Aside from fires or accidents that lead to fires, UF6 releases, or criticality accidents, no other significant accidents were identified. Explosions 

were not seen to yield as large a release unless they were followed by a fire. 

Tank ruptures were not identified as having a potential for significant releases. 

Earthquakes also were not identified as leading to significant releases unless 
they were followed by a fire. Tornados might cause large release, but would 

disperse the material so widely that significant doses would not result. 

The criterion for deciding whether an accident was significant is whether 

a release could cause a person outside the plant on the plume centerline to 

receive within 8 hours an effective dose equivalent of more than 1 rem," a. 

thyroid dose of more-than 5 rems, or an intake of soluble uranium exceeding 

2 mg. One rem whole body or 5 rems thyroid are at the lower end of the range 

of dose for which the EPA says protective actions should be considered. A 

rough screening identified 64 licensees authorized to possess quantities of 

radioactive matertals.for which an accident release could cause doses exceeding 

1 rem effective dose equivalent, 5 rems thyroid, or intake of 2 mg of soluble 

uranium. However, some of these licensees do not actually possess all of the 

materials authorized by the license. These licensees are likely to amend their 

licenses to lower their possession limits rather than submit a plan.. Other 

licensees could demonstrate that a significant release is not possible. In 

actuality, only about 30 plans would be submitted. For most of these licensees, 

however, the degree of hazard is smalH For most such licensees maximum doses 

for even the most severe postulated accidents are only a few reus. The areas 

within which people should take protective actions are small - for most 

licensees much less-than a square mile. At most sites these areas would contain 

few people. And the probability of a major release is small - less than 10-4/yr. 

The probability of a major release simultaneous with highly adverse meteorology 

is less than 10-5/yr. Thus the probability of even a single person's receiving 

a dose in excess of 1-rem at.most of ;these facilities has about the same 

probability as a core melt accident with containment failure at a nuclear power 

plant. . 

A further feature of-major ace-rderits at such licensed facilities is that 

airborne releases are likely to occur rapidly with Tittle warning. Ruptures 

of UF6 cylinders and fires would give little or no warning. Releases could 

start before a fire is detected or shortly thereafter. Plume travel time to 

nearby people is likely to be no more than a few minutes. Releases would in a 

majority of cases end within half an hour to an hour when the fire department 

controlled the fire. In most instances, actions taken half an hour after 

accident detection would be fairly ineffective. Actions taken 15 minutes 

after accident detection could on the average be expected to be roughly 50% 

effective. Thus protective actions would have to be taken quickly. This 

requires predetermined actions to be taken by people in the vicinity of the 

ii 
P.5 

site to notify nearby residents. There is not likely to be enough time for 

dose projections or complicated decisionmaking during the accident, nor for 

participation of personnel not in the immediate vicinity of the site. 

An appropriate plan would (1) identify Occidents for which protective 

actions should be taken by people offsite. (2) list the licensee's responsibilities for each type.of accident, including notification of local 

authorities (fire and police generally), and (3) give sample messages for 

local authorities including protective action recommendations. This approach 

more closely follows the approach used for research reactors than for power 

reactors^ The low potential offsite doses (acute fatalities and injuries not 

possible except for UF6 releases), the small areas where actions would be 

warranted, the small number of people involved, and the fact that the local 

police and fire departments would be doing essentially the same things they 

normally do, are all factors that tend to make a simple plan adequate. 

m 
P.6 

CONTENTS 

EXECUTIVE SUMMARY . 

ACKNOWLEDGEMENTS . .. 

1. PROPOSED ACTION 

1.1 Description of the Proposed Action 

1.2: Need for the Proposed Action . . . 

TECHNICAL BASIS FOR THjE PROPOSED RULE 

2.1 Methodology ...... . y . . 

2.1.1 The Accident History Approach .... . 

2.1:2 Accident Source Terms ........ . 

2.1.3 Calculations of Doses ... . 

2.1.4 Protective Action Guides ...... . 

2.1.5 A Discussion of the Conservatism in the 

Calculations . . . '. 

2.2 Fuel Cycle Facilities 

2.2.1 Uranium Mining 

2.2.2 Uranium Milling ...... . 

2.2.3 UF6 Conversion Plants . . . . 

2.2.4 Enrichment Plants . 

2.2.5 Fuel Fabrication - Uranium . 

2.2:6 Fuel Fabrication -Plutonium 

2.2.-7 Spent Fuel Storage 

2.2,8-New Fuel Storage 

2.2.9 Reprocessing of-Spent Fuel . 

2.2.10 Research with Nuclear Fuels . 

2.3 Byproduct Material Facilities 

2.3.1 A Generic Overview ...... . 

2.3.2 Radiopharmaceutical Manufacturing 

2.3.3 Radiopharmacies and Hospitals . . 

2.3.4 Sealed Source Manufacturing . . . 

2.3.5 University Research Laboratories 

2.3.6 Waste Warehousing and Burial . . 

2.3.7 Depleted Uranium Products ... . 

2.4 Summary of Facilities to be Covered 

2.5 A Protective Action Strategy . . . . 

Pagjs 

i . 

ix 

1 

1 

3 . 

4 

6 

7 

12 

14 '. 

- 17 _-.-

17 

17 

25 

37 

39 

50 

56 

59 

59 

64 

66 

67 

81 

83 

84 

89 

90 

93 

96 

98 
P.7 

CONTENTS (Continued) 

"2.5.1 The Initial Response I 

2.5.2 The Relocation Phase ......... . 

2.5.3 The Assessment Phase . .. . . .... . 

2.5.4 The Food Pathway . . ... . . . . . ... . . . . . 

3. VALUE/IMPACT . . . ,,. . ..v. . . . 

3.1 Alternatives. .-. . . .. . ^ .......... . A.. . . 

3.2 -Value of the Propose Action . .-. • ... • • ... . . . . . 

. ,"' 3.3 Cost... . r ,; . . . ._ . . .. . ..... . . . . . . ". . . . 

3-r4 Value/Impact of Alternatives 

4. STATUTORY CONSIDERATIONS ..-,. 

4.1" NRC Regulatory Authority . . . . 

4.2 Agreement States . . 

4.3 Environmental Impact Appraisal .............. . 

5. CONCLUSIONS 

Page 

98 

104 

105 

107 

107 

107 

107 

109 

110 

111 

111 

111 

111 

112 

FIGURE 

1. Atmospheric dispersion calculated by CRAC2 for F stability, 

1 meter/second wind

 :speed .. . . ."•...-• ..... . ...." 

v TABLES 

1. Fires in Uranium Mills ... . 

2. Uranium Mill Tailings Releases, 1959-1984 

3. Accident Source Terms and Doses From Uranium Mill Accidents . . 

4. Accidents Involving UF^ Releases . 

5. Criticaiity Accidents InFuel Cycle Facilities -...'.... . 

6. Fires and Explosions Involving Uranium and Thorium 

7. Other Accidental Releases from Uranium Fuel Fabrication Plants 

8. Amounts of Radioactive-Materials Released to Room Air Due to 

a Critical ity Accident ...... . . . . . . . ., 

9. Offsite Doses Calculated for Fuel Fabrication Plants 

10. Calculated Releases and Doses from Spent Fuel Storage Accidents 

11. Fires and Explosions Involving Release of Byproduct Materials . 

12. Accidental Releases of Byproduct Material Except Fires 

and Explosions ..... . 

13. Quantities of Radioactive Materials Requiring Evaluation of the 

Need for Offsite Emergency Preparedness 

14. Radiopharmaceutical Manufacturing: Maximum Possession Limits, 

Release Fractions, and Doses Due to a Major Facility Fire . . . 

15. Radiopharmacy: Maximum Possession Limits, Release Fractions, 

and Doses Due to a Major Facility Fire 

11 

19 

20 

22 

27 

40 

42 

43 

45 

47 

60 

68 

69 

75 

82 

85 

Vi 
CONTENTS (Continued) 

TABLES (Continued) 

Page 

16. Sealed Source Manufacturing: Maximum Possession Limits, 

Release Fractions, and Doses Due to a Major Facility Fire . . . . 87 

17. University Research Laboratories: Maximum Possession Limits . 

Release Fractions, and Doses Due to a Major Fire ........ . 90 

18. Waste Warehousing Airborne Releases and Doses Due to a Major 

Facility Fire .. . . . ,. . 91 

19. Beaufort Wind Scale (Simplified) . _ 99 

20. Pasquill's Relation of Turbulence Types to Weather Conditions . . 100 

21. Dose for Different,Meteorologial Conditions Relative 

to F, 1 m/s ........... . . ....... . 100 

22. Downwind Distances to Which .Protective Actions Should Be 

Considered for an Accident Involving Five Times the Quantity 

of Material in Table 13 .:.:.. . 101 

23. Downwind Distances to Which Protective Actions Should Be 

Considered for.an Accident Involving 50 Times the Quantity 

of Material in Table 13 .:. . . 103 

24. Comparison of Costs and Benefits of Special Emergency 

Preparedness 110 

vi 1 
P.9 

ACKNOWLEDGEMENTS 

We would like to thank the following people for helping us prepare this 

Regulatory Analysis. 

David Bennett, Eugene Runklej Daniel Alpert, Kenneth Adams, and 

David Aldrich of Sandia National Laboratory for performing dose calculations 

and meterological modeling. 

Peter Owczarski, Sue Sutter, Jofu Mishima, and Mark Halverson of BattellePacific Northwest .Laboratory for performing release fraction experiments, 

analyzing release fraction experiments, reviewing the literature on release 

fractions, developing accident scenarios, and analyzing accidents. 

Moshe Simon-Tov and Reid Williams of Martin-Marietta Energy Systems, Oak 

Ridge, for performing analyses of UF6 releases. Robert Just for discussions 

on the health effects of UF6. 

Andreas Jensen of the John Hopkins University Applied Physics Laboratory 

for descriptions_of the U.S. Department of transportation accident modeling:: 

methods. - .'_".'••-

Joseph Logsden of the U.S. Environmental Protection .Agency" for discussions 

on the use of the EPA's protective action guides. 

Paul Morrow of the -University of Rochester and McDonald Ed Wrenn of the 

University of Utah for discussion on the chemical toxicity of uranium. 

Duane Hall of 3M for hosting a tour of the 3M sealed source manufacturing 

plant in New Brighton, Minn, and for discussion on-jthe release fractions of 

polonium-210. . - ' .. " 

Edward Janzow of Monsanto Research Corporation for hosting a tour of the 

Monsanto Dayton, Ohio, sealed source" manufacturing.plant and for information 

on the particle size of americium oxide powder. 

Melita Rodeck of the Federal Emergency Management Agency for helpful 

comments. 

Al Arcuni and Sandy Mullen of International Energy Associates Limited 

for helpful comments. 

ix 
On the NRC's Office of Nuclear Material Safety and Safeguards staff: 

Fritz Storz, Lee Rouse, and John Roberts for assistance in preparing the 

section on spent fuel storage; Dennis Sollenberger and George Gnugnoli for 

assistance in preparing the section on uranium milling; Fred Fisher, Ed Shum, 

and R. Gerry Page for suggestions throughout the analysis; Joseph Wang for 

assistance in preparing sections in byproduct material licensees; Ron Cardarelli 

for help in collecting information in emergency plans submitted to NRC. 

On the NRC's Office of Nuclear Regulatory Research staff: Steven Bernstein 

and Don Solberg for assistance on accident scenarios and release fractions, as 

well as contract assistance; Robert Kornasiewicz and Leta Brown for comments 

and assistance on meterological modeling; James Martin and Leonard Soffer for 

comments and assistance with regard to existing NRC policy, on emergency preparedness, especially*as it relates to nuclear power plants; James Norberg, 

Philip Ting, Robert Bernero, James Malaro, and Karl Goller for many helpful 

comments and valuable support. 

On the NRG's Office of Nuclear Reactor Regulation staff: Irwin Spickler 

for many valuable comments^and discussions on meterological modeling. 

On the NRC's Office of Analysis and Evaluation of Operational Data staff: 

KathleenBlack and Eugene Trager for help in collecting information on accident 

histories. 
P.11 

A REGULATORY ANALYSIS 

ON 

EMERGENCY PREPAREDNESS FOR FUEL CYCLE 

AND OTHER RADIOACTIVE MATERIAL LICENSEES 

1. . PROPOSED ACTION _ • _-

1.1 Description of the Proposed Action .-

This regulatory analysis evaluates the need for a proposed rule to require 

additional emergency preparedness for certain fuel cycle =and other.radioactive 

material licensees. The purpose of the rule would be to~require, for certain 

licensees who are authorized to possess radioactive materials in large quantity, 

"emergency procedures for offsite releases of radioactive"materials. =These 

would include: 

(1) A brief description of the licensee's facility and the area near, 

the site. -

(2J _An identification of each type of accident for which"protective 

-actions for the public may be needed. -_ 

(3) An analysis of the potential doses to.the public from each type of 

accident, including doses from naturally occurring and accelerator 

produced radioactive material in addition to all licensed material. 

(4) Identification of the means of detecting each type of accident in a 

timely manner. 
(5) A brief description of the means and equipment for mitigating the 

consequences of each type of accident, including those provided to 

protect workers onsite. 

" - - - . - * . - - -

(6) A brief description of the methods and equipment to assess releases 

of radioactive materials. . 

(7) A brief description ofthe responsibilities of licensee personnel 

should an accident occur, including identification of personnel 

responsible, for notifying offsite authorities and notifying the NRC. 

(8) A brief description of the means of promptly notifying offsite 

authorities and of requesting offsite assistance resources. 

(9) A brief description of the methods for assuring that recommended 

protective actions and distances are unambiguously communicated to 

_- response organizations. 

(10) A brieT description of any special instructions the licensee would 

give "to fire, police, medical and other emergency personnel. 

(11) A brief description of the means of restoring the facility to a safe 

condition after an accident. 

{1-2) Provisions for conducting onsite drills at intervals not to exceed 

two years. : -

. The question is not whether licensees should have any emergency preparedness, 

that question was addressed long ago. The NRC has long required licensees to 

be prepared to cope with emergencies. The question is whether there should be 

additional requirements. For example, should NRC require formal written state 

and local government plans for coping offsite with serious radiation accidents? 

Such plans might include provisions for early evacuation by the public or 

notifying them to take shelter indoors. 

The question is also not whether State and local governments should have 

emergency preparedness capabilities for dealing with radiation accidents. 
P. 13 

Police departments, fire departments, and other agencies that are routinely 

prepared to cope with emergencies already exist. This rulemaking is intended 

to assure that, where needed, there exist emergency procedures for mitigating 

and coping with offsite releases. 

We must also distinguish between emergency response and formal emergency 

procedures. If an accident happens, the licensee and State and local governments can be expected to respond to the best of their abilities whether or not 

there are any formal written emergency procedures for offsite releases. The 

question is whether the response would be adequate if there were no Written 

emergency procedures for coping with the emergency. 

This regulatory analysis considers plausible accidents for various types 

of fuelcycle and other radioactive material licensees. The analysis then 

determines for.which types of facilities the response to an accident would 

involve so many people, such large areas, and require such prompt actions that 

the response would not be adequate unless there were available additional 

special resources beyond those already existing. 

1.2 Need for the Proposed Action 

The NRC has always.required that its licensees take steps to reduce the 

likelihood of serious accidents to a minimal level, but yet be prepared" to cope 

with accidents should they occur. However, during the Commission's deliberations on rulemaking concerning nuclear power plant emergency preparedness 

following the accident at Three Mile Island, the Commission directed the staff 

to evaluate the jieed to strengthen the emergency preparedness requirements for 

fuel: eyele-and other radioactive material licensees. 

In late 1980, the staff reevaluated previously submitted emergency plans 

for fuel fabrication plants and found some weaknesses in the plans as written. . 

For example, some plans did not describe (1) timely alerting-of potentially 

affected public to a hazard, (2) recommendations for specific actions the 

public should take to protect itself, such as sheltering or evacuation, and 

(3) arrangements for prompt notification of NRC and state and local government 

agencies. 

In February, 1981, the NRC issued orders to 62 licensees to either submit 

comprehensive radiological emergency plans or lower their possession limits 

for radioactive material. About half of the licensees submitted plans, and' 
" » 

P.14 

half lowered their possession limits or surrendered their license. Then, an 

Advance Notice of Proposed Rulemaking on the subject of offsite emergency preparedness was published in the Federal Register on June |3, 1981 (46 FR 29712). 

2. TECHNICAL BASIS, FOR THE PROPOSED RULE y 

2.1 Methodology 

This regulatory analysis identifies the classes of fuel cycle and other 

radioactive material licensees that could have accidents that might result in 

radiation doses to the public exceeding protective action guides established 

by the EPA. (Chemical toxicity is also considered for the special case of 

uranium hexafluoride and soluble uranium releases.) The plausibility of 

exceeding the EPA's protective "action guides was considered.from two points of 

view: (1) the accident history of fuel cycle and byproduct material licensees, 

and (2) theoretical calculations of the releases and offsite doses of accidents 

considered to= be possible. -" '_ -: •'•• = 

2.1.1 The Accident. Hi story Approach . ' -

The historyof accidents involving radioactive byproduct material (Part 30 

licensees), source material (Part 40 licensees), special nuclear material 

(Part 70 licensees), and spent fuel storage (Part 72 licensees) was surveyed. 

In.summary, we found, no evidence that any accidental release of radioactive 

material from facilities of these"types has ever caused an effective dose 

equivalent to any individual offsite exceeding even 1% of the EPA's 1-rem 

protective actidrv guide. - -

The value of the historical review is that it identifies types of accidents 

that have occurred and draws our attention toward accidents similar to those 

that have occurred. In addition, accidents that have happened cannot have 

violated the physical laws of nature. This statement cannot be made about 

theoretical calculations. Theoretical calculations can use simplifying assumptions that are internally inconsistent or inconsistent with the laws of nature. 

More information on the nature, causes, and frequency of accidents is 

available for fuel cycle and other radioactive material licensees than is 

available for nuclear power plants because there are so many more fuel cycle 
and other radioactive material licensees than nuclear power plants. Currently 

the NRC regulates about 9,000 non-reactor licensees. In addition, Agreement 

States regulate roughly another 12,000 non-reactor licensees. A large number 

of these licensees have operated for many years, and the combined experience 

of these licensees approaches half a million licensee-years. By contrast, 

there are only about 70 nuclear power plants operating with about 1,000 plantyears of combined operating experience. Thus more experience exists for fuel 

cycle and other radioactive material licensees than exists for nuclear power 

plants. 

Operating experience may be more relevant for these licensees than for 

nuclear power plants because of the nature of the accident driving force. In 

nuclear power plants the driving force is the enormous amount of heat in the 

reactor. The available energy is so large that some unique occurrences are 

conceivable, such as molten cores, large-scale metal-water reactions, and 

rupturing the containment by overpressurization. Because these events have 

never happened they can. only be^studied theoretically. The dominant driving 

forces for accidents at_non-reactor licensees are common industrial accidents— 

fires, chemical explosions, leaks, and the like. A great deal of industrial 

accident experience can be drawn upon in analyzing these.potential accidents. 

Much information on the accident history of fuel cycle and other radioactive material licensees is available. NRC regulations require the reporting 

of all significant events. All licensees must notify NRC of (1) the overexposure 

of any individual to radiation [10 CFR §20.403 and §20.405], (2) the airborne 

release of large quantities of radioactive material [10 CFR §20.403 and 

§20.405], (3) the loss of one day or morerof the operation of any facility 

[10 CFR §20.403], (4). damage to property in excess of $2000 [10 CFR §20.403], 

(5) the loss or theft of licensed material [10 CFR §20.402], (6) excessive 

radiation levels or contamination on packages received [10 CFR §20.205], and 

(7) major defects in equipment or noncompliance with regulations that have 

major safety significance [10 CFR §21.21]. In addition, there are other 

reporting requirements in NRC regulations that apply to specific classes of 

licensees. The reports that licensees have filed provide extensive information 

for evaluating the history of accidents in this part of the nuclear industry. 

Since January 1975, the NRC has carefully and systematically published 

reports describing accidents of significance, "Report to Congress on Abnormal 
P.16 

Occurrences," NUREG-0090. Accidents in the fuel cycle and at byproduct material facilities are included for NRC licensees and Agreement State licensees. 

All these reports for the period from January, 1975, to September, 1983, were 

reviewed for this Regulatory Analysis. 

Also, a comprehensive file of over 5,000 events from the period of 1950 

through 1978 was compiled at Argonne National Laboratory. The file includes 

primarily events from NRC licensees (formerly AEC licensees) and governmentowned laboratories run by DOE (formerly ERDA and AEC), and some reports on 

accidents from Agreement State licensees and facilities in foreign countries. 

The file inciudes events from nuclear power plants, the nuclear fuel cycle, 

and other radioactive material usage.. - A brief summary of 1634 events in the 

commercial nuclear fuel cycle was published in 1981.?" The Argonne file also 

includes some events from 1979, but is not complete for that year. 

In addition, other sources were searched for this analysis, such as NRC's 

annual reports on radiation exposures, NRC's preliminary notification of 

unusual event reports, the memory of NRC staff members, and accounts published 

in the open literature.

 : . = -

Using all these sources of information this analysis should include most 

significant accidents for the years front 1950 through 1983. Although some 

events may have been omitted, it is beTieved that all relevant events in the 

United States that caused a significant release of radioactive material outside 

a restricted area have been included. 

2.1.2 Accident Source Terms •- - . -

Many plausible accident scenarios were considered for various types of 

facilities. The NRC considers inSafety Evaluation. Reports a number of possible 

accidents and their effects on public health and safety befpre issuing licenses ._ 

for fuel cycle facilities. The NRC requires applicants to evaluate possible 

accidents. Additionally, the NRC performs its own analyses of several severe 

accidents to determine whether there is adequate protection of public health 

and safety. The NRC's analyses are then issued when the license is issued. 

This regulatory analysis makes use of those NRC staff analyses in developing 

accident source terms. 

*Deborah J. Bodeau et al., Data Base for Radiation Events in the Commercial 

Nuclear Fuel Cycle, 1950-1978, NRC Report NUREG/CR-2429, 1981. 
The significant accidents were determined to be UF6 releases, fires, and 

criticality accidents. Aside from the special cases of UF6 releases and criticalities, the release fractions for fires were considered to be larger than 

the release fractions for other types of accidents. Thus, release fractions 

for fires, as described in Section 2.3.1.2, are used to determine the need for 

emergency preparedness. _. 

2.l.3 Calculations of Doses 

Doses from airborne.releases were calculated by assuming release fractions 

for radioactive materials, assuming a atmospheric dispersion model, and calculating doses from three pathways-inhalation, cloud-shine, and ground-shine from 

particulates deposited on the ground. In general, the highest doses come from 

inhalation. 

Two kinds of doses were calculated: effective dose equivalent and child's 

thyroid dose.* The effective dose equivalent is the sum of the 50-year dose 

equivalent commitment to each body orjan multiplied by a weighting factor for 

each organ as given in ICRP Publication 26 and, for the skin, ICRP Publication. 28.** For the inhalation pathway, dose conversion factors from ICRP 

Publication 30*** were used. Thyroid dose is the dose equivalent delivered to 

the thyroid by inhaled radioiodines. The child's thyroid dose is calculated 

by multiplying the value calculated for an adult by two. 

For the two external dose pathways dose conversion factors from Kocher**** 

were used. An 8-hour ground exposure time and a 0.7 shielding factor (30% 

*David E. Bennett et al., Preliminary Screening of Fuel Cycle and Byproduct 

Material Licenses for Emergency Planning," NUREG/CR-3657, March, 1985. 

. **Statement from the 1978 Stockholm Meeting of the ICRP, Publication 28, 

International Commission on Radiological Protection, Pergamon Press, 

Oxford, 1978. 

***Limits for Intakes of Radionuclides by Workers, Publication 30, 

International Commission on Radiological Protection, Pergamon Press, 

Oxford, 1979 and 1980. 

****D. C. Kocher, "Dose-Rate Conversion Factors for External Exposure to Photon 

and Electron Radiation from Radionuclides Occurring in Routine Releases from 

Nuclear Fuel Cycle Facilities," NRC Report NUREG/CR-1918, 1981. Also D.C. 

Kocher, "Dose-Rate Conversion Factors for External Exposure to Photons and 

Electrons, "Health Physics, 45, 665, 1983. 

f'Reactor Safety Study," NRC Report WASH-1400, Appendix VI, page 11-23, 1975. 
reduction) for a not perfectly flat surface were used to calculate the external 

exposure from radionuclides deposited on the ground. The basis for the 8-hour 

exposure to ground shine is that even if there is no pre-existing offsite 

emergency preparedness it should be possible to locate areas with high dose 

rates due to ground shine (greater than 100 mR/hr) and move people out within 

.8 hours. The contaminated areas from which people would need to relocate would 

be small and generally near the site. 

The atmospheric dispersion model is a standard Gaussian plume model. Doses 

were calculated for two sets of meteorological conditions: stability class F 

with 1 m/s wind speed and stability class D with 4.-5 m/s wind speed. The F, 

1 m/s" assumptions are those traditionally used by NRC in hazard evaluations and 

represent very adverse weather conditions. The D, 4.5 m/s assumptions are 

those traditionally used by DOT in calculating evacuation distances for accidents 

involving toxic chemicalst and represent more typical weather. DOT considers 

evacuation distances based on D, 4.5 m/s adequate to protect public health and 

safety as demonstrated by experience with toxic chemical releases. The NRC's 

proposed rule associated with this analysis bases the need for emergency 

preparedness on the traditional NRC assumptions if F, 1 m/s. The doses calcuy 

lated using the DOT assumptions of D, 4.5 m/s are included for perspective to -

show doses that would be expected under more typical or realistic conditions. 

The intercept fraction for inhalation of 10-6 is considered to be about 

the maximum value likely to be inhaled in an accident.* Using F, 1 m/s meteorology and the assumptions described below, 10-6 corresponds to a distance of 

100 meters for the entire duration of the accident. Limiting the intaketd 10-6 

in effect means that a person on the plume centerline in dense smoke closer than 

100 meters from the release point will move out of the smoke before the release 

ends. Thus, the distance at which doses were calculated Was taken to be 100 m -

from the release point in our mathematical model. This distance results in an •>-_ 

intercept fraction of 0.89 x 10-6 for radioactive materials that deposit on the 

t"Hazardous Materials-Emergency Response Guidebooks," Materials Transportation . 

Bureau, U.S. Department of Transportation, DOT Publication D0T-P5800.2, 1980. 

*"Upgraded Emergency Preparedness for Certain Fuel Cycle and Material Licensees," 

46 Federal Register 29712, 1981; "Criteria for Selection of Fuel Cycle and 

Major Materials Licensees Needing Radiological Contingency Plans," NUREG-0767, 

1981; and Allen Brodsky, "Resuspension Factors and Probabilities of Intake 

of Material in Process (or Is 10-6 a Magic Number in Health Physics?)" 

Health Physics, 39, 992, 1980. 

8 
\ 

} 

around. In other words a person on the plume centerline is assumed to inhale 

at most about one one-millionth of the material released regardless of the 

«eteorolpgy. 

Dosjas were calculated for a person standing in an open field in the downwind direction on the plume centerline breathing at a rate of 2.66 x IO-4 m3/s. 

Atmospheric stability class F and wind speed of 1 m/s were assumed. Doses to 

people in buildings would be smaller than the doses given in this analysis 

because buildings provide shielding and some respiratory protection. Doses to 

people standing outside in urban and suburban areas or wooded areas would be 

less than those given here because obstacles to wind flow would cause the plume 

to broaden. 

Doses were calculated using a slightly modified version of the CRAC2 

computer code.*. The CRAC2 code has been used extensively by the NRC for calculations of doses that could result from nuclear power plant accidents. 

The key input parameters for the CRAC2 calculations are building size, 

release duration, release height, dose conversion factors for each radionuclide, radioactive halflife of each .radionuclide, and-deposition velocity 

for particulates. 

- - Building size determines the building wake factor or the initial plume 

dimension. In most cases the building size was assumed to be 25 m wide by 

•10-W'high. For buoyant releases no building wake factor was used. At close 

in distances the building wake effect from a 250 m2 building significantly 

reduces the concentrations of airborne materials from a release. At distances 

of 1000m, the building wake factor is relatively unimportant. 

The release duration determines the amount-of plume meander. Plume meander 

was not included at 100 m because the plume is considered to be still in the 

building wake'.-'--' 

Release height determines plume height and thus affects ground level . 

concentrations. Greater release heights cause lower ground level air concentrations. In this regulatory analysis, the release height was assumed to be 

ground level or at 20 meters. The 20-meter release height was used to account 

for buoyant plumes. 

*L.T. Ritchie et al., "CRAC2 Model Description," NRC Report NUREG/CR-2522, 1983. 
P.20 

The CRAC2 code accounts for the radioactive decay of materials in transit. 

In most cases this correction is negligible. However, Kr-89 with a 3-minute 

half-life dominates the external dose from a criticality accident, so that in 

this case the decay correction is significant. 

The CRAC2code does not calculate radioactive decay while a material is 

held up in a building before release. A separate correction factor Was developed to account for radioactive decay before release to the atmosphere. This 

factor was used to reduce the quantities of short-lived radionuclides for critical ity accidents. 

In calculating external dose due to clouds, the CRAC2 code performs the 

calculation for a finite-size cloud rather than an infinite-size .cloud. The , 

difference between the two can be substantial at distances as close as 100 m for 

stable atmospheric conditions. For example, failure to correct for this factor 

at 100 m from a point release (i.e., no building wake) during class F stability 

would cause external doses to be overestimated by a factor of almost 40. However, the finite cloud correction factor is much smaller when building wake 

factors are used. " -'..- ' 

Jhe results of the atmospheric dispersion calculations for inhalatiorvare 

shown in Figure 1 for both F, 1 m/s and D, 4.5 m/s assumptions. Figure 1,-

giving x/Q in s/m3, can be used to calculate inhalation dose D in rems due to 

a released quantity Q in uCi by using the equation: 

D = DCF x B x x/Q x Q 

where: DCF = dose conversion factor, rems/uCi inhaled, as given in 

Table 13 and _•"." . 

B =' breathing rate", which is 2.66 x 10-4 m3/s. 

The doses due to ground-shine and cloud-shine should be added to the 

inhalation dose to obtain a person's total dose. As a practical matter, however 

ground-shine and cloud-shine doses will be considerably smaller than the inhalation dose except for a few radionuclides (xenon, krypton, Na-24, Mn-56, Tc-99m, 

and Ru-105). 

10 
P.21 

i 

600 800 

Distance, Meters 

Figure 1. Atmospheric dispersion versus distance. 

1400 

11 
P.22 

2.1.4 Protective Action Guides 

Protective action guides are expressed in terms of projected doses to 

individuals in the population which warrant taking protective action. The EPA 

has published draft guides for taking protective action in order to avoidexposure to radiation as the result of an accident at a nuclear power plant.* The. 

.EPA recommends that protective actions should be considered by responsible 

Officials if projected whole body doses are in the range of 1 to .5 rems. The 

loWer dose of 1 rem is a level which "should be used if there are no major 

local constraints in providing protection at that level, especially to sensitive populations" (children and pregnant women). The EPA believes that "in no 

case should the higher value (5 rems) be exceeded ih determining the need for 

protective action." Put another way, protective actions may be considered 

optional at 1 rem, to be taken if readily feasible, but are highly recommended 

at 5 rems if at all feasible. Note that the 1-rem and 5-rem doses are projected 

doses that might occur after the protective action decision. Doses received 

prior to the decisionmaking and during the protective action implementation 

time are nbt^considered in the decisionmaking. Only dose savings as a result 

of taking a protective action are to be used in "determining whether such protective action is warrented. 

For radioactive materials that deliver dose to the body nonuniformly after 

they are inhaled, the resulting effective dose equivalent is compared to the 

EPA's protective action guides. The effective dose equivalent, as defined in 

the previous section, is the sum of the external gamma dose equivalent, the 

dose equivalent delivered to each body organ multiplied by a weighting factor 

for the organ from ICRP Publication 26,_ and the external beta dose equivalent 

delivered to the skin multiplied by the weighting factor from ICRP Publication 28. 

The factors used to convert intake of radionuclides to an effective dose equivalent are.taken from ICRP Publication 30. 

The Commission's policy on the use of the EPA's protective action guides 

to establish planning zones for nuclear power plants is stated in NUREG-0654.** 

*EPA-520/l-75-001, "Manual of Protective Action Guides and Protective Actions 

for Nuclear Incidents," Draft Revision of June, 1980. 

**NUREG-0654/FEMA-REP-l, Revision 1, "Criteria for Preparation and Evaluation 

of Radiological Emergency Response Plans and Preparedness in Support of 

Nuclear Power Plants," joint NRC/FEMA Report, 1980, page 12. 

12 
P.23 

"The size (about 10 miles radius) of the plume exposure emergency planning 

zone was based primarily on the following considerations: 

a. projected doses from the traditional design basis accidents would 

not exceed protective action guide levels outside the zone; 

b. 

c. 

•er.-

projected doses from most core melt sequences would not exceed 

protective action guide levels outside the zone; 

for the worst core melt sequences, immediate life threatening doses 

would generally not occur outside the zone; 

detailed planning within 10 miles would provide a substantial base 

for expansion of response efforts in the event that this proved 

necessary." _[•• 

( 

For nuclear power plants the tradition design basis accident assumes the 

containment does not fail. In other words, one major barrier to a radioactive 

release remains.. Failure of all engineered safeguards is evaluated with respect 

to life threatening doses rather than protective action guides in the 1 to 5 rem 

range. In addition, emergency preparedness does not guarantee there will be no 

loss of life for the worst case imaginable. 

A later report elaborates on the Commission"'-policy.-'• NURE6--0771* states:. 

"For the small releases, the lower ranges of the protective action 

guides (PAG) would be used as the appropriate measure to base planning. For somewhat larger releases, the criteria.shift to the upper 

ranges of the PAG and levels of exposures which Would still be less 

than medically detectable. For intermediate level releasesearly 

injuries would be used as the measure to base the EPZ distance judgment on. Finally, for the most severe accidents, early fatalities 

become the immediate concern and therefore the measure to base the 

criteria upon..." 

*NUREG-0771, "Regulatory Impact of Nuclear Reactor Accident Source Term 

Assumptions," for comment, June, 1981, page 35. 

13 
For application to fuel cycle and byproduct material licensees, the lower 

end of the range (1 rem) of the EPA's protective action guides is used in conjunction with calculations of releases and offsite radiation doses due to severe 

accidents, such as a major facility fire, to establish the need for a plan, 

thus the lower range of the protective action guides is used to determine the 

need for offsite emergency preparedness. . 

The actual assumptions that were used for each facility type are discussed 

in separate sections in the remainder of this report. The reasons for selecting 

the assumptions are also discussed. 

2.1.5 . A Discussion of the Conservatism in the Calculations 

The Commission's policy is that, "Emergency planning should be. based on 

realistic assumptions regarding severe accidents."* 

The.doses calculated in this Regulatory Analysis have been conservatively 

calculated. Doses to people near a plant experiencing a severe accident are 

likely to be far below the doses in this analysis, probably by an order of "mag- ~ 

nitude or more, except in-very unusual circumstances. The accident history of -

such facilities in the U.S. is that there is no known case of a member of the 

public receiving even as much as 1% of the doses calculated in this analysis 

as the result of an accidental airborne release from any nonreactor facility.** 

A number of factors which cause this analysis to be conservative are discussed 

below. 

*"U.S. Nuclear Regulatory Commission Policy and Planning Guidance.- 1985," 

U.S." Nuclear Regulatory Commission, NUREG-0885, Issue 4, 1985, page 6. /'"'-

**For a 1962 release of high-enriched UF6 from the NFS "plant, Erwin, Tennessee, 

a plume centerline dose equal to 4% of the 1-rem effective dose equivalent 

guide was calculated using conservative assumptions (no deposition, open 

field diffusion parameters, no wind direction shift, etc.) However, the 

report stated, "No specific information regarding the presence of individuals 

during the releases was available." Because ho one is known to have stood 

on or near the plume centerline, we can say there are no known exposures 

exceeding 1% of 1-rem. The dose calculations are contained in an unpublished 

report, "Dose Assessment of Airborne Releases from NFS-Erwin Fuel Facility -

1972-1981," U.S. Nuclear Regulatory Commission, Region II, Atlanta, Georgia, 

May, 1983. 

14 
2.1.5.1 Conservative Factors 

1. Entire possession limit assumed to be involved. In calculating the 

quantities of radioactive material for which an emergency plan would be needed, 

this analysis generally assumed that the licensee's entire or nearly entire.... 

possession limit would be involved. In actuality, most licensees at any 

particular, time possess much less material than they are legally authorized 

to possess. In many cases the possessed material will be located at different 

locations and will not all be subject to release during a particular accident. 

For example, the National Institutes of Health is authorized.to use and store 

licensed material in more than 1,000 different laboratories. -

2. Worst-case release fractions. The release fractions due to fires 

(the accidents with highest potential release) were determined from experiments 

designed to maximize releases. In such experiments a finely powdered material 

is typically placed on top of a large amount of combustible material. Having 

the entire licensed inventory unenclosed on top of a large quantity of combustible material would be most unusual. Radioactive materials are usually within 

shielded "pigs" and kept in metal safes or well shielded hot cells or glove "_ .'. 

boxes. Amounts of combustible materials present are generally kept low.. 

3. No credit for engineered safeguards or response efforts. No credit . . 

is .generally given for design or operating features that could reduce releases. 

No credit is given for sprinkler systems designed to stop fires. Generally, no 

credit is given for filter systems during a fire. No credit rs-given for. fire 

fighting efforts to stop the fire before it reaches radioactive materials^ 

Little or no credit is given for holding up the release of the material by means, 

of depositionor plateout. For UF6 releases outdoors, no credit is given for 

knocking the uranium out of the air using fire hoses. 

4. - The exposed individual makes no response. In the case of fires and 

UF6 releases, the dose is calculated for a person who stands directly on the 

plume centerline for 30 minutes. Such a person would be standing in dense smoke 

pr irritating acid fumes. Realistically, people can be expected to move from 

such positions to avoid smoke inhalation. People close in would only have to 

move about 20 meters to get completely out of the plume. 

5. No plume-rise for smoke. Even where the assumed accident is a large 

fire no credit is given for plume rise due to buoyancy in calculating the 

quantities of radioactive material for which an emergency plan would be needed. 

The smoke is assumed to be released at and remain at ground level. 

15 
P.26 

6. Conservative dosimetry. The material is assumed to have the solubility which would result in the highest dose per curie inhaled. Particulates are 

generally assumed to have a size of 1 micron making them highly respirable and 

transportable. 

7. Adverse meteorology. Quantities of radioactive material for which 

an emergency plan would be needed were calculated for atmospheric stability 

class F with a 1 m/s windspeed. These conditions result in minimal dilution 

and high plume centerline doses, but also very narrow plumes. It is probable 

that the actual weather would cause lower doses. For example, doses during a 

moderately sunny day with average winds would be a.factor of 50 times smaller 

than the doses calculated for the analysis. 

8. Open-field site assumed. A rural open-field site is assumed. Greater 

atmospheric dispersion ahd-thus lower doses would occur at an urban or suburban 

site. Buildings, trees, or other obstacles in the plume path would broaden the 

plume. Heat sources would increase the plume height. 

9. No wind shifts. Doses are calculated only on the plume centerline. 

It is assumed that no wind direction shifts occur during the accident. In addition^ correction factors for plume meander are conservative;"the factors were 

selected to envelope the experimental data. Normally greater plume meander 

would be expected. - . ." • 

10. 8-hour criticality. The source term assumes a pulsating criticality. 

with a total of 48 bursts occurring over 8 hours (see Section 2.2.5.2). This 

is a highly conservative source term. 

11. There may be no one standing on the plume centerline. The doses are 

calculated for single point, and they fall off rapidly as one moves away from 

the point. Even with no protective actions, the highest dose anyone would 

-receive is likely, to be well below the assumed dose. 

2.1.5.2 Nonconservative Factors 

On the other hand there are certain assumptions in the dose calculations 

that may be nonconservative in certain instances. These factors are discussed 

below: 

1. Adult doses. Doses are calculated for adults rather than children 

(except for radioiodine doses which are calculated for children). This is 

because dose conversion factors for children using modern dosimeter models are 

generally not available. For some inhaled radionuclides a child standing in 

16 
siv 

KA nlufne may perhaps receive a dose 2 or 3 times higher than an adult standing 

a t the same location. 

2 Breathing rates. The breathing rate used in the dose calculations 

£9 66 x IO-4 m3/s) represents an average breathing rate. Breathing rates for 

^ove average activity would be higher. 

3 Site-specific factors not considered. The table of quantities of _'••• 

••tcrlal for which emergency planning should be considered was based on assumptions (for example building wake) that would usually be conservative, but may 

not be conservative for all instances. For example, the building wake factor 

for * particular building could be less than assumed although it would generally 

b« larger. This should be a minor factor. Any increases in dose due to such 

factors would not be significant in size by comparison with the sizes of the 

conservatisms discussed above. 

2.2 Fuel Cycle Facilities 

-.2.2.1 Uranium Mining" 

Uranium mining is not considered in this-report because the NRC has no 

regulatory jurisdiction over uranium mining. Uranium mining is regulated 

Instead by the Mine Safety and Health Administration, the Environmental Protection Agency, and the individual states. 

2.2.2 Uranium Milling 

Uranium mills extract uranium from ore that typically averages about 1 part 

par 1000 uranium. The mills produce concentrated uranium compounds, which are 

shipped out in 55. gallon drums, and waste "tailings," which contain radium-226 

And thorium-23Q not removed from the ore by the mill processes. In late 1984 

there were about 10 full-scale uranium mills operating in the U.S. In addition, 

there are smaller facilities that perform some of the processes found in milling, 

toughly half the mills are licensed by the NRC. The others are licensed by 

Agreement States. 

In addition, this section considers "in-situ" solution uranium mining, 

in which a solution that has leached uranium from the ground is pumped up and 

uranium extracted from the solution. 

17 
P.28 

2.2.2.1 Accident History 

i major fires ai 

or -taiVings li.i 

Uranium mills have experienced major fires and accidental releases from 

tailings ponds due to dam failures-or-taiVings .line-breaks.• 

Uranium-mills have a potential for Targe fires because large quantities 

of kerosene are used as a solvent to extract uranium in one mill process. The 

kerosene contains dissolved uranium and is contained in large open tanks. Two 

large fires in solvent extraction circuits have occurred. Table 1 lists fires 

known to have occurred. It is notable that the fires that have occurred have 

caused little release of radioactive material even though two of the fires 

were very intense.

 r 

Aside from fires, the other notable type of accident at.uranium mills 

has been tailings pond releas.es. There have been at least 16 instances where 

uranium mill tailings solids and liquids were released from tailings impoundments. Table 2 describes these releases. In no instances were there radiation 

dose rates that would cause doses to the public in the range of the EPA's 

protective action guides. In nojcase was drinking water contaminated above, :" 

NRC limits (Appendix B of 10 CFR Part 20).-"-'. . ~ - . _ 

2.2.2.2 AccidentrSource Terms 

Potential releases.of radioactive.materials and potential doses that 

could result from accidents at uranium mills are shown in Table 3. The largest 

potential releases were determined to occur as af.result of: (1) fires, 

(2) undetected failures of air cleaning -systems, _(3) tailing pond releases, 

and (4) tornadoes. Some reported calculations, of the quantities released and 

projected offsite doses are shown in Table 3. "." 

These calculations show that the largest offsite radiation doses would be 

due to a fire in the solvent extraction circuit or an undetected failure of 

the "air cleaning system servicing the yellowcake drying area. However, the 

undetected failure of'the air cleaning system servicing the yellowcake drying 

area does not provide a basis for actions to protect the public. As long as 

the failure remains undetected no emergency plan can be activated and no protective actions can be taken. As soon as the failure is detected, the release 

can be stopped by turning off the ventilation blowers. Airborne concentrations 

18 
Table 1. Fires in Uranium Mills 

P.29 

•BE 

Date Mill 

Fire 

Description Offsite release 

3-19-59 Vanadium Corp., 

of America, 

Durango, CO 

6-25-65 American Metal, 

Grand Junction, 

CO 

2-68 Western Nuclear, 

Jeffery City.iWY 

11-10-68 Petrotomics Co., 

Shirley Basin, WY 

12-25-68 _ Atlas Corp., 

Moab, UT 

10-23-80 Minerals Exploration, 

- Sweetwater, WY 

1-2-81 Atlas Corp., 

Moab, UT 

Fire in yellowcake 

dryer 

Fire in ore 

dryer for 3-5 min. 

$2600 damage 

Workers started a 

fire to thaw a frozen 

ore dryer. Fire 

ignited propane from 

a leaking tank. 

Solvent extraction 

circuit. 

$300,000 damage 

Solvent extraction. : 

circuit. Cause 

unknown. 

$1,000,000 damage -

Major fire burned, 

in mill before it 

started operation 

Fire in yellowcake 

scrubber stack for 

15 min 

None detected 

None detected 

None detected 

None detected 

None detected 

None. Radioactive, 

material was 

not yet being . 

processed. 

None detected 

19 
P.30 

Table 2. Uranium Mill Tailings Releases, 1959-1984 

Date Mill Type of Incident Release 

8-19-59 Union Carbide 

Green River, UT 

Tailings dam washed ~ 15,000 T sands lost to 

out river in flash floods; no 

increase in dissolving Ra 

was noted in river. 

8-22-60 Kerr-McGee 

Shiprock, NM 

12-6-61 Union Carbide 

Maybell,: CO -

6-11-62 

8-17^62 

Mines Development, Inc. 

Edgemont, SD : 

Atlas-ZincMinerals 

Mexican Hat, UT 

Raffinate pond 

dike failure 

Tailings dike 

failure 

Tailings dike 

failure 

Slurry pipeline 

rupture 

6-16-63 Utah Construction Precautionary 

- - Riverton^ WY_ release 

240,000 gal of raffinate 

releases into river 

~ 50 x 10-8 uCi/ml Ra-226; 

river samples collected 

several days after release 

showed no increase in 

Ra-226 background. 

~ 500 T solids released 

from tailings area; 200 T 

reached unrestricted area; 

no liquid reached any 

stream. 

200 T solids washed into" 

creek and some carried 

25 mi into reservoir. 

Est. 280 T solids + 240 T 

liquids released from 

broken tailings discharge 

line into draw 1.5 mi from 

river. 

Material released by 2 ft 

drainage cut made to prevent 

cresting due to heavy rains; 

material released below 

10 CFR Part 20 values. 

11-17-66. VCA 

Shiprock, NM 

Raffinate line" 

failure 

2-6-67 Atlas Corp. 

Moab, UT 

Line failure 

Est. 16,000 gals of liquid 

lost because of break in 

raffinate line; material 

spread over 1/4 acre; break 

occurred 1 mi from river 

with some small amount 

reaching river. 

440,000 gal lost; average 

Ra-226 concentration was 

was 5.5 x 10-8 mCi/ml. 

20 
MZ 

Table 2. (continued) 

Climax Uranium 

Grand Junction, 

CO 

Type of Incident Release 

Tailings dike 

failure -

Dike failure released 

1-10 acre-ft of waste 

liquid into Colorado River; 

no indication that Ra cone, 

in river exceeded 10 CFR 

Part 20 limits. 

11-23-68 Atlas Corp. 

Moab, UT 

Slurry pipeline 

rupture 

35,000 gal of tailings 

slurry lost; flowed 1/2 

mile to Colorado River; 

most solids settled.out 

in drywash. 

2-16-71 Petrotomics 

Shirley. Basin, 

WY 

J-23-71 Western Nuclear 

Jeffrey City, 

WY 

2-5-77 United NuclearHomestake 

Partners 

Grants, NM 

4-77 Western Nuclear, 

Inc. 

Jeffrey City, WY 

Secondary _tai1i ngs 

dike failure 

Tailings line and 

dike failure 

Slurry pipeline 

rupture 

Failure of tailings 

pond embankment 

2,000 gal of liquid lost 

to unrestricted area; spill 

froze in place. 

Break in slurry line casued 

a dike failure allowing sand 

tails to flow into natural 

basin adjacent to tailings 

site on licensee's property. 

50,000 tons of solids and 

slimes and somewhere between 

2 million and 8 million gal 

of liquid. All material was 

confined to company property. 

~ 2 million gal of liquid 

tailings and 55 yd3 of 

solids were released. No 

material was. released to 

unrestricted areas. 

9-26-77 - United Nuclear 

9-27-77 Church Rock, NM 

Release from 

tailings slurry 

line 

Approximately 1 ton of 

solids and 900 gal of liquid 

entered the watercourse." 

7-16-79 United Nuclear 

Church Rock, NM 

Tailings dike 

failure 

100,000,000 gallons of tailings solution and 1,100 tons 

of tailings solids. Most of 

the solids were deposited 

near the impoundment, but 

much solution reached a 

river. 

•eference: Regulatory Guide 3.11.1, "Operational Inspection and Surveillance 

of Embankment Retention Systems for Uranium Mill Tailings." 

.-•?$«•• 

21 
Table 3. Accident Source Terms and Doses From Uranium Mill Accidents 

Tornado Tailing Pond Release 

Fire 'in Solvent 

Extraction Circuit, 

Reference Release Dose Release Dose Release Dose 

Failure of the Air 

Cleaning System Serving the 

Yellowcake Drying Area 

Release Dose 

GEIS 

Sand Rock 

DES 

This 

Report 

11,400 kg U total < 2.2" x 10-7 rem 1400 tons solid Small. Cleanup < 13 kg U 

< 11,400 kg U 

respirable 

4550 kg U total 

< 4550 kg U . 

respirable 

to lungs at 500m 

< 1.1 x 10-7 rem 

at 400,0m 

(max. dose) , ,' 

14,000,000 gal:, 

liquids 

Same as GEIS 

assumed1 < 0.65 kg ttiqrlum* 

< 1.1 kg U 

< 1.36 rem* 

to bone at 

500 m 

11 kg Insoluble 86 mrem 

U oxides to lung 

lOver 8 hours 

rem 

1.3 kg U 

IO-7 

to bone 

at 8000 m 

(nearest 

residence) 

0.01 to 

0.1 rem EDE 

12 kg insoluble 

U oxides over 

8 hours 

at 2000 m 

10-2 rem 

to lung at 

8000 m 

(nearest 

residence) 

References •''.: y ,;;VV ' 

'••',' • ' " • • • ' ' ' > •<••• . 

GEIS: "Final Generic Environmental Impact Statement on Uranium Milling," NUREG-0706, Volume 1, pp 7-1 to 7-20, September, 1980. 

Sand Rock DES: "Draft Environmental Statement Related to the Operation of Sand Rocks Mill Project," NUREG-0889', pages 5-1 to 

5-12, March, 1982'..',. -,,.•' ,' •''''•.,-"','•"• " ,-' 

.'••.' - ' ' . •' .'••' ' i • . , 

*The dose value from GEIS is in error. The solvent extraction was assumed to contain 5% as much ThV230 as uranium by weight. The value should 

have been b% by^ activity. This error causes the dose to be overestimated'by a factor of about 50,000 times. 
£&y 

\«ty** >. -tix/p material promptly drop to very low levels. External exposure -0/ raflioaci.i vc 

"Haible because uranium is a very weak gamma emitter. 

Tbrnadoes could release a larger amount of radioactive material. However, 

iWad the material so greatly that resulting doses are very small,* as 

in Table 3. Because the doses that would be caused by tornadoes are so 

Her-than doses from other accidents,, releases and doses due to tornatMwn 

mch s»a 

dee* »re 

not discussed further in this analyses. 

Tailings pond failures also release a large quantity of material. However, 

•it* dose rates are less than 0.1 mR/hr and radioactive material concentrations 

so low that prompt emergency action is not needed to prevent anyone's dose 

froa exceeding the EPA's protective action guides. 

Thus we conclude that a fire in the solvent extraction circuit is:the 

accident of greatest significance for emergency preparedness. We assume the 

release from the building is 1.3 kg of uranium. The 1.3 kg release is based 

pa 0. 1% ofthe material in process becoming airborne and escaping from the 

building. Experiments on releases of uranium in a kerosene fire showed average 

releases of 0.025% when the residue is not heated with a propane torch after 

dryness-occurs.** Due to licensing policy requiring.automatic fire detection 

and supression systems*** (such as automatic, sprinklers., foam, or halon systems) 

significant heating beyond dryness would not be expected. Other experiments 

ihowed similarly low releases.! The uranium would be in insoluble form 

(solubility class Y) because a large kerosene fire would produce temperatures 

exceeding 400°C, the temperature at which the uranium should form insoluble 

oxides.**** The uranium may be class Y if the fire completely oxidizes the 

*NUREG-0706, Volume 1, "Final Generic Environmental Impact Statement.on ~~ 

Uranium Milling," September, 1980, p. 7-4. 

**Jofu Mishima and Lyle Schendiman, "Interim Report: The Fractional Airborne 

Release of Dissolved Radioactive Materials During the Combustion of 30 Percent 

Normal Tributyl Phosphate in a Kerosene-Type Diluent," BNWL-B-274, Pacific 

Northwest Laboratory, 1973. 

•**Regulatory Guide 8.31, "Information Relevant to Ensuring that Occupational 

Radiation Exposures at Uranium Mills Will Be as Low as Is Reasonably 

Achievable," Section C.3.4. 

tO. Whitney Tharin, Jr., "Burning of Radioactive Process Solvent," Savannah 

River Laboratory Report DP-942, Aiken, South Carolina, 1965. 

****R. C. Merritt, The Extractive Metallurgy of Uranium, Colorado School of 

Mines Research Institute, pp. 252-4, 1971. 

'&-' • 

23 
I . 

soluble class D uranium in the solvent extraction tanks and converts it to 

insoluble uranium. 

This source term ojf 1.3 kg of uranium is also considered to be suitable 

for "in-situ" solution wining. Some "in situ" mining processes use solvent 

extraction processes similar to those in uranium mills. More severe accidents 

than described above were not identified for "in-situ" mining. 

2.2.2.3 Calculations of Doses 

A 1.3 kg release of natural uranium, due ta a fire at an uranium mill could 

result in a potential dose of 0.1 rem effective dose equivalent during adverse 

weather (F, 1 m/s) or 0.01 rem during typical weather (D, 4.5 m/s). Assumptions 

were that the building size was 10 m high by 25 m wide, the release height was 

ground level, and the release duration was 30 minutes. The factors "for determining effective dose equivalent from ICRP Publication 30 for a particle size 

of 1 micron AMAD and class Y solubility are: 1.31 x 108 rem/Ci for U-234, 1.21 

x 108 rem/Ci for U-235, and 1.17 x 108 rem/Ci for_Q-238. If 1 curie of Uranium 

is composed of U-234, U-235, and U-238 in their naturally occurring-proportions, 

the dose conversion factor is 1.24 x 108 rem/Ci. - . - .-

The calculated dose from this accident is small (0.1 rem. or less).because 

of the very low specific activity of the uranium and the low volatility of the 

uranium compounds, which causes a low release fraction. 

Low release fractions are the reason why no offsite ground contamination 

was ever detected due to the fires listed in Table 1.. - . 7-

2.2.2.4 Implications for Emergency Preparedness 

On the basis of the very low doses calculated, the staff concludes that 

there is no need for offsite emergency protective actions on the part of the 

public at uranium mills or for "in-situ" mining. 

The staff concludes that no credible accident would justify emergency 

protective actions because radiation doses to the public offsite from an accident would be below the EPA's protective action guides. Also, the quantity of 

uranium inhaled is below the quantity where chemical toxicity effects are 

24 
Fi-li 

«rf * Thus, neither radiation doses nor chemical toxicity from licensed 

x i* -ic a concern with respect to the need for prompt protective actions. 

In the event of such a fire, the licensee would be required by existing 

__ -filiations to take certain actions. Among these, the licensee would be 

_Milred b y §26.201(b) to conduct surveys (offsite if appropriate) to determine 

Mt^_r tj,e NRC's limits on radioactivity in effluents to unrestricted areas 

120 106 were exceeded. A major fire would also require immediate notification of NRC by telephone and telegraph (§20.403). If appropriate, the NRC 

etm\d elect to immediately send an inspector to the site to make any necessary 

radiation measurements or evaluate the situation. 

With respect to tailing dam failures, rapid emergency response is not -

~M«ded to-avoid doses exceeding protection action guides because dose rates 

at • spill site are very low. An appropriate response is' to monitor drinking 

water especially'for radium-226, to be sure that drinking water standards-are 

M t . Gamma ray monitoring of the ground is also appropriate to determine where 

the tailings have been deposited. However, ground contamination presents little 

tjBMdiate hazard to the public because the gamma dose rates are low. Gamma doseretes in contact.with tailings should be less thafi 0.1 mR/hr. Since the EPA's 

protective action guides would not be exceeded, a rapid emergency response is 

= hot needed. A clean-up of the spilled tailings would be expected, but this 

could be done effectively without preexisting emergency preparedness. 

2.2.3 UFR Conversion Plants 

.Conversion plants convert yellowcake shipped from uranium mills into 

iiu« hexafluoride (UF6). Heated liquid UF6 is put into 10-ton or 14-ton~ " 

-. Qflinders, The cylinders are cooled for several days until the UF6 solidifies. 

eventually, the filled cylinders are shipped to enrichment plants to enrich 

the uranium in U-235. There are two NRC-licensed conversion plants: Kerr-McGee 

In Oklahoma and Allied Chemical in Illinois. 

The uranium is handled in many different chemical forms in UF6 conversion 

plants, but the UF6 itself is the only chemical form of uranium that is readily 

dlspersible. For example, the dispersibility of yellowcake is essentially the 

•R. A. Just and V. S. Emler, "Generic Report of Health Effects for the U.S. 

6eseous Diffusion Plants," DOE Report K/D 5050, Section VIII, Part 1, 

Pege 6, 1984. 

• % & 25 
same as that of yellowcake at uranium mills. Accidents involving yellowcake 

were previously discussed and found not to require offsite emergency 

preparedness. j 

The release of UF6 in significant quantity is possible because UF6 is 

volatile above room temperature. The UF6 released will react with water in" 

the air as follows:'" -

UF6;+ 2H20 = U02F2 + 4HF + 52.2 kcal/mole* 

The U02F2 forms a particulate, very soluble in the lungs, which will -be carried 

away by wind and will settle onto the ground. The HF is "a corrosive acid Vapor 

that can severely harm the lungs if sufficiently concentrated. The release of 

1 kg of UF6 combining with 0.1. kg of water results in release of 0.88 kg of 

U02F2 (which contains 0.68 kg of uranium) and 0.23 kg of HF. 

2.2.3.1 Acci-dent History 

Table 4 lists significant releases of UF6 that have occurred from all types 

of facilities, not iust conversion plants. There have been many releases of-U_F6. 

The releases have caused at least two prompt fatalities and several injuries. .'.'. 

The significant UF6 releases have consistently been with UF6 heated above its \ 

melting point (65°C). The releases have generally been fairly rapid—lasting 

from a few minutes to an hour. The plumes, where they are highly concentrated, ... 

have been visible and immediately irritating to the lungs. The escape of UF6 

can be-diminished greatly if the leak can be sprayed with water. 

Inhalation~of uranium due to a UF6 release can be verified by measurements 

of uranium concentrations in urinetaken within 48 hours of the exposure. The 

uranium from UFg has a biological half-life for expulsion via the urine of 4 to" 

6 hours.** Workers exposed tohigh concentrations have suffered edema of the 

lungs, presumably from exposure to HF, and kidney damage due to heavy metal 

*Minton Kelly, Oak Ridge National Laboratory, Sept. 1983. 

**M.W. Babcock and R. C. Heatherton, "Bioassay Aspects of a UF6 Fume Release," 

Proceedings of the 12th Annual Bio-Assay and Analytical Chemistry Meeting, 

AEC Report C0NF-661018, 1966, pp 147-159. 

26 
Table 4. Accidents Involving UF6 Releases 

Type of 

facility 

Quantity of 

UF6 released Cause and consequenc i e 

J949- AEC facilit y Experimental 

laboratory 

,

;,$$§y 

J949 AEC facilit y Not 

identifie d 

5-10-60 Babcock'& 

Wilcox, 

Apollo, PA 

Fuel 

fabrication 

11-17-60 _HJ'nioh Carbide, Uranium. -

- Oak Ridge, TN enrichment 

5-25-62 iluclear Fuel 

Services, 

- Erwih, TN 

yJ- 20- 64 

f-14-66 

Nuclear Fuel 

Services, 

Erwin, TN 

National 

Lead, 

Fernald, OH . 

Fuel 

fabrication 

metal 

Fuel 

fabrication 

metal 

- Feed -- -

material -

production. 

$-29-67 

7-19-68 

Kerr-McGee, 

Gore, OK 

Kerr-McGee 

Crescent, OK 

UF6 

conversion 

Fuel 

fabrication 

:':@v: 

Very large 

amount accompanied with 

live steam 

Believed to 

be 13 kg 

Not 

reported 

Not 

reported 

15 kg HEU in 

5 min. 6 kg 

recovered in 

plant 

1 kg in 2 hrs. 

Half recovered 

onsite 

2300 kg in 

1 hr. Much 

absorbed by 

water spray 

45 kg in 

15-20 min 

45 kg of 1.6% 

enriched U 

in 15-20 min 

Rupture or explosion 

of large tank. Two 

workers killed. Three 

other workers seriously 

injured, 13 others 

less seriously injured. 

Sudden leak in a hot 

cylinder. One worker 

received injury to 

respiratory tract, 

eyes, and kidneys. 

Leak in heat exchanger 

allowed U02F2 to escape 

to river water. 60 x 

MPC at discharge point. 

Rupture of 10-ton 

cylinder." 

An overheated 15-kg 

cylinder ruptured and 

released its contents 

in the building. 

Overpressure burst tube 

Operator accidentally 

removed valve on a hot 

10-ton cylinder, developed lung edema, 

hospitalized 6 days. 

No observed injury to 

kidney. 

Gasket leaked due to 

overheating. 

Valve accidentally left 

open during heating. 

27 
Table 4. (continued) 

Date 

11-12-68 

5-2-73 

4-20-74 

Facility 

Allied 

Chemical,. IL 

Goodyear 

Atomic 

Oak Ridge, TN 

Numec, 

Apollo, PA 

Type of 

facility 

UF6 

conversion 

UF 

processing 

Mixed oxide 

fuel 

fabrication 

Quantity of 

UF6 released 

43kg - y 

100 kg in 

20 min 

(inside) 

6 kg, slightly 

enriched 

Cause 

Valve failure 

Worker broke valve on . 

10-ton cylinder. 

'• ' 

12-2-76 Exxon Nuclear, Fuel 

Richland, WA fabrication 

Smal 1 Worker disconnected 

line but had forgotten 

to close valve. 

3-7-78 

12-3-78 

8-7-79 

5-20-80 

9-15-81 

10-12-81 

2-25-82 

12-83 

Portsmouth 

Gaseous 

Diffusion 

Plant, OH 

GE 

NFS, Erwin 

TN -

GE 

GE 

NFS, Erwin, 

TN 

Exxon 

Edlow International, East 

St. Louis, IL 

Enrichment 

plant 

Fuel -

fabrication 

Fuel 

fabrication 

Fuel 

fabrication 

Fuel

 r.-

fabrication 

Fuel, 

fabrication 

Fuel 

fabrication 

Warehouse 

9500 kg 

in 1/2 to 

1 hour 

not known 

<3 kg 

<1 kg 

<<74 kg 

0.05 to 0.1 -

kg, HEU 

« 2 5 kg 

None 

Rupture of dropped hot 

14-ton UF6 cylinder. 

Block valve opened 

Accidental venting of 

cylinder to stack. 

Pipe flange failure 

Gasket leak 

Release via main 

scrubber stack. 

Gasket leak 

Fire in warehouse. . 

__ 

28 
P.39 

•soning from uranium. At least two workers were killed . Persons injured or 

kille d in thi s manner have al l been workers in a room working close to a UF6 

cyliner. 

Two of the cases involving the most serious exposures were reported by 

Howland.* These involved two fatalities, four serious injuries, and slight 

injury to 13 other people. One of the fatalities showed by autopsy roughly 

1000 mg of. uranium:in the lungs. Howland concluded that the most serious 

injuries (observed on the skin, eye, mucous membrane of the upper respiratory 

tract, esophagus, larynx, and bronchi) were all caused by the action of the 

fluoride ion on the exposed tissues. Uranium produced transient urinary-tract 

changes. The conclusion is"that HF and uranium both have adverse, effects, but 

that the HF effects are the more severe. -

In the National Lead-Fernald accident, one worker suffered lung edema, 

presumably from exposure to HF.** - No injury to his kidneys was observed. He 

excreted in urine over 1 mg of uranium in the first two" days after the accident, suggesting a total intake of roughly 2 to 3 mg of uranium. 

The largest release of UF6 was the 1978 accident at the Portsmouth, -Ohio 

gaseous diffusion plant. In this accident a heatedythin-walled cylinder con-. 

taining 14 (short) tons of natural UF6 was dropped 8 to 10 inches and ruptured 

below the liquid level.*** Within-one hour, about 9500 kg of UF6 escaped. 

This is equal to about 6400 kg of uranium. The release was outdoors. The air 

temperature was 32°F, the wind speed.Was 2 meters/sec, and a mixture of snow 

and freezing rain was falling. Snow covered the grbund. About 550 kg of 

uranium were recovered on the ground afterwards. Agglomeration is likely to 

have increased the settling. About 4800T kg of uranium (75% ofthe release) 

were estimated to have become airborne and-dissipat_ed in" the air, much thereby 

leaving the-site. The site boundary in the downwind direction was at a 

distance of 2.2 km. . """." " 

*Joe W. Howland, "Studies On Human Exposures to Uranium Compounds," in 

Pharmacology and Toxicology of Uranium Compounds,-edited by Carl Voegtlin 

and Harold Hodge, McGraw-Hill, New York, 1949, p 993. 

**Babcock and Heatherton, op. cit. 

***"Investigation of Occurrence Involving Release of Uranium Hexafluoride from 

a Fourteen-Ton Cylinder at the Portsmouth Gaseous Diffusion Plant on March 7, 

1978," DOE Report ORO-757, June, 1978. 

29 
P.40 

Water samples from a drainage ditch located near the release had a peak 

uranium concentration of 450 mg/l, 10 times the NRC's radiological limit for 

water to be released to unrestricted areas.* 

The reported environmental effects were minimal. Workers who drove 

through"the plume showed no detectable uranium in samples of their urine. 

Significant ground and water contamination were confined to distances of a 

few hundred yards from the release point. Airborne concentrations at the site 

boundary (2.2 knv) were calculated to be not high enough to be harmful for brief 

exposures. 

Another large release of U.F6 occurred in France in. 1977.t As the result 

of a handling error a valve ruptured on a container heated tor90-95°C- The 

UF6 immediately started to spill out onto the ground. The liquid flow lasted .: 

10 to 15 minutes until the level of liquid in the container had fallen belowv 

the valve opening. Then UF6 continued escaping as a gas until the valve was 

plugged with a wooden peg 30 minutes after the rupture. Of the 8800 kg of 

liquid UFe in the container, 7100 kg escaped. --'• 

Water "and Carbon dioxide were used to prevent the escaped UF6 "from becoming 

airborne. However, 330 kg of uranium and 1600 kg of HF were not recovered. 

-Thus 7% of the uranium and 98% of the HF that escaped the container apparently 

became airborne. "Weather conditions favored rapid dilution. It was a warm 

and sunny afternoon with a windspeed of 9 m/s. 

The French workplace limit for HF of 2.4 mg/m3 was exceeded up to a 

distance of 1200 meters. Ground contamination by uranium of up to 10 mg/m2 

was observed up to 600 meters. The area on which virtually "all the^solid 

uranium compounds settled did not exceed 1000 m2. 

-No injuries were observed. Urine samples were taken from 449 people.. 

-Two workers excreted more than 0.5 mgduring the first day, but_no physiological symptoms were observed. No symptoms of the HF exposure were observed. 

*the NRC limit for water in unrestriced areas in 10 CFR Part 20, Appendix B 

is 3 x 10-5 microcuries/ml. Using the specific activity of natural uranium 

of 6.77 x 10-7 microcuries/microgram, the effluent water standard is equivalent to 44 milligrams/liter. 

tA.J. Docouret, "An Experience of Accidental Release of UF6," Comurtex Plant, 

Pierrelatte, France. 

30 
P.41 

U ^ S 

gf« " 

addition to gaseous UF6 releases, conversion plants have released 

to rivers. On Dec. 1, 1978 the Kerr-McGee conversion plant accidental" 

•leased 750 ,kg of natural uranium in the form of uranyl nitrate into a 

rlvtr 

for 

The liquid released had a uranium concentration of "1.4 times.the MPC 

... which would then be diluted by the river water. 

w/ti1-' » 

2.2.-3.2 Accident Source Terms 

The NRC staff, Sutter at Pacific Northwest Laboratory, and M. Simon-Toy* 

n*k Ridge National Laboratory have-recently_analyzed potential accidents at 

* mnversion plants to estimate potential releases of JUF6.-

Uf«

 c 

The largest release postulated by the NRC staff is contained in an Environta 1 Impact Appraisal for the Allied Chemical conversion plant.** The NRC 

•aff assumed that the largest release of UF6 would be caused by the rupture of 

heated 14-ton cylinder. The staff assumed that 9500 kg of UF6. would escape,. 

-nd that the material would hydrolyze. As a result, 4800 kg of natural uranium 

would be released with the chemical form U02 F2,a highly soluble compound.. 

Sutter*** considered a number of possible accidents. These include: 

1. The rupture of two 14-ton UF6 cylinders outdoors in conjunction with 

a fire fed by 100 gallons of gasoline due to a truck crash 

2. A leak of UF6 from a pipe 

3. A tornado strike 

4. Fires _-_--"-. -.. y 

5 / Chemical explosions - : 

"-. 6.~ Natural gas explosions -"•"' 

The accident determined by Sutter to cause the most significant release -

1s the rupture of two 14-toh UF6 Cylinders along with a gasoline fire. The 

initiation is assumed to be a truck accident in which thevtruck hits the 

*M. Simon-Tov et al., "Scenarios and Analytical Methods for UF6 Releases 

at NRC-Licensed Fuel Cycle Facilities," NUREG/CR-3139, 1984. 

""Environmental Impact Appraisal for Renewal of Source Material License, 

No. SUB-526, Allied Chemical Company UFR Conversion Plant, Office of 

Nuclear Material Safety and Safeguards, NUREG-1071, May, 1984, page 4-28. 

***S.L. Sutter, et al., op. cit. 

31 
mmm^am^mmm_ 

cylinders, ruptures its gas tank, and catches on fire. A total release of 

up to 3800 kg of UF6 was calculated. The amount of material that could be 

released is limited by the amount of heat available to vaporize the solid UF6. 

Heat required to raise the temperature of the cylinder and UF6 is neglected. 

If the UF6 cylinder is not.ruptured, the heat is sufficient to raise the 

temperature of the.UF6.from 20 to 100°F. The pressure produced would not be 

enough to rupture the cylinder. 

Simon-Tov's work was directed toward determining accident scenarios, and 

analysis methods for UF6 releases. His work is the most recent and most 

comprehensive. Twenty-five release scenarios are described in his report 

(Chapter 5>. The scenario most appropriate for this analysis is the rupture 

of a heated liquid^filled cylinder outdoors. At a temperature of 100°C, 

57% of the liquid UF6 could be vaporized. At 120°C, 65% could be vaporized 

(Figure 11, page-58). The most important parameter for determining the release 

is the temperature of the cylinder. Thusvthe largest release is from a cylinder 

just-filled. Analyses of plausibile fire scenarios involving cooled cylinders^ 

show that"the UF6 cannot be heated sufficiently to cause as large a release as. 

from a hot cylinder. "_ -

For the purpose-of this regulatory analysis, the release to be evaluated 

for UF& conversion plants will be one similar to the one that occurred at the 

Portsmouth gaseous diffusion plant. That accident involved the rupture of a 

hot 14-ton UF6 cylinder outdoors with a release of 9500 kg of UF6 (equivalent 

to 6400 kg of natural uranium). It is assumed that 4800 kg of natural uranium 

becomes airborne and the remainder settles on the ground due to agglomeration 

.and impaction"- _The release was calculated by W. Reid Williams as likely to 

occur in about 15 minutes. There would be no advance.warning. Because the

 : 

"release is assumed to be outdoors, no automatic detection or alarm system 

would detect the release. Rather, plant personnel are assumed to detect the 

release arid then take emergency measures. 

The plume would be readily.detectable to the; human senses because of the 

HF and its resulting irritation. Therefore no monitoring instruments are 

needed to detect high concentrations. 

32 
2.2.3.3 Calculations of Doses 

The release of UF6 presents a chemical rather thaniradiological hazard. 

Exposures lethal due to uranium chemical toxicity or HF^burns on lung tissue 

would not result in radiation doses exceeding 1 rem effective dose equivalent. 

Therefore, radiation, doses are not calculated. The release assumed-is.the 

escape of 9500 kg of UF6 in 15 minutes due to the rupture outdoors of a heated 

14-ton cylinder. The mass of uranium in 9500 kg of UF6 is 6400 kg. Some of 

the uranium will be removed from the air initially by agglomeration and impaction. We "assume 4800 kg of uranium becomes airborne. The corresponding mass 

of HF is 1620.kg. " 

Intakes are calculated for atmospheric stability class F with a wind speed 

of 1 m/s as well as stability Class D with wind speed of 4.5 m/s. The plume is 

assumed initially to have a centerline near ground level. The heat from the 

chemical reaction of UF6 combining with the moisture in the air will cause the 

plume to become buoyant. Calculations by W. Reid Williams indicate the plume 

would lift off within 20 to 30 meters and a plume centerline height of about 

20 meters would be obtained within 200 to 300 meters. However, we do nc»t consider Williams' model to be sufficiently developed at this time (May, 1985) to 

use in this Regulatory Analysis exclusively. Therefore, we present a range of 

intakes, assuming the plume is neutrally buoyant for one end of the range and 

assuming a 20-meter plume centerline height (but not building wake) for the 

other end of the range. 

The equation for uranium intake lis: 

I = Q x B x'= 

where Q = the released quantity (4800 kg), 

B = the breathing rate (2.66.x 10-4 m3 /s), and 

x/Q = the atmospheric dispersion value from Figure 1. 

33 
P.44. 

NRC FORM U S 

NRCM 110}. 

U l NUCHA * REGULATORY COMMISSION 

BIBLIOGRAPHIC OATA SHEET 

SIC INSTRUCTIONS ON TM t REVERSE 

Rt'OR T NUMBER tA„.t*rOt>, HOC tat) Vo' No. .1 tnri 

NUREG-1140 Draft 

1 TlTLC ANO SU8TITL I i LEAVE BLANK 

A Regulatory Analysis on Emergency Preparedness for 

Fuel Cycle and Other Radioactive Material Licensees 

Draft Report for Comment 

4 DATE REPORT COMPLETCO 

9 AUTHORISI 

MONT H I YEAR 

May ' 1985 

Stephen A. McGuire 

6 OATE REPORT ISSUED 

- MONTH I YEAR 

June ' 1985 

7 PERFORMING ORGANIZATION NAME AND MAILING ADDRESS ttnclt/dt Z*> Ctxitl 

Division of Risk Analysis and Operations 

Office of Nuclear Regulatory Research 

U. S. Nuclear Regulatory Commission 

Washington. D.C . 20555 

8 PROJECT/TASK/WORK UNIT NUMBER 

9 FIN OR GRANT NUMBER 

10 SPONSORING ORGANIZATION NAME AND MAILING AOORESS llnditOf im Cotti i i r TYPE OF REPORT 

Draft 

Same. 

t> PERIOD COVEREO llnclvt.f (ttntl 

12 SUPPLEMENTARY NOTES 

T 3. ABSTRACT 1200 wotxti ar Itul 

Potential accidents for 15 types of fuel cycle and other radioactive material licensees 

were analyzed. The most potentially hazardous accident, by a large margin, was determined 

to be the sudden rupture of a heated multl-ton cylinder of UFg. Acute fatalities offsite 

probably not credible. Acute permanent Injuries may be possible for many hundreds of are meters, and clinically observable transient effects of unknown long term consequences may be 

possible for distances up to a few miles. These effects would be caused by the chemical 

toxicity of the UF,. Radiation doses would not be significant. 

The most potentially hazardous accident due to radiation exposure was determined to be 

a large fire at certain facilities handling large quantities of alpha-emitting radionuclides 

(I.e., Po-210, Pu-238, Pu-239, Am-241, Cm-242, Cm-244) or rad1o1od1nes (1-125 and 1-131). 

However, acute fatalities or Injuries to people offsite due to accidental releases of these 

materials do not seem plausible. The only other significant accident was Identified as a 

long-term pulsating cr1t1cal1ty at fuel cycle facilities handling h1gh-enr1che<t uranium or 

plutonium . _ . . - _--.-•'..-, - ' - ^ 

An Important feature of the most serious accidents 1s that releases are Ukely to_ start 

without prior waring. The releases would usually end within about half an hour. Thus 

protective actions would have to be taken quickly to be effective. There 1s not Ukely 

to be enough time for dose projections, complicated decisionmaking during the accident, 

or the partlcpatlon of personnel .not 1n the Immediate vicinity of the site. The appropriate 

response by the facility 1s to Immediately notify local fire, police, and other emergency 

personnel and give them a brief predetermined message recommending protective actions. 

Einergency personnel are generally well qualified to respond effectively to small accidents 

of these types, 

14 DOCUMENT ANALVSIS - • KEYWORDS/DESCRIPTORS 

b IDENTlflERS/OPENCNOED TERMS 

XXX. AVAILABILITY 

STATEMENT 

Unlimited 

IB SECURITY CLASSIFICATION 

On'cT'ssified 

ITtitl rtfiorll 

Unclassified 

1? NUMBER OF PAGES 
P.45. 

Uranium intake due to the airborne release of 

4800 kg of uranium 

Distance 

(meters) 

100 

200 

300 

500 

700 

1,000 

1,500 

,2,000 

5,000 

10,000 

15,000 

20,000 

Fv 1 m/s 

buoyant 

0 

6 

46 

110 

110 

92 

62 

44 

11 

3 

1 

0.6 

Uranium intake (mg) 

F, 1 m/s D, 4.5 m/s 

ground buoyant 

4000 6 

2000 53 

1100 59 

530 40 

310 28 

170 17 

80 10 

45 6 

6 1.6 

0.8 0.5 

0.2

 : 0.3 

0.04- 0.2 

D, 4.5 m/s 

ground 

540 

210 

120 

54 

32 

18 

10-

6 

1.5 

0.5 

0.3 

0.2 

The exposure to concentrations of HF can be calculated similarly, 

sures due to the ai=rborne releas.e of 1620 kg- of HF are shown below. 

ExpoHF exposure, due to the airborne release, of 1620 kg of HF 

-

Distance 

(meters) 

- 100 

200 

300 

500 

,700 

1,000 

1,500 

2,000 

5,000 

10,000 

15,000 

20,000 

" - .- -

F, 1 m/s 

buoyant 

_0 

_9_ 

-68_ 

160 

160 

140 

92 

65 

16 

5 

1.8 

0.9 

HF exposure (mg/m3) 

F, 1 m/s D, 4.5 m/s 

ground buoyant 

5900 9 

2900- 77 

1700 86 

770 59 

450 41 

- 250 25 

120 14 

67 9 

.: 9 2.3 

1.2 0.8 

0.2 0.4 

0.06 0.3 

-

D, 4.5 m/s 

ground 

790 

310 

170 

79 

47 

27 

15' 

9 

2.2 

0.:7 

0.4 

0.2 

34 
Intake 

243 

45 

8.6 

4.5 

(

m

9) 

2.2.3.4 Implications for Emergency Preparedness 

Of all the accidents considered in this Regulatory Analysis, the rupture 

of a heated 14-ton cylinder of UF6 is clearly and by far the most hazardous to 

people offsite, The corrosive effects of exposure to HF and heavy metal poisoning due to uptake of uranium are discussed separately below. 

Heavy metal poisoning: We consider the best estimates of the health 

.effects of uranium intake to be those in two DOE reports* based on the work of 

a panel of experts on uranium toxicity. The effects are summarized below: 

Health Effect , 

50% Lethality 

Permanent damage 

Renal injury (transient) 

No effect 

It is not likely from the calculated results that lethal intakes are 

actually plausible^for outdoor releases of UF6. In order to calculate lethal 

intakes it is necessary to assume little or no-buoyahcy, which is believed to 

be incorrect, and little br no effort on the part of the exposed individual to 

escape the plume, which may not be a reasonable assumption. We conclude that 

lethal intakes of uranium by people offsite are not really plausible under 

realistic conditions. 

Permanent kidney damage, on the other hand, does appear possible. From the 

intakes calculated above permanent kidney damage could occur as far as 2000 m 

(1.2 miles) under very adverse weather (F, 1 m/s) and no attempt to escape the 

plume, Under more typical conditions (D, 4.5 m/s, some .buoyancy, and-attempted 

| escape) permanent kidney damage would not be expected offsite. 

S Transient kidney injury appears to be quite plausible. Under highly adverse 

{ conditions (F, 1 m/s) it might be possible as far as five miles away. . Under 

-: more typical conditions (D, 4.5 m/s and some escape attempt) transient injury 

might occur as far as 1 mile away. 

*R. A. Just and V. S. Emler, "Generic Report on Health Effects for the U.S. 

Gaseous Diffusion Plants," DOE Report K/D 5050, Section VIII, Part 1, 1984. 

35 
P.47 

It is Commission policy for nuclear power plant accidents to plan to avoid 

acute fatalities and serious injuries for the worst case accidents and protective action guide doses for more probable accidents. With this in mind, the 

recommended protective action for rupture of a 14-ton cylinder would be evacuation to a distance of 1 mile, if possible. This would avoid acute fatalities 

and serious injuries for.worst-case accidents and transient kidney injury under 

more typical conditions. 

HF: Estimates of the health effects are from a recent DOE report.* The 

effects described here are based on concentration as applied to a 15 minute 

exposure: "--."'•'•-- • - -

Health effect HF concentration (mg/m3) 

Lethal (15 min) 

Unbearable for 1 min 

Irritation (15 min) 

Detectable by smell but 

no health effects 

3500 

100 

13 

2.5 

From the calculated HF exposures given above, lethal exposures offsite 

are not plausible. y - _ -.• . . 

Unfortunately, levels for permanent injury are.not knowrv. As a consequence 

we are substituting the concentration of 100 mg/m3 as the level considered to 

be "unbearable" for more than a minute. Such levels may occur out to about 

1500 meters under adverse conditions. Generally, they would not be expected 

to occur offsites under typical conditions (Dy 4.5 m/s). if one discounts somewhat the ground level release values. _ . '""*-"' '-

Irritation appears possible out to at least 5000 meters (3 miles) under 

adverse meteorology and roughly 1500 m (1 mile) under typical-conditions. 

Thus the consequences, of HF exposure are similar in severity to those from 

uranium intake. Consequently tbe one-mile evacuation suggested for the rupture 

of a 14-ton cylinder of UF6 is appropriate for protection against both uranium 

and UF6. ; -

*R. A. Just and V. S. Emler, "Generic Report on Health Effects for the U.S. 

Gaseous Diffusion Plants," DOE Report K/D 5050, Section VIII, Part 1, 1984. 

36 
$ A' 

\ 

The U.S. Department of Transportation has also established evacuation guides 

HF releases.* For small leaks (drum, small container, small leak from a 

.» tn e DOT recommends isolation in all directions to a distance of 240 feet 

7S meters). . For a large spill from a tank (i.e. railroad tank car) the DOT 

ends isolation in all directions to a distance of 490 feet (150 meters) 

•nd then evacuation in a downwind direction to a distance of 1.8 mile and a 

<dth of 1-2 mile. The DOT distances, however, are based on a larger quantity . 

f HF. Thus, the one-mile distance suggested here is consistent with DOT recommendations. DOT evacuation distances are based on atmospheric stability Class D 

and wind speed of 4.5 m/s. DOT states that-evacuation distances based on those 

assumptions have proven to be adequate under actual accident situations. -

ce 

2.2.4 Enrichment Plants " 

At present there are no NRC-licensed enrichment plants, nor are there any 

immediate prospects for one. Basically, however, enrichment plants receive 

UP6 from conversion plants and ship UF6,Eenriched in U-235, to fuel fabrica-" 

tion plants. Thus the-types of potential accidents are similar to those at" 

conversion plants and fuel fabrication plants. 

2:2.4.1 Accident History 

j .-

Several large releases of UF6 have occurred at enrichment plants, as 

shown in Table 4. These have been the result of the ruptures of heated large 

10-ton or 14-ton cylinders. The largest release was the 1978 cylinder rupture 

at the Portsmouth, Ohio gaseous diffusion plant, which released 9500 kg of "_. 

UF6. . -

*"Hazardous Materials-Emergency Response Guidebook' 

Transportation report D0T-P5800.2, 1980, op. cit. 

U.S. Department of 

37 
2.2.4.2 Accident Source Terms 

Sourde terms for two types of accidents are considered: UF6 releases and 

criticality accidents.-

The UF6 release for natural uranium is considered to be the same as for v 

the UF6 release previously discussed for UF6 conversion plants in Section 2.2.3.2. 

The UF6 releases for enriched uranium are considered to be the same as those 

for fuel fabrication plants that will be discussed in Section 2.2.5:2. 

The criticality accidents is assumed to be the same at the criticality 

.accident for the fabrication plants that will be discussed in Section 2.2.5.2. 

2.2.4.3 Calculations of Doses 

Doses due to a UF6 release from the rupture of a 14-ton cylinder of 

natural uranium are the same as those given in Section 2.2.3.3 for UF6 conversion-plants. Doses due to_UF6 releases of low and high:enriched uranium are 

the same as those that will be given in Section 2.2.5.3 for fuel fabrication 

plants . . _ '•-•'.:.'•-•_ ' 

Doses due to a criticality are the same as those given for a criticality 

at a fuel fabrication plant in Section 2.2.5.3. 

2.2.4.4 Implications for Emergency Preparedness 

-Offsite emergency preparedness at uranium enrichment plants should be " -

based dh chemical toxicity from a large UF6 release. Thus, uranium enrichment 

plants should be considered a potential chemical hazard, not a radiation hazard. 

Basically, uranium enrichment plants should have the same level of offsite 

emergency preparedness as UF6 conversion plants. Currently, enrichment plants, 

if licensed, would be covered under Part 50 of NRC regulations. The emergency 

preparedness requirements in Part 50, which were developed for nuclear power 

"plants, are clearly excessive for enrichment plants. However, because NRC does 

not currently license any enrichment plants, the discrepancy is academic. 

38 
pr c PIIPI Fabrication - Uranium 

c 1 fabrication plants(generally receive UF6 enriched in the uranium-235 

convert it generally "into highly refractory uranium oxides, form the 

oxides into pellets, and load the pellets into metal-clad fuel elements. 

Moment to nuclear power plants. In most cases the uranium-235 is enriched 

than 5%, but at several plants the enrichment exceeds 93%. However, 

Ity one 

licensed plant (Nuclear Fuel Services, Erwin, Tennessee) currently 

P'^P^-mlts'the volatile.UF6 in highly enriched form. 

{• f vi. ~ Accident History 

:- A'y"--

A^yMirz 

•**i..?. 

Among the...accidents that have occurred in processing uranium are criti-

;^iHtt1Uy accidents, fires, and releases of UFg. 

.- -:«ii - $j n c e the first successful self-sustaining nuclear chain-reaction there 

been no less than 37 occasions when the power .level of fissile systems 

^ y:1jsi|i unexpectedly-because of unplanned or unexpected changes in^system reac--

"v.$~t.lvity. Of these-37 cases, six cases, caused eight deaths, two of which occurred 

:'-'-.rfa"4.ne early", rushed pace near the end of World War II. 

y-:3> •'" Of these 37 criticalities, eight occurred in fuel cycle facilities (7 in 

^^^tfct'U.S. and one abroad) and are thus relevant to this analysis. The remaining 

•Jt occurred in nuclear reactors or critical assembly experiments. The seven 

rtUvant.LL. S. fuel cycle facility criticalities are listed in Table 5. One 

tyyifcftttfrred-in a licensed facility (Wood River Junction, R.I., 1964). 

-There are several lessons about criticalities that can be learned from 

y.ystudying these accidents. Accidental criticalities can occur and occasionally 

When they "occur the. doses to workers can be very large, sometimes fatal, 

vjv^.—— sometimes requiring hospitalization. Radioactive solutions can be ejected 

can contaminate workers and the plant area. No offsite contamination or 

lation doses have been reported. 

39 
Table 5. Criticality Accidents In Fuel Cycle Facilities 

Date Location Process Cause 

Total 

Fissions Duration Personnel Exposures 

Contamination 

Ih-plant 

Contamination 

Out of 

Building 

June 16, Y-12 Process1958 ing Plant, 

Oak Ridge, TN 

Dec. 30, Pu Process1958 1ng Plant, 

Los Alamos, NM 

Oct. 16, 

1959 

o 

Jan. 25, 

1961 

April 7, 

1962 

July 24, 

1964 

Oct. 17, 

1978 

Idaho Chemical 

Processing 

Plant, Idaho 

Reactor Test 

Site 

Idaho Chemical 

Processing 

Plant, Idaho 

Reactor Test 

Site 

Hanford Works, 

Richland, WA 

Scrap Recovery 

Plant, Wood 

River Junction, 

RI 

Chemical 

Processing 

Plant, NRTS, ID 

Recovery of highly 

enriched uranium 

by chemical 

methods. i 

Recovery of 

plutonium from 

scrap. 

Transfer of highly 

enriched uranium 

solution. 

1.5 x 101 7 1 sec 

Wash water added to 1.3 x, 101 8 18 min 

U 02 (N03 )2 solution 

1n 55-gal. drum. 

Liquid phases of 

plutonium separated 

out'. 

Solution transferred 

to unsafe geometry 

(5000 gal tank) 

4 x 101 9 15 to 

20 m1n 

Transfer of highly Solution transferred 6 x IO17 1 sec 

enriched uranium 

solution. 

Plutonium 

processing ' 

Recovery of highly 

enriched uranium < 

Solvent extraction 

column 

to unsafe geometry 

Plutonium solution 

incorrectly siphoned 

Solution hand-poured 

Into unsafe geometry 

8 x IO17 37.5 hr 

1.3 x 101 7 

3 x 101 8 

2 short 

pulses 

1.5 hrs 

apart 

15 min 

8 people. Doses of 461, Small local None 

418,, 413, 341, 298, 87, contamination reported 

29 rads. No fatalities. 

3 people.' Doses of 

12,000, 134, 53 rads. • 

One fatality. 

19 people. No direct 

gamma or neutron dose 

because tank was 

shielded, but beta 

doses from released 

radio activity of 

50 rem, 32 rads, and 

smaller amounts for 

17 other people. 

None. Shielded 

operation 

None 

reported 

Yes 

airborne 

beta 

activity 

i • 

None 

reported 

None 

reported 

Not 

reported 

None 

reported 

3 people. Doses of 110, 

43, 19 rads. 

3 people. Doses of 

10,000, 100, 60 rads. 

One fatality. 

Less than .13 rem. 

(In shielded cell) 

None 

reported 

20% of 

solution 

splashed 

out of tank 

A1r monitors 

detected 

considerable 

activity 

T, 

None 

reported 

None 

reported 

Filters 

removed 

most 

particles 

References: William R. Stratton, "A Review of Criticality Accidents," AEC Report LA-3611, 

October 17, 1978,"' DOE'Report ACI-362, November, 1978. 

>j < 1 hr 

1967 and ."Recovery of ICPP from Criticality Event of 

mm 
•rf**' 

A number of fires and explosions involving uranium or thorium, which would 

similarly, have been reported. Uranium metal is pyrophoric. Uranium 

•^y - i heated or in powdered form or heated as a solid will spontaneously ignite 

ff & nosed to air. Reported fires and explosions involving uranium or thorium 

* Included in Table 6 below. What is noteworthy is that these fires have had 

^ ««tt1e consequence with regard to. either personnel exposure or ground cohtamina-

' Ktttx Reported offsite contamination levels were generally below the levels 

t th e NRC allows on equipment to be released for unrestricted use.* 

By comparison, the accident record for plutonium, which has a much higher 

t-t-tr activity, is much different. Plutonium accidents have been characgpec< ii*- , 

t*rized by extensive radioactive contamination and personnel exposures. Yet 

avtn the most serious of these accidents, the Rocky Flats fire, caused only 

small fraction of the plutonium involved to be released. Of hundreds ofkllograms of plutonium involved in the fire, only 0.003 g was released through 

A damaged exhaust system.** Thus the overall release fraction for plutonium 

^gf about 10-8, based on the estimated release quantity compared to the 

y -aantity involved in the fire.** = -

Table 7 lists other accidents involving uranium fuel fabrication, but 

not including UF6 releases, fires, or exposions which were listed previously. 

These accidents in Table 7 all involved ventilation systems. None of the 

accidents listed in Table 7 caused any offsite doses approaching the 1 rem 

lower limit of the protective action guides. 

•Surface contamination levels for uranium allowable on equipment to be 

released for unrestricted use are average: 5000 dpm alpha/lOO^cm*, maximum: 

15,000 dpm alpha/100 cm2, and removable: 1000 dpm jilpha/100 cm2.- These 

" values are found in Regulatory Guide 1.86, "Termination of Operating Licenses 

for Nuclear Reactors," and "Guidelines for Decontamination of Facilities and 

Equipment Prior to Release for Unrestricted Use or Termination of Licenses 

for Byproduct, Source, or Special Nuclear Material," USNRC, July, 1982. 

**H. K. Elder, "Technology, Safety and Costs of Decommissioning Reference 

Nuclear Fuel Cycle and Non-fuel Cycle Facilities Following Postulated 

Accidents," NUREG/CR-3293, page 3.3, 1985. 

41 
Table 6. Fires and Explosions Involving Uranium and Thorium 

Date Facility Release Description 

6-27-49 

6-12-53 

Los Alamos 

Laboratory, NM 

10-29-52 Truck in Kansas 

City, MO 

12-9-52 AEC facility 

None reported 

Considerable 

U.S. 

None 

Onsite contamination up to 15,000 

dpm/100 cm2. 

8-20-56 

9-21-56 

6-23-58 

9-26-60 

9-20-63. 

6-29-67 

9^2-72 

3-12-81 

AEC contractor 

Truck in Detroit, 

MI 

AEC contractor 

Attleboro, MA 

M&C Nuclear 

Attleboro, MA 

Controls, Inc. 

Attleboro, MA 

Kerr-McGee 

Crescent, OK 

United Nuclear 

Fuel fabrication 

facility 

Nuclear Metals 

Concord, MA 

100,000 dpm/ 

100 cm2 onsite. 

500 dpm/100 cm2 

offsite. 

None 

No material 

loss 

Enriched U 

no exposures. 

None 

detected 

Minor 

Uranium at 

15 times MPC 

Minor inplant 

contamination 

Fire broke out in a drum 

containing uranium metal 

turnings.i 

Truck carrying uranium 

metal burned. Uranium 

ignited and. much was lost. 

Molten uranium metal was 

being cast in a vacuum. 

Spill ruptured vacuum. 

Uranium-then burned. 

Explosion of powdered 

uranium and CC14" in 

glovebox. 

Thorium explosion 

Drum containing thorium •-

metal started to burn. 

No contamination. No 

exposures, y "'-..•. 

Fire in slightly . .yenriched uranium scrap 

in perchloroethylene. ••-..-

Magnesium explosion in 

vacuum induction furnace. 

Fire in filter box . 

exhausting enriched 

uranium. No contamination 

on or offsite. - - - ''"... 

Explosion in ion exchange 

column. - '"•-'-'' 

Flash fire caused by 

organic contaminants in 

ductworks. Considerable . 

damage. 

Fire in scrap-packaging . 

building from spontaneous 

combustion of 10 lbs of 

uranium turnings. No 

overexposures. 
Table 7. Other Accidental Releases from Uranium Fuel Fabrication Plants 

(UF6 releases, fires, and explosions excluded) 

Date Facility Release Accident description 

.4-2-71 

2-28-73 

8-24-73 

9-6-74 

1-24-75 

1-79 

Babcock and Wilcox 

Research Faci1ity 

Lynchburg, VA 

General Electric, 

Wilmington, NC 

10 Microcurie in 

plant 

Below MPC 

Babcock and-W-i'lcox .Decontamination 

Apollo, PA " required on and 

offsite. Release 

•_-'••'• was 6.3 microcuries. 

12-8-73 Babcock and Wilcox Offsite release of 

Apollo, PA 

Westirighouse 

Columbia, SC 

natural urani urn 

4 times MPC. Offsite 

decontami nati on 

required. 

UO, 

Babcock and Wilcox Enriched uranium 

Apollo, PA 

General Electric, 62 Kg of low 

Wilmington, NC enriched U02 powder 

stolen. 

Contamination outside a 

hot cell due to a plug 

bei ng i hstal1ed wi thout 

a sealing bellows. 

HEPA filter failed. 

Enriched uranium released 

when corroded scrubber 

spray nozzle did not 

provide enough scrubbing." 

Inadequate ventilation 

of caieiner allowed 

uranium to escape 

through canopy exhaust. 

Leak in a roughing filter 

allowed U02 power to be 

discharged directly to 

the air. 

Malfunction of scrubber/ 

ventilation system. 

Two cans of powder stolen 

and used in extortion 

attempt. Thief arrested, 

convicted and imprisoned. 

Powder recovered. 

43 
2.2.5.2 Theoretical Calculations of Releases 

Criticality accidents, UF6 releases, fires, explosions, and tornadoes 

have been considered in various analyses of accidents in a fuel fabrication 

plant. The most serious accidents appear to be criticalities and UF6 releases. 

Thus, we consider those accidents here. 

Criticality Accident: the NRC staff has developed a set of assumptions 

on the release of radioactive fission products from a criticality accident 

occurring in a solution. The assumptions are published.in Regulatory Guide 

3.34, "Assumptions Used for Evaluating the Potential Radiological Consequences 

of Accidental Nuclear-Criticality in a Uranium Fuel Fabrication Plant," July 

1979. '" 

Regulatory Guide 3.34 assumes a criticality excursion occurs in a vented 

vessel of unfavorable geometry containing a solution of 400 g/1 of uranium 

enriched in U-235. The excursion produces an initial burst of 101 8 fissions 

in 0-5 second followed successively at 10 minute intervals by 47 bursts of 

-1.9 x 1017_ fissions "each for a total of 101 9 fissions in 8 hours. The excursion 

is assumed to be-termi hated by evaporation of 100 liters of the solution. The 

amounts, of radioactive materials assumed to be released from .the solution 

vessel to the room air are given in Table 8. 

Regulatory Guide 3.34 allows credit for removal of fission products by 

filters in the ventilation system. In order to escape to the environment the ^ 

room air must pass through a filter system by means of the building ventilation 

system. Even if doors are opened the ventilation system should exhaust the 

fission products through _the filters. In this analysis the filters are assumed 

to remove 75% of the iodine but none of-the inert gases. This analysis assumed 

the ventilation-system operates at 5 air changes per hour. 

The NRC staff has recently analyzed the potential consequences assuming a. 

criticality accident for Exxon Nuclear in Richland, Washington,*"Combustion 

Engineering in Hematite, Missouri,** and Nuclear Fuel Services in Erwin, 

*"Environmental Impact Appraisal, Exxon Nuclear Company, Nuclear Fuel 

Fabrication Plant, Richland, Washington," Docket 70-1257, NRC Office of 

Nuclear Material Safety and Safeguards, August, 1981. 

**"Environmental Assessment, Combustion Engineering, Inc., Nuclear Fuel 

Fabrication Plant, Hematite, Missouri," Docket 70-36, NRC Office of 

Nuclear Material Safety and Safeguards, November, 1982. 

44 
P.56 

Table 8. Amounts of Radioactive Materials Released 

to Room Air Due to a Criticality Accident 

(IO18 fissions initially and 1.9 x IO17 

fissions every 10 minutes for 8 hours) 

... 

Radionuclide 

Kr-83m 

Kr-85m 

Kr-87 

Kr-88 

Kr-89 

Xe-133 ." '" 

Xe-135m " 

Xe-135 

Xe-137 

Xe-13'8 

1-131 

1-132 

1-133 

1-134 

I-135_ 

- -_ ; 

Half-life 

in hours r 

1.8 

4.5 

1.27 

v 2.8 

0.05 

125. 

-= 0.26 •_ 

9,1 

0.06 

0^24 

192, 

2.3 

21. 

. 0._88 

6.6" 

Radioactivity 

•0-0.-5-.hr 

22 

21 

140 

91 

5900 " 

4 

310 

: 50 

6900 

1800 

0.3 

38 

5.5 

- 160 

17 

in Curies 

0.5-8 hr 

140 

.130 

850 

. 560 

36,000 

23 

1900 

310 

42,000 

11,000 

1:9 

240 

35 

980 

100 

Source:. Regulatory Guide 3.34, Table 1, July, 1979. 

"""..-• The values for radioiodines in Table.lof 

: the guide were reduced by a factor of 4 to 

••••- account for retention in the solution water. 

Tennessee.* In each _case the radionuclide releases from Regulatory Guide 3.34 

were used. - y_ 

Low-enriched UFfir " The Exxon, arid-Combustion Engineering analyses** also 

considered releases of UF6, as did a recent analysis for the General Electric 

fuel fabrication plant in Wilmington, North Carolina.t The worst-case accident 

in the Exxon and Combustion Engineering analyses was assumed to involve the 

-release of UF6 as might occur from valve or line failure of a heated cylinder 

being unloaded. Assuming that a full cylinder of UF6 (2500 kg) at elevated 

^"proposed New Emergency Preparedness License Conditions at NFS-Erwin," NRC 

Commission Paper SECY-82-311, July 23, 1982. 

**0p. cit. 

t"Environmental Impact Appraisal, General Electric Company, Wilmington 

Manufacturing Department," Office of Nuclear Material Safety and Safeguards," 

NUREG-1078, June, 1984, page 69. 

45 
temperature started to leak and that no additional heat was supplied after 

cylinder failure, the NRC staff estimated that about 22 percent of the material 

would be released before the UF6 would be cool enough to solidify and have a 

vapor pressure low enoujgh so that the release would stop. The NRC staff estimated that such a release would last for 15 minutes,, and ,540 kg of UF6 would 

be released. This has a uranium content of 360 kg. The staff assumed the. 

uranium released would react with water in the air and form highly soluble 

U02F2 of a respirable particle size. 

High-enriched UFB: In evaluating the need for offsite emergency preparedness at Nuclear Fuel Services, Erwin, Tennessee, the NRC staff concluded that, 

the UF6 accident to be considered-was release from a 15-kg cylinder containing 

high enriched uranium.* This is largest cylinder used at the site for highly 

enriched UF6. An accident in 1962 breached one cylinder. Of the 15-kg contents, 

6 kg was recovered'but 9 kg was not recovered.and presumably much of the 

material escaped from the plant. 

2.2.5.3 Calculations of Doses ' _ . 

Potential radiation doses due to criticality accidents and UF6 releases 

previously calculated by the NRC staff for Combustion Engineering, Exxon, and 

Nuclear Fuel Services are summarized in Table 9. 

Criticality accident: To calculate the dose due to a criticality accident, 

the Exxon analysis assumed a wind speed of 1 m/sec, atmospheric stability 

class F, and a building wake factor of 1.0 beyond 500 meters (i.e. no building 

wake assumed). The building ventilation rate is.assumed to be 30 air changes/hr. 

The whole body doses were calculated to Be 0.004 rem at 3600 m (the nearest 

residence) and 0.009 rem at 2000 m (the nearest industrial site). The doses 

to the thyroid were calculated to be 1.7 rem at 3600 m and 4.5 rem at 2000 m. 

If one assumed only 25% of the iodines would pass through the filter system,. 

the thyroid doses would be 0.4 rem at 3600 m and 1.1 rem at 2000 m. 

The NRC analysis for the Combustion Engineering plant made similar assumptions. The whole body dose at the nearest residence (800 m) would be 0.27 rem. 

The thyroid dose at 800 m would be 1.7 rem. If the filters reduced the iodine 

concentrations by 75%, the thyroid dose would be 0.4 rem. 

*"Proposed New Emergency Preparedness License Conditions at NFS-Erwin," NRC 

Commission Paper SECY-82-311, July 23, 1982. 

46 
Table 9. Offsite Doses Calculated for Fuel Fabrication Plants 

Analysis 

Key 

Assumptions 

Criticality UF6-low enrich. 

Effective DE Thyroid DE Effective DE Bone DE 

UF6-high enrich. 

Effective DE 

NUREG-1140 

Combustion 

Engineering 

Exxon 

NFS, Erwin 

Building size: 250 m2 

Wind: F, 1 m/sec 

Release height: ground 

Building size: 0 

Wind: F, 1 m/sec 

Release height: stack 

Building size: 0 

Wind: F, lm/sec 

Release height: ground 

Building size: 0, , 

Wind: G, 0.5 m/sec 

Release height: same 

level as residence 

0.5 to 

2.6 rems at 

100 m 

0.27 rem 

at 800 m 

0.009 rem 

at 2000 m' 

1.1 to 

8.2 rems 

at 100 m 

.(child's 

thyroid) 

1.7 rems 

at 800 m 

4.5 rems 

a!t 2000 m 

5 rems 

at 1000 m 

0.05 rem 

at 800, m 

0.11 rem 

at 2000 m 

0.82 rem 

at 800 m 

1 . 7 reijis 

at 2000 m 

0.2 to 

1.5 rem 

at 100 m 

-1 rem 

at 1000 m 
This analysis calculated an effective dose equivalent due to the airborne 

release from a criticality as 0.5 to 2.6 rem at 100 m. The dose from prompt 

gammas and neutrons from excursions after the first one should be added to that 

dose, but those doses have not yet been calculated. [ 

We calculated the thyroid dose to a child due to the radioiodine releas.e 

from a criticality accident, to be 1.1 to 8.2 rems at 100 m. 

Low-enriched UFB: For the UF6 release, the Exxon plant analysis assumed 

a ground level release and calculated a dose to the bone of 1.7 rem at 2000 m. 

The whole body dose was calculated to be 0.11 rem. In the analysis the NRC 

staff published for the Combustion Engineering plant, the release was assumed 

to be through a stack. The bone dose was calculated to be 0.82 rem at 800 m. 

The whole body dose was calculated to be 0.05 rem." 

For purposes of this Regulatory Analysis uranium intakes and HF concentrations have been calculated. Radiation doses were not calculated since they are 

of lesser concern. 

The release was assumed to be 540 kg of UF6i at ground level. The uranium 

content -is 364" kg. Atmospheric dispersion values from Figure 1 were used. 

Uranium intakes are shown below for ground level releases. Buoyant releases 

are not calculated because there may not be enough material to create buoyancy. 

Uranium intakes due to a ground 

level release of 540 kg of UF6 

Distance 

(meters) 

100 

200 

300 

400 

500 

1000 

1500 

2000 

5000 

Uranium 

F, 1 m/s 

320 

150 

90 

42 

24 

14 

6 

4 

0.5 

intake (mg) 

-

D^ 4.5 m/s 

43 

16 

9 

3 

2.2 

1.4 

0.8 

0.5 

0.1 

High-enriched UFR: For the Nuclear Fuel Services Plant, the NRC staff 

previously calculated the effective dose equivalent at 1000 m due to the 

release of 15 kg of UF6, high-enriched, to be 1 rem. For a 9 kg release of 

high enriched uranium, we calculated an effective dose equivalent of 0.2 to 

48 
1.5 rems at 100 m, a uranium intake of 0.3 to 2.6 mg, and exposure to HF at a 

concentration of 1.0 to 7.8 mg/m3. A building size of 10 m x 25 m, neutral 

buoyancy for the plume, ground level release, 5-minute release deration, 1.5% 

U-234 with, solubility Class D for uranium and F, 1 m/s and D, 4.5 m/s meteorology" 

were the assumptions. ..."'. 

2.2.5.4 Implications for Emergency Preparedness 

The implications of criticality accidents and UF6 releases are discussed 

separately below. - -.- _'_".. 

Criticality accident: Usingwhat are believed to be reasonable assump- ... 

tions, at 100 m effective dose equivalents of 0.5 to 2.6 rems for F, 1 m/s _andD, 4.5 m/s-meteorologies were calculated compared to. a 1 to 5 rem protective 

action guide. A child's thyroid dose of 1.1 to 8.2 rems was calculated compared 

to a 5 to 25 rem protective action guide. The calculated doses exceed the lower 

end of the range where protective actions should be considered out to about 

200 to 250" m. "_- " y~. 

Low-enriched UFfi release: For the release of 540 kg of lowTenriched. UF6, 

lethal intakes (242 mg) offsite do not seem plausible. -Intakes sufficient to 

cause permanent kidney injury (45 mg) are calculated for adverse meteorology 

with not buoyancy to about 500 meters, although consideration of buoyancy might 

easily eliminate calculated permanent injury under any conditions. Transient 

kidney injury might occur to 100 meters under worse case meteorology and to 

perhaps about 300 meters under more typical meteorology. There would.be-no 

observable effects at 400 meters for typical meteorology* Therefore, in keep-_ 

ing with the Commission's policy on nuclear power plant emergency preparedness, 

avoiding fatalities and serious health effects for worst case and protective 

action guide doses for more probable events, a response distance of roughly 

400 meters is recommended. 

High-enriched UFB: For the release of 9 kg of high-enriched UF6 during 

F, 1 m/s meteorology, the dose at 100 m is 1.5 rems effective dose equivalent 

and the uranium intake is 5 mg. For D, 4.5 m/s meteorology the dose at 100 m 

is 0.2 rem and the uranium intake is 0.7 mg. Buoyancy is not considered in 

either case since the quantity of material is so small. Protection actions 

to reduce dose may be appropriate during F, 1 m/s meteorology to a distance 

of 150 m. 

49 
T 

P.61 

2.2.6 Fuel Fabrication - Plutonium 

There is currently no plutonium fuel fabrication being done in the U.S., 

but accidents for facilities fabricating plutonium fuel have been analyzed in 

NUREG-0002, "Final Generic Environmental Statement on the Use of Recycle 

Plutonium in Mixed Oxide Fuel in Light Water Cooled Reactors," 1976, usually, 

called: "GESMO." Because of the quality arid thoroughness of that report, we 

will directly quote relevant sections for this analysis* 

2.2.6.1 Accident-History* 

"A summary, of operational accidents in U.S. Government facilities, from 

1943 to 1970, is given in WASH-1192. For some facilities and operations having 

a general"resemblance to the various mixed oxide fuel cycle steps (in the areas 

of fuel-reprocessing, fuel fabrication, and scrap recovery) there have been:a 

number of accidents. Those which involved the possibility of environmental 

-release include, tbe following: -•• = 

Five solution criticality events (1958—2 events; one each in 1959, 1961, 

y and-1-962) in reprocessing or recovery operations involving highly enriched 

uranium or plutonium. All were of small consequence in terms of property 

damage or releases of radioactivity to the environment, but one fatality 

and several high radiation exposures occurred among operating personnel. 

Chemical- explosion in evaporator (.1953), related to fuel reprocessing. 

Explosion and fire in plutonium purification facility (1963). 

Metallic fuel fire (reactive metal) in process dissolver (1960), related 

to fuel reprocessing. ; 

Spontaneous fire in radioactively contaminated, combustible waste (1951). 

*From GESMO, pages 11-45 to 11-48. The references given in GESMO are not 

included here to save space. The reader wanting the references should refer 

to the original GESMO. 

50 
P.62 

Two fires at the Rocky Flats plutonium fabrication and recovery facility, 

(1957) and (1969). The 1969 fire caused $45 million in property damage. 

Both fires are attributable to spontaneous ignition of plutonium metal 

which is not involved in the mixed oxide fuel cycle. 

Fire around an anion exchange column, fuel reprocessing plant. (1964). 

"Some of these accidents occurred during the early years of operations 

With new facilities and newly developed technologies. All were investigated 

and corrective actions were taken (e.g., design changes) to make the events 

unlikely of recurrence.- Such corrective actions have been carried forward, 

where applicable, into design practices for new facilities, both government 

and commercial. During the past decade, criticality accidents have disappeared 

from the accident scene, and fire or explosion involving reactive metals has 

become the predominant major accident in government facilities. 

"On a comparable basis, accident experience in commercial facilities to 

date include: ' ~ 

A solution criticality accident in recovery operations involving highly 

enriched uranium, fatal to operator (1964). 

A A series of dissolver "fires" (reactive metal), fuel reprocessing plant 

(1967-1968). 

Final HEPA filter bank failure (inadequate mechanical support), fuel 

reprocessing plant (1968). 

Fire in plutonium contaminated wastes, fuel fabrication facility {1973). 

There was no detectable release of plutonium to the environment. 

Explosion in plutonium glovebox, fuel fabrication facility (1972). 

5.6 uCi of alpha activity was released via the stack. 

About 

51 
"The measured and estimated quantities of long-lived alpha activity 

released from the Rocky Flats plant during its first 20 years of operation are 

summarized in Table 11-12.... 

"The specific and extensive modifications made to all plutonium handling 

facilities at Rocky Flats subsequent to the 1957 fire (especially the substitution of "flame resistant filters for those formerly used, and the addition of 

fire protection in the filter banks and plenums) were clearly responsible for 

the vastly improved containment of alpha activity during the 1969 fire. The 

new plutonium recovery facility now under construction at Rocky Flats (as. a 

replacement for older facilities) is being built under criteria that should 

provide even greater assurance that the facility will be able to confine plutonium releases to exceedingly small values, even under severe abnormal circumstances— including natural events", such as tornados." 

z . "Table 11-12 

" LONG-LIVED ALPHA ACTIVITY RELEASED FROM ROCKY.FLATS 

Date Ci rcumstances Quanti ty 

1958-1968 

1957 

1969 

1953-1970 

Leakage of. Pu contaminated machine oil 

stored at the. Rocky Flats site 

Fire in Bldg. 771 resulting in major 

damage to fi'i ter system 

Fire in.Bldg. 776. 

Normal effluent releases (cumulative) 

5.3 Ci to soil at 

drum storage area 

60 uCi, airborne, 

mostly during fire 

0.2 uCi, airborne, 

over 6-day period 

during and after 

fire 

< 41 uCi, airborne 

91 uCi, liquid 

effluents 

52 
P.64 

2.2.6.2 Accident Source Term* 

"A wide spectrum of credible accidents for these plants has been analyzed 

and their potential consequences estimated. 

"Some incidents, such as punctures or tears of gloves or other glovebox 

malfunctions, are expected to occur as part of the normal operation. Other 

more serious accidents—such as glovebox window breakage—will occur far less 

often, although the offsite consequences from such accidents are judged to be 

insignificant, they are included in the estimate of airborne effluents resulting from normal operations. Upper limit accidents that may occur include a . 

criticality incident, an explosion, or a fire.... 

Criticality 

"There have been no criticality accidents to date in process operations 

where undermoderation is a primary method of control, and few in-aqueous or 

moderated systems. The number of fissions has-been estimated to be IO1 8 in an 

accidental criticality. In calculating the effects, it is assumed that all of 

the noble gases krypton and xenon and 25% of the iodine formed by the fissions 

would escape. In addition, it is assumed that 500 grams of Pu would become airborne in a glovebox by the accidental criticality excursion. The ventilation 

filters are assumed to remain intact because a criticality is not an explosive 

process. The decontamination factor of three HEPA filters in series has been 

assumed to be 2 x IO7, lower by a factor of 50 than tbe decontamination factor 

assumed for normal operating conditions. [Pu release = 0.29 uCi alpha] Calculations show-that the maximum offsite individual dose commitment results from 

absorption of fission product iodine in the thyroid and amounts to 360 mrem. 

This is comparable to the dose to the closest theoretical resident from a criticality accident at.a U02 fabrication plant. The slightly different fission 

product yield and the presence of small amounts of plutonium particulate do not 

significantly alter the effects of a Pu02 criticality accident relative to 

those of a U02 criticality accident. 

'From GESMO, pages IV D-37 to IV D-39. 

53 
• H 

Fire 

"Unlike a criticality excursion or an explosion, a fire usually is not 

an instantaneous event and very(often starts from a small flame source. The 

design, construction, and operation of fuel fabrication plants considers in 

detail, the possibility of a fire and equipment and procedures for fire prevention. Regulatory Guide 3.16.-presents methods acceptable to the NRC for a fire 

protection program that should prevent, detect, extinguish, limit, or control 

fires and explosions and their hazards and damaging effects. Licensees must 

operate witbin these or acceptable equivalent constraints. Under these conditions, the probability of a fire of the magnitude considered in this statement 

is considered highly unlikely. In general, operators have-time to react to and 

extinguish small fires. The.process materials, oxides of uranium and plutonium, 

are not themselves flammable. The final filters are protected against fire by 

water spray systems installed in the duct some, distance upstream of the final • 

filters. Mist deflectors or collectors are installed between the water spray 

system and the filters to remove large drops of water. The water from the sprays 

collects in the bottom of the ductsand flows to a fire-water collection tank. 

This tank is either a" safe-geometry tank or a fixed-poison-controlled tank to 

preclude the possibility of a criticality accident as a result of a fire. 

"The final HEPA filters are located some distance from the gloveboxes. 

This separation distance and the water spray system should be sufficient to 

protect the filters against the effects of an" explosion or fire, but the fire 

or explosion is assumed to destroy the local filters on" the gloveboxes. 

Plutonium and uranium oxides reach the final filters". Based on an assumed 

room volume of 1,000 m3 and an air loadingof 100 mg/m3 for plutonium oxide 

particulate, 100 grams of plutonium would reach the filters.. Each of the 

filters is expected to remove.99.9% of the particulate reaching it, so that 

a total of 0.1 mg of Pu passes through the filters. [53 uCi alpha] 

54 
Explosion 

"An explosion might occur in a mixed oxide fuel fabrication plant at locations where an explosive mixture of vapors in air $ould be present. There is a 

potential for the existence of combustible gases at the sintering furnance and 

in the clean scrap reduction operation. In addition, flammable solvents are 

used in the dirty scrap recycle operation and may be used for cleaning fuel rods 

.and during'cleanup and maintenance operations. These operations are essentially 

the only ones that have a potential for supporting an explosion..... 

"The consequences of an explosion are similar to those of a fire. The 

amount of plutonium reaching or passing through the filters is estimated to be 

the same as that estimated for the fire [53 uCi alpha] and would have the same 

relatively minor offsite consequences." [End of quote] 

2.2.6.3 Calculations]of Doses 

For the accidental releases above, GESMO assumed the nearest resident was 

at 500 m. GESMO used a chi/Q of 1.3 x 10-4 for Pasquil! Stability Category D, 

wind speed 3 m/s, and 10 meter release height. GESMO assumed the plutonium to 

be soluble. 

The following dose commitments were calculated for GESMO: 

Type of accident Dose commitment (rem) 

Criticality 0.36 (thyroid) _.. -

Fire 0.02 (bpne)_ 

Explosion 0.02 (bone) 

Using the standard assumptions in this Regulatory Analysis, doses would 

be slightly higher than those in GESMO. 

2.2.6.4 Implications for Emergency Preparedness 1 

The GESMO analysis concluded that the design of plutonium fuel fabrication 

plants is adequate to prevent releases that would cause doses exceeding protective action guides. Thus special emergency preparedness would not be needed 

offsite. 

55 
In addition, GESMO (page 11-10) concluded there was little threat from the 

ingestion pathway. 

"A study of indigenous and experimental animals kept for longl periods in 

areas heavily contaminated with plutonium indicates that direct uptake of plutonium was small. Plutonium uptake by plants from soil and growth media, has 

been investigated in the field and in the laboratory under a variety of conditions. The concentration of plutonium in plants on a dry weight basis was never 

more than one thousandth of that in the growth medium, and only about one ten 

thousandth of that in the soil. The fraction of available plutonium absorbed 

from the gastrointestinal tract of-animals grazing on contaminated vegetation 

is less than one ten thousandth the total intake of the element and:measurements 

of plutonium transfer from the blood stream to milk suggest a further reduction 

in plutonium concentration by another factor of at least 10. Consumption of 

animal products by man will introduce another reduction factor of at least 10-4 

in the plutonium concentration entering the systemic circulation, except in the 

very young infant where the factor may approach 0.01. =It appears, therefore, 

that the possibility of transfer of plutonium from soil to man by way of the 

food chain is negligible." ~ . -

2.2.7 Spent Fuel Storage* 

Spent nuclear power plant fuel may be stored in pools of water, in. dry 

storage casks, in drywells, or dry vault storage.. Each of these methods is 

discussed in this section. ' - - • ,v .- -

*Source terms in this section were developed with the assistance of Fritz Sturz, 

Advanced Fuel and Spent Fuel Licensing Branch, Office of Nuclear Materials 

Safety and Safeguards, NRC. 

56 
P.68 

2.2.7.1 Accident History 

There have been no accidents associated with spent fuel storage that have 

had. any significance for offsite radiation exposure. Radioactive material has 

escaped to storage pool water when casks containing damaged fuel elements were 

opened. However, these events have no significance for offsite emergency 

preparedness. 

2.2.7.2 Accident Source Term 

Pool storage: Pool storage for spent fuel storage in pools, possible 

accidents and their effects are discussed in a generic environmental impact ... 

statement, NUREG-0575.* -

The accidents considered are: 

1. The rupturing of-fuel pins due to thf drop of a"fuel.assembly 

_" - .2. A tornado driven utility pole strikes the pool at-the worst-possible 

angleand ruptures-a 45 foot row of assemblies 

-3. Fires and explosions 

4. A criticality. accident 

5. High radioactivity in the pool water 

6. Rupture of a waste tank or piping 

-7— - 7. Lowering of the water level in the pool ^ 

; 8. Loss of the ability to cool the pool water - _ - .-

Of these accidents, the generic environmental impact statement for pool 

storage estimated that the most serious of these accidents is the rupture of 

a large number of fuel assemblies- by a tornado-driven missile. The statement 

calculated the release of radioactivity to be 19,000 curies of krypton-85 and 

0.00006 curies of iodine-129. (Actually, 1-129 is of only academic interest. 

Due to saturation, the 5-rem thyroid dose used as the protective action guide 

cannot be reached.) 

*NUREG-0575, Generic Environmental Impact Statement on Handling and Storage of 

Spent Light Water Power Reactor Fuel, pp 4-17 to 4-22, Volume 1, August 1979. 

57 
A more recent NRC report on pool storage evaluated accident consequences 

for the General Electric fuel storage facility in Morris, Illinois.* The 

most serious accident was considered to be the drop of a fuel storage basket 

in the water of the storage pool. The maximum drop would be about 7 m. While 

experience with similar drops indicates that only minor damage to a fuel 

assembly would result, the calculations assumed that all the fuel rods in four 

PWR fuel bundles would rupture and that all the plenum gases would be released 

to the pool water. The release was calculated to be about 6000 Ci of Kr-85 

and 0.00008 Ci of 1-129. 

Several other types of accidents were also analyzed for. the GE-Morris. 

facility. & loss of basin water was considered to be not .credible., an earthquake was estimated to cause minimal offsite radiological consequences, a 

tornado-driven missile was estimated to cause the same release as the fuel- _-., 

basket drop, and a criticality was estimated to cause minimal offsite doses.-

Dry cask storage: The accident assumed for this analysis is the removal 

of the lid of a dry cask in which" all the fuel rods have been damaged. The 

gaseous-activity in the gap between the fuel and cladding is assumed to be r 

released.. From NUREG-0575** 10% of the krypton-85 and 1% of the iodine-129 '-•-

activities are assumed to be in the gap. The cask is assumed to hold 24 PWR-•-~. 

spent fuel assemblies. The fuel is assumed to be less than 5% by weight 

uranium-235. " 

The fuel burnup for this analysis is assumed to be 33,000 megawatt-days 

per metric ton of uranium. . The fuel is assumed to have been removed from the 

reactor core b years earlier. Using these assumptions the activity released 

from a cask would be 8,000 curies of kryptori-85 and 0.004 curies of iodine-129. 

Drywell and dry vault storage-- While the number of fuel rods may be 

larger than-assumed for dry cask storage above, it is reasonable to assume that 

a single accident would not damage a larger number of fuel rods than assumed 

above for dry cask storage. Therefore, the dry cask storage source term is 

also "appropriate for drywell and dry.vault storage. 

*NUREG-0709, "Safety Evaluation Report related to the Renewal of Materials 

License SNM-1265 for the Receipt, Storage, and Transfer of Spent Fuel 

Pursuant to 10 CFR Part 72 - Morris Operation - General Electric Company -

Docket Nos. 70-1308 and 72-1," July, 1981, Chapter 7. 

**NUREG-0575, op. cit., Volume I, p. 4-18. 

58 
P.70 

2.2.7.3 Calculations of Doses 

Dose estimates previously published by the NRC staff for the pool releases 

described above are given in Table 10. 

Doses-.for'the dry cask storage accident described above calculated in this 

analysis are also shown in Table 10. The effective dbse equivalent for F, 1 m/s 

meteorology would be 0.003 rem and the child's thyroid dose would be 0.04 rem. 

For D, 4.5 m/s meteorology, the child's thyroid dose would be 0.005 rem. 

2.2.7.4 Implications for Emergency Preparedness 

The doses-shown i.n Table 10 are below the EPA's protective action guides 

for taking protective action after an accident. Therefore offsite emergency 

preparedness is-'not necessary for spent fuel storage either, in dry casks or in 

pools. 

2.2.8 New Fuel Storage 

New fuel will"at times be stored prior to being loaded into a nuclear 

power plant".core.- Sto.red new fuel does not require any offsite emergency 

preparedness because of its. minimal hazard. By comparison with spent fuel 

storage just discussed, no fission products are present. Thus no volatile 

radioactive materials are present, and no driving force, such as decay heat, 

is present to cause the uranium fuel to escape its cladding and become airborne. 

2.2.9 Reprocessing of Spent Fuel 

Spent fuel reprocessing is the mechanical and chemical processing of spent 

nuclear fuel to extract "enriched uranium and plutonium. from the fuel elements 

so tbey can be used in new fuel elements. Radioactive fission products are 

removed from the spent fuel and processed into high-level radioactive waste. 

Currently, no reprocessirig plants are licensed by NRC to operate in the 

U.S., nor are there any near term prospects for licensing any reprocessing 

plants. However, the Nuclear Fuel Services reprocessing plant in West Valley, 

New York, operated as a commercial plant under NRC license for many years, and 

59 
Table 10. Calculated Releases and Doses from Spent Fuel Storage Accidents 

O 

Reference 

Storage in pools: 

Generic Environmental 

Impact Statement, 

NUREG-0575 

Storage in pools: 

GE-Morris SER, 

NUREG-0709 

Dry cask, drywell, 

or dry vault 

storage: NUREG-1140 

Accident 

Tornado driven 

missile followed 

by calm 

Drop of a fuel 

storage basket 

Removal of cask 

lid with all fuel 

elements ruptured 

i 

Kr-85 

Release 

19,000 Ci 

6*000 Ci 

8,000 Ci 

• . 

Skin 

Dose 

I'MI 

0.06 rem 

at 275 m 

.Not 

calculated 

Not 

calculated 

,-

Effective Dose 

Equivalent 

Not calculated 

. 

• ' 

0.016 rem 

at 150 m 

0.003 rem 

at 100 m 

, 

1-129 Release 

0.00006 Ci 

0.00008 Ci 

0.004 Ci 

• 

Thyroid 1 

Dose 1 

0.03 rem • 

at 275 m • 

0.0004 rem 1 

at 150 m I 

0.005 to 8 

0.04 rem I 

within I 

100 m • 

(child) • 

I ' 
reprocessing plants have been operating in the U.S. weapons program for over 

40 years. In addition, reprocessing plants are operating in several foreign 

countries. 

I 

2.2.9.1 Accident History -

. The Nuclear Fuel Services, reprocessing plant was plagued by many small 

releases into ground water, surface water, and air as well as unusually high 

occupational radiation exposures. However, the plant never had an accident of 

significance for offsite emergency preparedness. 

Several criticality-accidents;-as listed in Table 5, have occurred in 

spent fuel reprocessing or in processes similar to those that would be used 

in spent fuel reprocessing plants. 

No other accidents of significance offsite are:known to have happened 

in spent fuel reprocessing. 

2.2.9.2 Accident Source TermsJ* = 

Accident source telrms for a reprocessing plant were analyzed in GESMO, . 

and are quoted below. 

"Upper level accidents that may occur at separations facilities or Pu02 

conversion facilities include: 

Criticality y - " - • . 

High level radioactive waste concentrator or calciner explosion 

Plutonium product concentrator explosions 

Criticality Accident -_'• 

"A criticality accident is unlikely in a separations facility or Pu02 

conversion facility, because equipment and process limitations are designed to 

prevent such incidents. Safe spacing is assured in storage basins by physically 

spacing the fuel elements in storage racks in a safe array. Process systems 

and controls are designed to prevent an unsafe condition. Nevertheless, a 

*From GESMO, pages IV E-39 to IV E-42. 

61 
criticality accident of IO19 fissions is assumed. This yield is approximately 

an order of magnitude greater than the yield that has been experienced for Pu 

systems in past accidents. It is further assumed that all noble gases and 50% 

of the halogens (or halides) are discharged fromlthe plant stack. The dose 

commitments would be essentially the same for U02 fuel or MOX fuel. 

Waste Concentrator Explosion Accident 

"During operation of the separation facility solvent extraction process, 

solvent degradation products are generated and may be carried oyer into the 

waste streams. Under extreme conditions in early pilot plant operations", 

these nitrated degradation products (red oil) have caused concentrator explosions. However, red oil explosions can be prevented by installing equipment 

to eliminate the accumulation of organic materials in the waste, and by controlling the process temperature in the concentrator. 

"Concentrators are installed in highly shielded cells, having a volume of 

"about 100,00.0 cu ft (3,000 m 3 ) . In the unlikely event of an accident, the --' 

explosion is estimated to disperse about 150 gallons (600 liters) of high 

level radioactive.waste solution into the cell in the form of a finely divicted 

mist. A substantial portion of the mist would rain out or plate out on the 

cell surfaces. Droplets remaining in the air (10 mg/m3) would be carried 

through the ventilation ducts to the high efficiency filters. Moisture separators upstream of the filters would knock out most of the mist. 

"The j)lant ventjlation filters are located some distance from the separation-plant process cells. Most of the explosive energy would be expended in 

destruction of the concentrator. Pressures developed by the explosion would 

be dampened by expansion into the cell and would be further attenuated in the 

ductwork. Thefinal filters are not expected.to.be affected. 

"It has been estimated that plateout of the droplets oh the cell walls 

and floors and removal by the filtration system will result in a reduction in 

the fraction of material released to 3.6 x 10-8. Material leaving the final 

filter has been estimated to be 30.5 mg of high level radioactive waste solution 

in the form of an aerosol. 

62 
"Table IV E-16 identifies those nuclides that would contribute significantly 

to the offsite dose, and summarizes the offsite bone dose commitment that might 

result from this hypothetical accident. The maximum offsite dose commitment to 

an individual is estimated to be about 2.6 mrem (bone) for U02 fuep, and about 

6.9 mrem (bone) for. MOX fuels. 

"Table IV E-16 

WASTE CONCENTRATION EXPLOSION EFFECTS 

Nuclide 

i _ * 

Pu 

241 Am 

242CnT 

244 Cm 

9 0 Sr 

ioeRu _ _..-

144 C e " -

Other F.;P/ 

Half Life 

458y 

162d ": 

18y 

29y 

: iy - .. -

284dv 

' •-

Radioactivi 

in 

-

U02 Fuel 

0.02-

0.007" 

0.76 

0.05 

2.80 

2000. -

27. 

:l-5? 

Ace 

mC 

-

ty Released 

ident 

iM0X Fuel 

0.15 

0.05 

12.3 

3.25 

1.62 

3400. 

23. ; 

1.5 

Accident Bone 

Dose Contribution 

mrem 

U02 Fuel MOX Fuel 

0.01 

0.02 

0.04 

0.07 

1.92 

0.02 

0.04" 

0.48 

0.06 

0.17 

0.60 

4.38 

1.14 

0.04 

"TJ.03 

0.48 

Total 2.6 6.9 

"Table IV E-17..shows the radionuclide releases and the bone dose commitment to the maximally exposed offsite individual.... 

.

 v -•-- -. Table IV E-17 

"PLUTONIUM PRODUCT-EVAPORATOR EXPLOSION EFFECTS 

__ 

Plutonium 

Isotope 

238 

239 

240 

241 

- . 

- ' - . - --

Half Life 

86y 

-" 2.4 x I04y -

6,540y 

13y 

Radioactivity Released 

in Accident 

mCi -

U02 Fuel MOX Fuel 

1.02 

0.08 

0.11 

2.94 

2:il

 ; 

0.04 

0.13 

4.01 

Total 

End of 

Accident 

Dose Contr 

mrem 

Bone 

ibution 

U02 Fuel MOX Fuel 

5.94 

0,51 

0.76 

3.98 

11.2 

Quote 

12.27 

0.28 

0.88 

5.42 

18.9 

63 
• ^ 

Plutonium Concentrator Explosion Accident 

"The postulated explosion of a plutonium concentrator in the reprocessing 

plant is typical of upper level accidents by which plutonium could be released [ 

to a cell or glovebox area. Typically, the plutonium processing equipment tends 

to be smaller, and installed in smaller rooms (cells or gloveboxes) than the 

waste concentrator previously discussed. The release rate is derived by assuming 

that the room (cell or glovebox) atmosphere contains the same mass of aerosol 

(10 mg/m3) as the atmosphere of the waste concentrator cell. For a 1,000 m3 

plutonium concentrator cell volume, the postulated accident would release 

about 2.2 mg of plutonium. 

2.2.9.3 Calculations of Doses 

The doses below for these accidents are taken directly from GESMO, page 

IV E-40. 

i - "Maximum Offsite Individual Dose Commitment (rem) 

Accident PWR MOX Fuel 

Criticality 0.056 (thyroid) 

Waste Concentrator .Explosion 0.0069 (bone) 

Pu Evaporator Explosion 0.019 (bone) 

Fire 0.0135 (bone) 

End of Quote 

2.2.9.4 Implications-for Emergency Preparedness 

According to the GESMO analysis accidents at reprocessing plants would 

not cause doses in excess of protective action guides. This is. primarily due 

to lack of.strong driving forces and extensive containment systems. Thus. 

special emergency preparedness is not needed offsite. 

2.2.10 Research with Nuclear Fuels 

These facilities perform research and development related to nuclear 

power plant fuel manufacturing and testing. They use special nuclear materials 

64 
in forms ranging from powders to solutions, although larger quantities are 

usually in the form of fuel pellets. The work takes place in laboratories and 

glove box trains. Processes such as blending, crushing, milling, sintering, 

grinding, and solvent extraction may take place. 

2.2.10.1 Accident History . 

A number of the accidents previously discussed for other parts of the 

fuel cycle are relevant to research with nuclear fuels. For example, if the 

facility haridles large quantities of enriched uranium, criticalities as listed 

in Table 5 could occur, yIn addition, fires and other types of accidents such 

as those listed in Tables 6 and 7 could occur. Since these types of accidents 

have been discussed previously, they will not be discussed further here. 

2.2.10.2 Accident.Source Terms 

Sutter* bas analyzed nuclear, fuel research facilities and concluded- that " 

the potential-accidents arercriticality.accidents, spills.and leaks, tornados, 

earthquakes, .fires outside the facility, fires inside the facility, explosions, 

and fuel handling accidents.-7-~- "=-

If large quantities of material are handled "a criticality accident may be 

a possibility. If this is the case, the releases and offsite doses would be 

the same as those discussed in Section 2:2.5.2. for criticalities at fuel 

fabrication plants. "-."-" _ 

The other possible accidents_of" significance would be a major fire in-the 

facility such as a fire in a glovebox train.- If trie-fire does not break the 

glove.box filters and the final HEPA filters, the release to the environment 

would be negligible. However, if both sets.of filters are breached a considerable release is plausible. Assuming 13 glove boxes are in the train, 2 kg of 

material is present in each, and 0.1% becomes airborne due to the fire,:a 

release of 0.026 kg would be calculated. This would be uranium or plutonium, 

whichever was beirig processed, although at this time no licensees are handling 

such quantities of plutonium. 

*Sutter, Op. cit. 

65 
P.77 

2.2.10.3 Calculations of Doses 

Doses due to a criticality would be the same as discussed previously. A 

fire involving low enriched uranium (solubility class Y) would result in an 

effective dose equivalent within 100 m of 0.007 rem for F, 1 m/s meteorology 

and 0.001 rem for more typical D, 4.5 m/s meteorology. Inhaled quantities are 

well below levels where chemical toxicity is observed7 The dose calculations 

assumed that the building size was 25 m x 10 m, the release was at ground 

level, and the release duration was 30 minutes. 

If plutonium-239 were the fuel, rather than uranium, the effective dose 

equivalent from the fire would be about 67 rems fpr typical D, 4.5 m/s meteorology and 500 rems for conservative F, 1 m/s meteorology. The dose calculations 

for Uranium and plutonium consider an insoluble class Y compound. However, the 

plutonium dbse was'calculated assuming that plutonium facility would have the 

same containment capability as uranium facilities would be expected to have. 

Since superior containment is provided for plutonium, the offsite doses 

presented here are probably larger that could actually-be by quite a large 

margin. For .example, we assume 0.1% release for the involved plutonium. By 

comparison the actual release for the Rocky Flats fire discussed previously was 

10-8,"a hundred-thousand times lower2.2.10.4 Implications for Emergency Preparedness 

Criticality, as discussed previously, may require,some emergency preparedness. Other accidents at facilities handling uranium would not seem to require 

emergency preparedness. Plutonium processing, "on~the other hand could cause 

large doses offsite iff-not contained. We therefore conclude that for plutonium 

research and development activities the need for offsite emergency preparedness 

should be evaluated on a case by case basis. 

2.3 Byproduct Material Facilities 

There are six types of byproduct material licensees that handle large 

enough quantities of radioactive material not in sealed form so that need 

for offsite emergency preparedness should be considered. These are: radiopharmaceutical manufacturing, radiopharmacies, sealed source manufacturing, 

66 
P.78 

university research laboratories, waste warehousing, and fabrication of 

depleted uranium products. 

2.3.i (A Generic Overview 

We will consider each of these types of facilities separately rather than 

as a single group. The reason is that we Wish to determine whether accident 

scenarios and release fractions developed for each type can be generalized 

into a single set of accident scenarios and release fractions. But first, we 

will consider certain common or generic characteristics of byproduct material 

licensees. ' ' j . -.-_..."• - ." 

2.3.1.1 Accident History y 

The accidents involving release of byproduct material from all types of 

facilities handling byproduct material are listed below in Tables 11 and 12. 

The reason for combining all types of facilities=in these tables .is-that the 

experience of one type of facility may be relevant to other types of facilities. 

For example, a fire or-explosion-in a glovebox or hot cell "is an accident-that 

could happen at almost any type of major byproduct-handling facility. The relevant accidents listed in Tables 11 and 12 will be discussed in later sections. 

Overall, accidents involving byproduct material have led to small offsite 

doses. Releases have always been below the EPA's protective action guide lower 

limit of 1 rem. Thus, no emergency protective actions have ever-been necessary 

to protect people offsite from airborne releases of radioactivity,.--

2.3.1.2 Release Fractions for Accident Source Terms* •-..-

The release fractions "selected are given below along with the reasons for 

selecting them. 

Noble gases (1.0): Kr and Xe were assigned a release fraction of 1 because 

they are always gases at room temperature, they do not plate out, they are not 

retained by filters, and they do not react chemically to form less volatile 

compounds. 

*This section prepared with the assistance of Mark Halverson, Pacific Northwest Laboratory, who compiled the references upon which the release fractions 

are based. 

67 
Table 11. Fires and Explosions Involving Release of Byproduct Materials 

Date Facility Release Description 

4-23-50 

2-21-55 

10-8-59 

11-10^60 

12-29-60 

.4^5-61 

9-25-64 

4-20-69 

8-30-76 

Lawrence Radiation Mi 

tab, S,F. , CA - • 

AEC Contractor 

Mound Laboratory, 

Miamisburg, OH 

(Fuel R&D) 

Laboratory at 

Univ. of Calif. , 

Berkeley, CA 

University of 

Alabama _ 

(Laboratory) 

U.S. (location 

unspecified) 

AEC 

contractor 

Babcock and 

Wi-1 cox 

Hanford Site -

Plutonium 

Plant 

nor 

Mi nor 

Pp-210 

contamination 

in laboratory. 

39 Ci released 

onsite 

None 

Some 

1 mCi. Minor 

contamination 

2.5 g of 

americium 

Co-60 

contamination 

of incinerator 

of 10 mR/hr 

300 Ci Am-241 

Worker seriously 

contaminated. 

Negligible 

release to 

environment. 

Multistory research building destroyed by fire. 

Spontaneous ignition fire 

broke out in nitric acid 

saturated rags and paper 

in a contaminated waste 

storage area of a chemical 

process laboratory. 

Explosion in drybox 

disperses Po-210. Lab tech 

combined acetone wash with 

nitric acid solution. 

An overheated oil both 

started a fire in a processing cave handling curium. 

No exposures and no contamination outside cave. 

Fire caused loss of some 

radioactive materials. 

Radiochemistry building 

fire. 

Shipping container 

explosion. 

A bag of Co-60 contaminated 

paper was put:iri rionradio^ 

active trash arid burned in 

incinerator. 

Explosion of an ion . '-

exchange column containing 

100 g (300 Ci) of Am-241. 

1 to 5 Ci deposited on a 

worker's skin and clothing. 

5 mCi remained on body 

after initial washing. 

Inhalation uptake by lung 

estimated at 0.05 mCi. 

68 
Table 12. Accidental Releases of Byproduct Material Except Fires and Explosions 

Date Facility Release 

H-3, minor [ 

1-131 leaked 

from cask. 

Dose of 5 R/hr 

on surface 

Sr-90 

Description 

7-25-58 

8-24-62 

1-23-64 

1-15-67 

Los Alamos 

Scientific.Lab., 

Los Alamos, NM 

Phillips Petroleum 

Co., Idaho Reactor 

Test Site, Idaho 

Falls, ID 

Hanford Laboratory-,• 

Richland, WA 

Babcock and Wilcox, 

Apollo, PA 

Ir-192 onsite 

4-4-67 

9-27-68 

5-5_-69 ... 

8-6^70 

9-17-73 

5-2-74 

"Savannah River 

Laboratory, 

Aiken, SC 

U.S. (location 

not reported) 

-U.-5.. Naval Ammunition Facility 

Lawrence Radiation 

Laboratory, 

Berkeley, CA 

Rocky Flats, 

Golden, CO 

Savannah River 

Laboratory, 

Ailen, SC 

H-3 released 

through stack. 

Minor 

Low level 

waste 

contamination 

Minor quantity 

of Kr-85 

H-3 accidentally 

discharged 

through 

100-ft stack 

H-3 released 

to water 

H-3, 50 g 

H-3 gas escaped. 

Leaky shipping cask contaminated 

truck and cask. 

While workers were replacing 

an agitator on a waste storage. . 

"tank", convection currents -

carried contaminated vapor from 

the open top of the tank to the 

environment. 

Technicians cut into an Ir-192 

pellet in hot cell. Ventilation 

tion imbalance allowed Ir-192 to 

escape ta working area. 

An electric welder malfunctioned 

causing the failure of a.seal 

tube which released H-3 to -

stack. """•- - .. • ~ . 

A cask of canned waste was 

dumped into an underground 

caisson. Radioactive dust 

escaped contaminating the cask, 

its truck, and workers.. 

Released to atmosphere 

Automatic safety devices 

failed. 

Water not known to be contaminated with H-3 released to 

plant waste stream. 

Failure of a pipe fitting 

allowed H-3 to be discharged 

through stack. 

69 
P.81 

Table 12. (continued) 

Date Facility Release Description 

"Normal operating losses," Some 

food prepared riearby contained . 

H-3 in concentrations above 

EPA drinking water standard 

of 0.02 uCi/£ (4 mrem/yr for 

2 liter/day consumption). (This 

was not an accidental release, 

but is included because of the 

large quantity of material 

released.) 

Airborne Am-241 while renovating 

contaminated gloveboxes. 

Released during opening of 

2 capsules containing 30 Ci 

of S-35. 

Ir-192 surface contami nati on 

found in lab. Escape path from 

hot cells riot discovered. 

A stuck well logging source 

was cut by a drill bit during 

recovery operations. Because 

the leak went unrecognized, 

some Am-241 activity was 

tracked into homes and 

businesses. 

A 20 Ci Am-241 sealed source was 

accidentally cut open on a cut 

open"on a lathe during machining. 

Six employees were exposed. 

2 Ci Cs-137 sealed source cut 

open on lathe. Some CsCl powder 

spilled out. Shop contaminated: 

15 homes contaminated by Cs-137 

from, workers shoes and clothing. 

1978 -

calendar 

year 

9-1-80-

7-24-81 

11-19-81 

12-12-81 

8-27-82 

2-8-83 

9-13-83 

American Atomics 

-Corp, Phoenix, . AZ 

287,000 Ci 

of H-3 

New. England Nuclear, Am-241 inside 

N. Bill erica, MA _ plant 

New England Nuclear7 3.2 Ci of S-35 

N. Billerica, MA -

Tech/Ops, 

Burlington, MA 

Consol idation "Coal 

Company, Library, 

'-P-A' -"-•--." 

Gulf Nuclear, -Inc. 

Webster, TX y - -

She!wel1 Servi ces, 

Inc., Hebron, OH 

Ir-192 

surface 

Up to 0.5 Ci 

of Am-241 in 

drilling mud 

Am-241 inside 

plant 

Cs-137 on 

;workers shoes 

and clothing 

70 
P.82 

Volatile and combustible compounds (0.5): A release fraction of 0.5 was 

assigned to compounds of hydrogen, carbon, phosphorous, sulfur, and iodine. 

Compounds of these elements would be expected to be less than 100% due to 

these factors: (1) some of the compounds of the elements may form some 

novolatile compounds (ash) in a fire, (2)-some of the compounds of the.elements 

will plate but and deposit on internal surfaces, (3) some of compounds of the 

elements will be subject to retention by filters, (4) not all containers -"' 

possessed by a licensee would be likely to be breached in an accident, (5) at 

any particular time the actual inventory may be below the licensed possession 

limit, and (6) some particles formed may not be respirable. Consideration of 

site specific factors could cause Considerable reductions in the release 

fractions for specific facilities. •;...--• ' 

7 With respect to 1-131, experimental releases from a fire with the 1-131 

in a flammable solvent were 65%.. If the container was subsequently heated 

afterwords with a propane torch the release reached 83%.* Other factors as 

mentioned above would reduce the quantity that would actually escape the 

building. ' . 

Semi-volatile compounds (0.01): These include compounds of the elements 

selenium, mercury, cesium, polonium, bromine, tellurium, and ruthenium. 

Releases of Cs-137 at 1000°C have beerr measured at 1%/hr,* 1.5%/hr,** and 

4.2%/hr.t Additional removal by -filter retention and condensation was 

measured at 35 to 93% removal, and an additional 30% removal by deposition 

was measured.tt Based on the assumption of a 30-minute fire, an assumed 

release fraction of 0,01 is believed to be reasonable for cesium and selenium 

compounds. ^ " _ -

Unknown form: (but not" generally volatile "or combustible) <0.01): For use 

in screening analyses in which the chemical form of a radioactive material is 

not known, a release fraction of 0.01 is assumed. This value does not apply 

! 

*A. E.. Albrethsen.and L. C. Scheridiman, "Volatilization of Fission Products 

from High-Level Wastes," BNWL338, Pacific Northwest Laboratory, 1967. 

**W. J. Gray, "Volatility of a Zinc Borosilicate Glass Containing Simulated 

High-Level Radioactive Waste," BNWL-2111, Pacific Northwest Laboratory, 1969. 

tD. Walmsley et al, "Volatility Studies of Glasses for the Fingal Process, 

AERE-R-5779, England, 1969. 

ttR. K. Hilliard, "Fission Product Release from Uranium Heated in Air," HW-60689, 

Hanford Atomic Products Operation, 1959. 

71 
to Kr, Xe, H, C, P, or S, which were assigned different values above. It does 

not apply to U, Pu, Am, or Cm, which are assigned a release fraction of 0.001 

on the basis of their general form as nonvolatile powders as explained below. 

It also does not apply to Co, Ta, and Ir, which are assigned a release fraction 

of 0.001, as explained below.-

The 0.01 release factor for uriknown form should be used only for screening. 

For specific facilities, the actual chemical :form of the radioactive material 

could be used to determine an appropriate (generally lower) release fraction. 

Nonvolatile powders (0.001): Release fractions for nonvolatile compounds 

are given as the fraction of material released which is of respirable size. 

Most experiments report-total release fraction. To convert these to respirable 

release fraction, the release of particles larger than 10 microns was excluded 

from the respirable release fraction. Most reported experiments provide 

enough information to allow thisdetermination, but some do not. For experiments providing no information on particle size, it is necessary to estimate 

respirable release fractions from total release fractions by assuming that the 

respirable proportion will be similar to that reported in similar experiments. 

A release fraction of 0.001 was assigned to nonvolatile compounds in _ 

powder form. The mechanism is.not volatilization. Rather it is entrainment 

of the particles in .an airstream.. Even finely ground powders will generally 

contain less than a few percent of.the powder iri respirable size. The fraction of particles of respirable size is kept small by the difficulty in producing all small particles and subsequently by agglomeration and weathering, 

processes which cause micron-sized particles to. stick to surfaces, to larger 

particles, and to themselves. Particles larger"than respirable size (>10 

microns aerodynamic median diameter) quickly settle out of the air, and if 

inhaled seldom are deposited in the lungs. '-•_ -

The release fraction of 0.001 is suggested by a number of experiments 

generally designed to maximize the release.* Tbe experiments usually found 

*J. Mishima, L. Scbwendiman, and Radasch, ''Plutonium Release Studies III. 

Release from Heated Plutonium Bearing Powers," BNWL-786, Pacific Northwest 

Laboratory, Richland, WA, 1968. 

J. Mishima and L. Schwendiman, "Fractional Airborne Release of Uranium 

(Representing Plutonium) During the Burning of Contaminated Wastes," BNWL-1730, 

Pacific Northwest Laboratory, Richland, WA, 1973. 

S. Sutter, Johnson, and J. Mishima, "Aerosols Generated by Free Fall Spills 

of Powders and Solutions in Static Air," NUREG/CR-3093, 1981. 

72 
W41UB 

releases of respirable size particles of about 0.001 or less. In a few special 

cases designed to produce maximum releases, values above 0.001 were found. 

These conditions were for highly ground powders on certain flammable surfaces 

such as rubber or plexiglass, for high velocity air flow, or for highly pressurized releases in which all the material is violently thrown into the. air, 

which is then sampled before significant settling can occur. Such conditions 

are not corisidered representative of realistic accident conditions. 

Uranium metal and plutonium metal (0.001): These materials are pyrophoric. 

The release fraction of 0.001 is representative of experimental measurements 

described in Section 2.3.7.2. 

Nonvolatile solids (0.0001): For nonvolatile compounds in solid form 

rather than powder form, a release fraction of 0.0001 was assumed to reflect -

the lower amount of material that would be of respirable size. This value 

could be applied to cobalt, iridium, and tanalum in.solid form on a case by 

case basis. : 

Nonvolatile elements in flammable liquids (0.005). A release fraction Of 

0.005 was assigned to nonvolatile compounds in flammable liquids. 

Experiments with strontium in a flammable solvent yielded a release 

fraction of 0.002 from a fire.* . - . . 

Releases of uranium in flammable solvent averaged 0.00025, releases: of 

cesium averaged 0.0024, releases of cerium averaged 0.0065, and releases of 

zirconium also averaged 0.0065.** When the container in those experiments 

were subsequently heated with a propane torch uranium releases were 0.003, 

cesium releases averaged 0.006, cerium releases averaged 0.0074, and zirconium 

averaged 0.004.** _

 _ ~ 

For U02"powder in gasoline, release fractions were 0.0012.when airflow 

was 1.8 m/s.t Of this, 66% was less than 10 micron AEDft for a respirable 

release fraction of 0.0008. When airflow increased to 8.9 m/s, the release 

fraction increased to 0.013,t but only 7%tt was smaller than 10 microns, thus 

*S. L. Sutter et al, "Fractional Airborne Release of Strontium During the 

Combustion of 30% Normal Tributyl Phosphage in a Kerosine Type Diluent," 

BNWL-B-358, Pacific Northwest Laboratory, 1974. 

**J. Mishima and L. C. Schwendiman, BNWL-B-274, 1973, op.cit. 

tJ. Mishima and L. C. Schwendiman, BNWL-1730. 

ttJ. Mishima and L. C. Schwendiman, BNWL-1732. 

73 
the respirable release fraction was 0.0009, about the same as with the lower 

airflow. 

In other experiments on burning of kerosene-based solvent releases of 

Ru-106 we lie below 0.1% and releases of uranium and plutonium were much 

lower.* - -_'""•"'' 

Nonvolatile compounds in nonflammable liquids (0.001): Nonvolatile 

compounds in nonflammable liquids are assigned a release fraction of. 0.001. 

Several studies have measured releases in these circumstances. In general, 

release of these compounds can be expected to be small until the liquid is 

dried. After drying release fractions generally remain small because the 

material normally cakes dn the substrate or binds into particles too large 

to be respirable. 

"2.3.1.3 Quantities Requiring Consideration of Emergency Preparedness 

Table 13 lists the quantities requiring consideration of emergency 

preparedness based on the quantities needed to deliver a 1-rem effective dose 

equivalent offsite. The quantities Q. in Table 13 were calculated using thefollowing equation: 

Q>

 X 

1 RF i( H Ii * HGCi + HCSi> 

where Q. = the quantity of material i, curies 

" RF. = release fraction of material i. (Release fractions aire for a -

severe facility fire, which has been identified..as "the accident 

with the largest plausible release fractions.) 

HT- = the effective dose equiv.alerit from material i for the inhalation 

pathway-, rems/curie released. _.-..'...' 

*D. Whitney Tharin, Jr., "Burning of Radioactive Process Solvent, Savannah 

River Laboratory Report DP-942, Aiken, South Carolina, 1965. 

74 
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M MMMMMMMMMMMMMMMMMMMMMMMMMMMM' M 

t~\ r—s 

CO O 

ro • M 

IQ ro <*r 

xa \~r 

r^tt^s, 

co -P» 

O M 

- O 

o o 

o O TT 

IQ 

X - O 

xa ' 

O M O cn 

cn M co J> v i 

oooo o 

o 

o 

o 

2°2 ° _ £ ^ vIOMinOOOOO O 

222 2 ^ ° * ° t-1 OOOMO O o o oo o 

o oooioroJ>coororororoorororo^jroocoooooo o o o cocnooooooooooooo o 

3 » 

to to 

cr a. 

ro -*• 

-s 6 

_ i . Cu 

to n 

^— cr 

~ t . 

• < 

ro 

— i . /~ , 

3 1 

3- re 

cu 3 —• tn 

re v^ 

a. c 

v > O 

—*• 

> 

n co •i-l O 

Cg —< 

(/» c 

tn cr 

^. —^ 

- »^. r r 

<< 

-h ;rj 

-i re to —• 

o re 

rf to 

•_J. tn 

o re 3 

<-v, 0 

« C 

re oi 

-<• 3 

t a e* 

3" -•• 

rt- rr 

v>«< 

in 

—J. 

o 3 

-tt 

to 

o 

r r 

O 

-5 

O 

O 

tn 

re 

o 

o 3 

< re 1 

1 

- H 

to CT 

.—J 

re 

M 

co 

* 

o 

o 

3 

r r 

—t. 

3 

c 

re CL 

^^ 

/-vO 

n c 

c to 

1 3 

-^ nre -'• 

v > fiID v j ro 
Table 13. (continued) 

Radioactive 

material 

Dose conversion factor 

(rems/uCi 

inhaled) 

Solubility 

class 

Release 

fraction 

Quantity 

(weight) 

Quantity .[' 

(curie)_ 

Irradiated material-solid, 

noncombustible 

Mixed radioactive waste, 

beta gamma except 

1-125 and 1-131 

Packaged waste, 

beta gamma* 

Any other alpha emitter 

Contaminated equipment, 

alphaT 

Packaged waste, 

alpha 

0.001 

0.01 

0.001 

0.001 

0.0001 

0.0001 

10,000; 

1,000. 

10,000 

2 

20 

20 

*Type B packages assumed to have no release and no need for emergency 

preparedness. 

'GCi - the effective dose equivalent from material i for the ground 

_ contamination pathway, rems/curie released. 

'CSi the effective dose equivalent from material i for the cloudshine. 

pathway, rems/curie released. 

For all materials of greatest interest for fuel cycle and other radioactive material licensees, the dose from the inhalation pathway H,. will 

dominate the dose. . , .;_.- •_._.-' 

HT. was calculated by assuming-a maximum plausible intercept fraction for 

non-depositing (non-particulate) materials of 10-6. Thus, 

H^. = 10-6 x DCF 

where DCF = dose conversion factor, rems/curie inhaled. This intercept fraction was found to be the intercept fraction that would be calculated a Gaussian 

plume dispersion model, a distance of 100 meters on the plume centerline, 

atmospheric stability class F, 1 m/s wind speed, release duration of 30 minutes, 

building size of 10 m by 25 m, no other obstructions to spread the plume, and 

no plume rise due to buoyancy. 

78 
P.90 

According to this model intercept fractions would exceed IO-6 at distances 

closer than 100 meters from the release point. Such larger intercept fractions 

are not used because they are not considered physically realistic. Larger 

intercept fractions require the assumption that the smoke from a fire will not 

rise over the heads of people closer than 100 meters and that the people would 

Stand i.n dense smoke for the full duration of the release. These assumptions 

are riot considered realistic. 

Next, a deposition model was added to the meteorological model described 

above. A deposition velocity of 1 cm/sec was assumed.. This reduces the 

intercept fraction at-100 meters to 0.9 x 10-6, but adds a dose contribution 

from ground shine.-- Exposure to ground shine for 8 hours was assumed. 

Table 13 includes all radionuclides -listed-on 20 or more of NRC's approximately 9,000 licenses. A quantity for 1-129 is not included because saturation would prevent the thyroid from absorbing enough 1-129 to reach the 5-rem 

protective action guide dose. Table 13 also includes all beta-gamma emitters 

listed on_more than one but less than 20 licenses if the quantity that might 

deliver ah effective dose="equivalent-e_xce_eding 1 rem is less than 10,000 curies. 

Table 13 also includes all-alpha emitters-listed on more than one but less than 

20 licenses if the quantity that might deliver an effective dose equivalent 

exceeding 1 rem is less than 2 curies.. 

The quantities in Table.13 are different from quantities previously published in NUR£G-0767, "Criteria for Selection of Fuel Cycle and Major Materials 

Licensees Needing RadiologicaJ Contingency Plans," and Federal Register Notices 

with orders (46 FR 12566) and-an Advance-Notice of Rulemaking (46 FR 29712). 

The main reasons are: (-1) dosimetric models" from ICRP Publications 26 and 30 

have been used instead of the older models from ICRP Publication 2 and 

(2) release fractions have changed as a.result of further study. Meteorological 

models are slightly different, but this has a small effect in most'cases. The 

intercept fraction for inhalation of a material that does not deposit on the 

ground is 10-6 in both cases. For particulates, which do deposit on the ground, 

the dose per curie released can-be as much, as 11% lower for-radionuclides with 

negligible external dose to about 5 times higher for radionuclides whose major 

dose pathway is groundshine from material deposited on the ground.* The values 

in Table 13 include consideration of groundshine. 

*See Table 5.5 of NUREG/CR-3657 for contributions of each pathway. 

79 
P.91 

Quantities are also given for certain frequently occurring mixtures of 

radionuclides. 

Mixed fission products are assigned the generally applicable release fraction of 0.01 and a quantity of 1000 curies. This value is typical for fission 

products, assumes that short-lived iodines will not be present, and assumes 

that no single nuclide wiIT account for more than perhaps 5 to 10% of the 

activity. In particular, Sr-90 is likely to be the dominant contributor to 

dose and is assumed to be no more than 5 to 10% of the activity. 

Mixed corrosion products are assumed to be bound to surfaces and thus 

assigned the smaller release fraction of 0.001. Cobalt-60 is likely to be the 

dominant nuclide. The quantity assigned is 10,000 curies. 

Contaminated equipment, beta-gamma is likewise assumed to have a lower 

release fraction of 0.001 due"to bonding to surfaces. The activity assigned 

is 10,000 Curies. Sr-90 is assumed to be less than 10% of the activity. 

Irradiated material is assigned the same values as mixed fission products 

because the forms and nuclides are likely to be similar. However, solid noncombustible irradiated material is assigned a lower release fraction of 0.001 and 

an activity _of 10,000 curies. - . y 

Mixed radioactive waste, beta-gamma, except 1-131 and 1-125 is assigned 

aa activity of 1000 curies. Iodine-131 and iodine-125 are assumed not to 

dominate the hazard due to their short half-lives. If significant quantities 

of iodine would be in the waste, their quantities should be calculated 

separately. 

Packaged waste, beta-gamma is assumed to have one-tenth.the release 

fraction of unpackaged material due to an assumed protection provided by the 

packaging. The quantity assigned is 10,000 curies." -- " - -_ 

Contaminated equipment, alpha is assumed to have one-tenth the release 

fraction of alpha emitters in general due to bonding onto surfaces. An activity 

is 20 curies is assigned. 

Packaged waste, alpha is assigned an activity of 20 curies because the 

packages are assumed to provide some protection against release. 

Type B packaged waste is assumed to be a sealed source and assumed to have 

no significant release. 

80 
Table 17. University Research Laboratories: Maximum Possession 

Limits Release Fractions, and Doses Due to a Major Fire 

P.92 

Radioactive 

material 

Maximum licensed Release 

possession limit (Ci) fraction 

Effective dose 

equivalent, rems 

H-3 

C-14 

P-32 

S-35 

Ni-63 

Sr-90 

Mo-99/Tc-99m 

1-125 

1-131 

Xe-133 

Po-210 

Am-241 

Cm-244 

Cf-252 

3000 

10 

5 

5 

1 

0.5 

10 

8 

1 

10 

10 

0.5 

1 

0.1 

0.5 

0.5 

0.5 

0.5 

0.01 

0.01 

0.01 

0.5 

0.5 

I. 

OAn. 

0.001 

0.001 

0.001 

0,002 to 0.2 

"• 0 to 0.01 

0 to 0.04 

0 to 0.01 

0 

0 to 0.005 

0 

0.06 to 5.5 

0.01 to 1. ( 

0 _ . •• 

0.009 to 0.9 

0.003 to 0.3 

0.003 to 0.3 

0 to 0,01 

(child's thyroid). -

2.3.5.3 Calculations of Doses 

A range of calculated effective dose equivalents is shown in Table 17. 

The largest potential doses are from 1-125,; Po-210, and Am-241. The range _ 

reflects the likelihood that release fractions wilVie considerably below _' 

those shown in the table and a range of possible meteorology. 

2.3.5.4 Implications for Emergency Preparedness 

Offsite emergency preparedness does not appear necessary for university 

research laboratories because potential effective dose equivalents are low. 

2.3.6 Waste Warehousing and Burial 

Ih waste warehousing, radioactive material in containers (drums) is stored 

for a period of generally not more than six months. Drums may be opened in the 

waste warehousing operation. They are stored and then transported to a licensed 

waste burial ground. 

90 
P.93 

2.3.6.1 Accident History 

Accidents involving waste warehousing and burial have been minor, such as 

very small releases, leaking containers, and containers having surface contamination above regulatory limits, There have been no offsite airborne releases 

of significance. _ 

2.3.6.2 Accident Source Terms 

Warehousing. Since most radioactive waste is in metal drums, the potential 

for accidental releases is low. An event of significant magnitude to breach the 

drums would be required to make material airborne. Accidents 

considered are tornadoes, earthquakes, fires, and explosions. 

A major facility fire is seen as yielding the largest releases. The 

activities of radioactive materials becoming airborne were estimated by Sutter. 

These activities are listed in Table 18. -

Table 18-. Waste Warehousing Airborne Releases and 

Doses Due to a Major Facility Fire 

Radioactive Quantity Release Effective dose 

material present (Ci) fraction equivalent, rem 

0.004 to 0.4 

0.002 to 0.2 

0.01 to 1. . - _ '.-

0.002 to 0.2 

0 

4 to 400. (child's thyroid) 

0.4 to 40. (child's thyroid) 

H-3 

C-14 

P-32 

S-35 

Cr-51 

1-125 

1-131 

6200 

160 

160 -

120 

60 

280 

20 

0.5 

0.5 

0.5 

0.5 

0.01 

0.5 

0.5 

91 
Burial. Accidents during waste burial were analyzed in a draft environmental impact statement.* A major trench fire of two hours duration was determined to be the accident with the largest potential airborne release. Dropped 

and ruptured containers were determined to cause smaller releases. The reader 

is referred to the referenced environmental statement for the details of the 

assumed releases. -

2.3.6.3 Calculations of Doses 

Warehousing. A rangeof effective dose equivalents outside the building 

due to a major warehouse fire are also shown in Table 18. The range represents 

the uncertainty in doses that might result from a severe accident. 

Burial. Doses calculated for a major trench fire are given in NUREG-0782 

as 0.006 rem"whole.body and 0.03 rem to the lungs. While these doses were not 

calculated by exactly the same calculatio,nal techniques as others in this 

report, they are so low that there seems to be no need to recalculate them. 

2.3.6.4 -Implications for Emergency Preparedness ? - . 

Since" potential accident doses for waste burial are far below the EPA's 

protective action guides, burial does not.seem to require special offsite 

emergency preparedness. For warehousing, the radioiodines 1-125 and 1-131 may 

make some special emergency preparedness appropriate. 

*0ffice of.Nuclear Material Safety and Safeguards, Draft Environmental Impact 

Statement on 10 CFR Part 61 "Licensing Requirements for Land Disposal of 

Radioactive Waste," NRC Report NUREG-0782, Volume 2, Section 6.2.2, 1981. 

92 
2.3.7 Depleted Uranium Products* 

Depleted uranium is used to make a number of products: radiation shields 

for radioisotopes and x-ray machines, aircraft counterweights, armor-pierci'ng 

bullets, and artillery shells. Among the processing operations performed are: 

reduction of "green salt" (uranium tetrafluoride) to metal; melting and casting 

of the metal; welding; extrusion, cutting and etching; and machining." Since 

uranium metal turnings and powders will burn, fires are a potential accident of 

concern. 

2.3.7.1 Accident History '.-_•-. -

Table 6 listed fires and; explosions involving uranium and is relevant to 

fabrication of depleted uranium products. The releases and offsite contamination that resulted were negligible in all cases. 

2.3.7.2. Accident Source Terms .: 

Three potential scenarios during three operations suggest themselves-as 

potentially resulting in the largest airborne releases of uranium. First, a 

large quantity of molten uranium is handled during the melting and casting 

operation. Second, moderate quantities of divided uranium scrap, which can be 

more readily ignited than bulk pieces, can be stored outdoors under water in 

containers. Finally, large quantities of uranium in the form of depleted 

uranium munitions are stored by the military. 

Up to about 700 kg of molten uranium-could be poured during a casting 

operation. If an operational or equipment failure resulted in the release of _• 

the molten uranium and ihe loss of the inert gas cover, the uranium would 

oxidize rapidly and a fraction would be made airborne. Carter and Stewart 

(1970) experimentally measured the airborne release from molten uranium and 

*This section prepared with the assistance of Dr. Jofu Mishima, Pacific 

Northwest Laboratories. 

93 
"W 

( • • 

measured airborne release rates ranging from 0.005% to 0.3% depending upon the 

conditions - ignition and burning, melting, or partial disruption of liquid 

into droplets.* The potential airborne release from this scenario range from 

0.04 kg to 2 kg-of uranium. Castipg operations occur in enclosed facilities 

and some of the airborne material will be lost due to natural processes such . 

as gravitational settling or deposition on surfaces during its transport to 

the release, point from the facility. Many such facilities are equipped with 

particle removal devices such as filters which further reduce, the emission. 

Scrap metal such as turnings can be stored under water in metal cans. . 

Prezbindowski (1983) analysed such ah event postulated to occur outdoors.** 

-190 kg of uranium turnings in a 30 gallon metal drum were assumed to ignite 

and oxidize to completion. The airborne release was estimated to be 0.1% 

resulting in 0.190 kg being released to the environment. 

The potentially most serious- accident would involve a fire in a munitions 

storage bunker (igloo) holding a large quantity of various types of munitions 

(depleted uranium, high explosive, etc.). It is postulated that 12,000 rounds 

of a-105 mm depleted uranium cartridge could Lbe present. Each cartridge would 

hold 3.3 kg of uranium resulting inVtotal of 40,000 kg of uranium. Othertype's of combustibles such as wooden crates and pallets, paper based packing 

materials, etc. wouldalso be present;. If the material present were ignited, 

the fire would initially spread slowly until sufficient flammable vapors could 

be generated and flashover occurs. Once flashover occurs, the entire contents 

of the enclosure are involved. The fire soon becomes oxygen limited due to 

the limited accessibility of air. Eventually, the cartridges themselves would 

be ignited and, if containment is lost"; -the fire would burn more vigorously 

due to the greater availability of oxygen.

 _ 

Igloos are designed to vent: in a perferred direction which does not 

involve adjacent structures. The flammable-vapors released may well burn 

outside the facility due to the high vapor generation rate reducing radiant 

heat transfer to the materials inside. 

*R. F. Carter and K. Stewart, "On the Oxide Fume Formed by the Combustion 

of Plutonium and Uranium," Inhaled Particles III, Unwin Brothers Limited, 

The Gresham Press, England, 1970. 

**6. L. Prezbindowski, "Uranium Oxide Facility Safety Analysis Report," 

UNI-M-157-DR, United Nuclear Industries, Richland, Washington, 1983. 

94 
_ _ ^ Pj7 

There are two types of cartridges cases used for depleted uranium munitions 

metal and combustible. The depleted uranium portion of metal cased cartridges 

were ejected from and unaffected by the fire in a large-scale, outdoors test.* 

Depleted uranium from combustible-cased munitions i|n similar tests were not 

ejected from the fire and were almost completely oxidized (83%:and .85.2%).** 

Collection.and analysis of the residual material did not indicate a loss of 

uranium and air samples taken during part of the burning and all of the recovery 

period did not show any significant airborne release.*** 

Experimental studies measured the rate of oxidation and airborne release 

during oxidation at elevated temperatures (400 C to 1200 C) of the depleted 

uranium portion of large-caliber munitions in air and a 50% air-50% carbon 

dioxide mixture. The maximum airborne fractional release rate measured during 

the outdoor test using combustible materials as the heat source was 2.2 x 

10-6/min by weight. The material was primarily U3 08 . About 50% of the material 

had a by weight aerodynamic equivalent diameter of 10 microns or less. The 

three depleted uranium specimens oxidized an average of 44% during the three 

hour test. Tlie velocityof air passing around the samples was 2.23 m/s (5 mph). 

Similar but lower rates were measured in laboratory studies at various temperatures and atmospheres. 

For the worst case accident involving depleted uranium it is postulated 

that 40,000 kg of depleted uranium are involved in an igloo fire and that it 

is completely oxidized. It is also assumed that the material is combustiblecased although the igloo limit is for metal-cased. Based upon the experimental 

studies reported by Elder and Tinkle,**** the material would be completely 

*R. L. Gilchrist, G. B. Barker, and J. Mi shi ma,_" Radiological-and 

Toxicological Assessment of an External Heat (Burn) Test of the 105 mm 

Cartridge, APFSDS-T, XM-774," PNL-2670, Pacific Northwest Laboratory, 

Richland, Washington, 1978. 

**C. 0. Hooker et al, "Hazards Classification Test of the Cartridge, 120 mm, 

APFSDS-T, XM-829," PNL-4459, Pacific Northwest Laboratory, Richland, 

Washington, 1983; 

***J. C. Elder and M. C. Tinkle, "Oxidation of Depleted Uranium Penetrators 

and Aerosol Dispersion at High Temperatures," LA-8610-MS, Los Alamos 

Scientific Laboratory, Los Alamos, NM, 1980. 

****J. C. Elder and M. C. Tinkle, "Oxidation of Depleted Uranium Penetrators 

and Aerosol Dispersion at High Temperatures," LA-8610-MS, Los Alamos 

Scientific Laboratory, Los Alamos, NM, 1980. 

95 
P.98 

oxidized in 400 min. At the rate of 2.2 x 10-6/min, 35 kg would be made airborne of which half are respirable particles. The total airborne release of 

respirable depleted uranium thus would be 18 kg. 

2.3.7.3 Calculations of Doses 

The effective dose equivalent for an 18 kg release of depleted uranium was 

calculated to be in the range from 0.001 to 0.06 rem. Heavy metal poisoning of 

the kidneys is not a factor because the uranium would not be in soluble form. 

2.3.7.4 Imp!ications for Emergency Preparedness 

No special offsite emergency preparedness is necessary for depleted 

uranium products because doses are below protective action guides. 

2.4 Summary of Facilities to be Covered 

Fuel Cycle: The accident with the greatest potential hazard appears to 

be release of a large quantity_of UF6. The release of-ajlarge'quantity of UF6 

presents a chemical toxicity hazard. The greatest potential hazard is at UFg 

conversion plants where hot 14-ton cylinders are handled outside. NRC licenses 

two such plants. The rupture of a hot cylinder is quite plausible and could 

lead to multi-ton releases of UF6. The release would begin instantly. In 

such a case evacuation to several kilometers downwind would be appropriate for 

very calm weather conditions. The plume is easilydetectable-by sight and by 

smell at levels well below levels likely to cause injury. Thus people downwind 

would be able to see the plume coming and would beable to judge for themselves 

when they have reached an area of safety. A protection action strategy is 

discussed in more detail in Section 2.5. 

In an actual accident the release could be greatly diminished by spraying 

the release point with water or carbon dioxide. This is probably the most 

effective action that can be taken to mitigate offsite consequences. 

The release of low-enriched UF6 could also occur from smaller 10-ton 

cylinders and could occur inside a building, which would prevent escape of 

96 
much uranium. For the 9 such facilities licensed by NRC the appropriate 

response would be similar to that above but would be limited to distances of 

perhaps about half as great. I 

The release of high-enriched UF6 from hot cylinders may not require offsite 

response due to the small cylinder size used. 

The release of UF6 from cold cylinders does not require a response because 

the quantities released would be quite small. With regard to the heating of 

cold cylinders in a fire and subsequent release, a response similar to that for 

hot cylinders may be appropriate, but additional warning time would generally 

be available. 

For a criticality accident, the lower end of the protective action guide 

range could be exceeded for a person standing outside on the plume centerline 

outvto a distance'of about 250 meters from the release point assuming adverse 

weather conditions. The calculated doses are based on the assumption of a 

pulsating criticality lasting 8 hours and the person standing outside on the 

plume centerline for the entire 8 hours. Three additional licensees mightbe in this category. " _. " - _ 

An-appropriate emergency response to a criticality accident during adverse 

meteorology would be immediate sheltering and closing windows up to a distance 

of 250 meters downwind. This response should be accomplished within about 3 br 

4 minutes after the criticality occurs. The initial pulse is likely to be the 

largest one. Plume travel time to 250 meters would be about 4 minutes for low 

wind speeds and 1 minute for average wind speeds. There would be some added 

. de]ay due to holdup in the liquid solution and the building atmosphere. After 

the immediate sheltering response, evacuation could be considered as an alternative to further sheltering. 

Preparedness for such a response is recommended for the area within 200 to 

250 meters from high-enriched uranium arid plutonium processing. 

Plutonium facilities (none currently operating) may also need special offsite emergency preparedness due to airborne releases of plutonium. However, 

the GESMO analysis discussed in Section 2.2.6 would indicate that extensive 

preparedness may not be appropriate. 

Byproduct material and plutonium licensees: A total of about 48 Part 30 

or Part 70 licensees were identified who were authorized to possess one or 

more radionuclides in unsealed form in excess of the quantities in Table 13 of 

this analysis (i.e., effective dose equivalent could exceed 1 rem offsite or 

97 
BB p.1 

thyroid dose could exceed 5 rems). However, some of the licensees have little 

or no need to possess as much as they are authorized to and some can do an 

evaluation to show releases would be smaller than assumed. We estimate that 

most of the licensees will elect to lower their authorized possession limits, 

or perform an evaluation rather than submit a plan. Thus, the total number ofaffected licensees is likely to be about 10 exceeding the limit for plutonium 

and 7 exceeding the limit for other materials, specifically Am-241, Po-210, 

1-125, 1-131, H-3, P-32, Sr-90, Cm-242, and Cm-244. 

In all cases a fire would be the accident of concern; The appropriate 

emergency response would be to evacuate people from the immediate vicinity (at 

least 100 meters) so as not to interfere with firefighting and to shelter or -

evacuate everywhere else where smoke can be smelled. These actions should be 

taken within a few minutes. 

2.5 A Protective Action Strategy 

- The most important characteristic of the accidents discussed is that there 

is likely to be little or no warning time before releases start." The most 

important accidents, UF6 releases and fires, are likely to be controlled within 

roughly half an hour in a majority of cases. Thus releases are often likely 

to stop or be.greatly reduced within a half hour. 

2.5.1 The Initial Response 

Quick decisions and prompt actions are necessary. The goal should be to 

make decisions on protective actions and start implementing those decisions 

within 5 or 10 minutes of discovering the accident. 

The heart of an effective protective action strategy is predetermined 

protective action decisions. The decisions should be made, agreed upon, and 

written down prior to an accident because accidents of concern are likely to 

happen so. quickly that decisions on protective actions must be immediate to 

be effective. Thus, the licensee's initial notification of police and fire 

officials should include a recommendation on what protective actions are 

appropriate and the distances to which the protective actions are appropriate. 

98 
Response distances can be calculated in advance based on how much material 

is subject to release relative to the values in Table 13 and on simple meteorological observables. The values in Table 13 are based on a meteorology of 

stability.class F and wind speed of 1 m/s. Stability classes A. through E and 

higher wind speeds; would each cause doses to be reduced. The protective action 

strategy developed here accounts for that reduction. 

The wind speeds at the time of an accident can be roughly estimated from 

simple observations. The correlation, called the Beaufort wind scale, is shown 

in Table 19. 

.'. Table 19. Beaufort Wind Scale (Simplified) 

Mean Wind Speed at _ - ' 

10 meters height Description of wind 

0.9 m/s Light air - direction of wind shownby smoke drift, but 

not by wind vanes. 

2.4 m/s Light breeze - wind felt onface; leaves rustle; ordinary 

. • • - _ vane moved by wind. 

4.4 m/s - _=." . Gentle breeze - leaves and small twigs in constant \ 

. - -.-'•''J motion; wind extends light flag. . 

6.7 m/s Moderate breeze - raises dust and loose paper; small 

. . _' branches are moved.-

9.3 m/s - Fresh freeze- small trees in leaf begin to sway; crested 

wavelets form on inland waters. 

In addition, atmospheric stability classes can be adequately determined 

from simple observations: whether it is day or night, the strength of the 

sun, cloud cover, and wind speed-;" The relationship, developed by Pasquill, is 

shown in Table 20. 

It is now possible to relate doses compared to the dose for F, 1 m/s 

meteorology. This is done in Table 21. The comparison is based on 100 meters. 

The table is slightly conservative for greater distances.; 

- The next step is to develop response distances for different weather conditions. First, consider a licensee for which the quantity subject to release is 

5 times the amount in Table 13. This represents a typical licensee to which the 

proposed rule would apply. The maximum potential effective dose equivalent offsite for F, 1 m/s weather is thus 5 rems. Protective actions should be considered 

99 
Table 20. Pasquill's Relation of Turbulence 

Types to Weather Conditions 

A—Extremely unstable conditions 

B—Moderately unstable conditions 

C—Slightly unstable conditions 

D—Neutral conditions* 

E — Slightly stable conditions 

F—Moderately stable conditions 

Surface wind 

speed, m/sec 

<2 

2 

4 

6 

>6 

Strong 

A 

A-B 

B 

c 

C 

Daytime insolation 

Moderate 

A-B 

B 

B-C 

C-D 

D 

Slight . 

B 

C 

c 

D 

D 

Th 

or 

Nighttime 

in overcast 

2 % cloudinesst 

E 

D 

D 

D 

conditions 

s V, cloudiness 

F 

E 

D 

D . 

•Applicable to heavy overcast, day or night. -' ' 

tThe degree of cloudiness is defined as that fraction of the sky above the local apparent 

horizon that.is covered by clouds. 

Table 21. Plume Centerline Dose for Different Meteorologial Conditions 

£ Relative to F, 1 m/s " - • 

Surface 

wind 

speed, 

m/sec 

<2 

2 

4 

6 

>6 

Dayt 

Strong 

0.03 

0.02 

0.01 

0.02 

0.01 

-

ime insolati 

Moderate 

0.04 

0.03 

0.02 

0.03 

0.03 

on 

• Slight 

0.05 

0.05 

0.03 

0.04 

, 0.03 

Nighttime 

Thin overcast 

or > V 8 

cloudiness 

1.00 

0.21 

>• 0.06 

_ 0:04 

0.03 

conditions 

<

 3 / 8 

cloudiness 

1.00 

0.50 

0.11 

0.04 

0.03 

Day or Night 

Heavy overcast 

0.23 

0.12 

0.06 

0.04 

0.03 

where the dose under actual meterological conditions could exceed 1 rem. Downwind distances to which actions should be considered are shown in Table 22 

below. Note that the table and its terminology has been simplied for use by 

night watchmen, firemen, and policemen. Figure 1 (page 11) has been used to 

estimate distances. Distances are rounded off to the nearest 100 meters. 

Note that the only time that actions would be considered for more than the 

immediate vicinity of the plant (i.e., beyond 100 meters) occur at night during 

light breezes. We would therefore recommend the following protective actions. 

In general, the fire and police departments should evacuate to a radius of very 

100 
Table 22. Downwind Distances to Which Protective Actions Should Be 

Considered for an Accident Involving Five Times the Quantity 

of Material in Table 13 

-

Wind 

Very light 

breeze - direction of wind 

shown by smoke, 

dri ft. 

Light breeze -

wind felt on 

face, leaves 

rustle. 

Gentle breeze -

leaves and twins 

in constant 

aotion; wind 

extends light 

flag. Or any 

stronger wind. 

Day 

(except 

heavy 

overcast) 

0 

meters 

0 

meters 

0 

Night 

Thi n overcast 

or 50+% Less than 40% 

cloud cover cloud cover 

400 400 

meters meters 

100 200 

meters meters 

0 0 

-

Day or Night 

Heavy overcast 

100 

meters 

0 

meters 

0 

roughly 100 meters. This would.be an appropriate action for any large fire to 

prevent onlookers from interferring with firefighting efforts. 

At night protective actions to 200 meters or 400 meters downwind would be 

recommended for light winds. Evacuation or sheltering are both appropriate 

responses. The local authorities should select one or the other based on their 

practicalities at the time. Sheltering has. an advantage in that most people 

would already be indoors at night. For sheltering to be effective, doors and 

windows should be closed and furnaces and air conditioners should be shut off. 

Upon termination of the release, doors and windows should then be opened to 

allow any radioactive materials that have gotten in to blow out. Evacuation 

of the area is also an acceptable alternative, but if it cannot be done within 

a few minutes it may not be as effective as sheltering. 

Downwind distances calculated in this manner should be taken as approximate guidelines, not as rigid criteria. In an actual situation firemen and 

policemen should notice where smoke is blowing and take reasonable steps to 

101 
prevent anyone staying in dense smoke for an extended time. In addition, since 

the upper ranqe of the protective action guides is never exceeded, the protective actions r[ecomended above should be considered optional, not mandatory. 

Now consider a licensee for which the quantity of material subject to 

release is 50 times the quantity in Table 13. This size release probably 

represents a practical upper bound for an accident at any currently licensed 

facility. The following table of downwind distances can be calculated. 

Note that in the daytime a rather modest response is appropriate, except 

for heavy, overcast days with light breezes. At night with very light breezes, 

protective actions should be considered to 1500 meters downwind. The table 

also shows distances downwind to which doses might exceed the upper end of the 

protective action guide ranges. 

102 
P.6 

Table 23. Downwind Distances to Which Protective Actions Should be 

Considered for an Accident Involving 50 Times the Quantity 

of Material in Table 13 

Night 

Wind 

Strong 

Sun 

100m 

Day 

Moderate 

Sun 

200m 

Slight 

Sun 

200m 

Thin .. 

overcast 

or 50+% 

cloud cover 

- 1500m 

(600m)* 

Less, 

than 40% 

cloud cover 

1500m 

(600m) 

Pay 

or Night 

Heavy 

overcast 

700m 

(200m) 

Very light 

.breeze - direction of smoke 

shown by smoke 

drift. 

Light breeze 

wind felt on 

face^ leaves 

rustle. 

100m 100m 200m 600m 

(200m) 

1000m 

(400m) 

400m 

(100m) 

Gentle breeze -

leaves and twigs 

in constant 

motion, wind 

extends light 

flag. 

Moderate breeze • 

raises dust and 

loose paper, 

small branches 

are moved. 

100m 100m 300m 400m 

(100m) 

300m 

100m 100m 200m 200m 200m 200m 

Moderately, 

strong breeze -

small trees in 

leaf begin to 

sway. 

100m 100m 100m 100m 100m 

*Distances in parenthesis are distances to which the 5-rem upper end of the 

protective action guide range .could be exceeded. 

103 
P.7 

I -. 

We have not defined an emergency planning zone for either the plume exposure pathway or the ingestion pathway as is done for nuclear power plant 

emergency planning. Instead we have calculated response distances. 

The concept of the planning zone has no practical application for the 

smaller accidents that might occur for these types of facilities. In general, 

the response would be under the direction of the local fire and police departments just as similar industrial accidents are handled. It is only necessary 

to identify the organizations that will be notified and will respond rather 

than a geographical area for which planning would take place. 

2.5.2 The Relocation Phase 

After the initial evacuation or sheltering, the next step is to relocate 

people out of "hot spots" caused by deposition of gamma ray emitters. The 

initial response described above included the assumption of exposure to ground 

shine for 8 hours in determining the distances to which a person might.receive 

a dose of 1 rem effective dose equivalent. The relocation phase is concerned 

with doses received; from ground shine for longer.periods of time. 

Surveys should-begin near the release point and;work outward so that the 

highest dose levels are likely to be found earliest, there are only a limited 

number of radioactive materials with the potential to case doses exceeding 1 rem 

within 24 hours outside the initial response area. These are: Na-22, Na-24, 

K-42, V-48, Mn-56, Cu-64, Tc-99m, Ru-105, Sb-126, and W-187. However, none of 

these materials is currently possessed by any licensee in quantities such that 

ground shine could exqeed 1 rem within 24 hours. 

As a practical matter, ground shine would not be expected to exceed 

40 mR/hr from any accident at a licensed facility, except possibly quite near 

the release point where fallout of very large, heavy particles may occur. 

"Hotspots" thus should be found within 8 hours inside the initial response 

area and within 24 hours outside. These times are believed to be practical 

and also adequate to allow taking protective actions in a timely manner. 

Since there are no official guidelines on dose rates at which people 

should be relocated pending decontamination, we suggest here a value of about 

2 mR/hr to be roughly consistent with NRC's concept of restricted areas and 

unrestricted areas. Under normal circumstances licensees must keep doses to 

people in unrestricted areas to less than 2 mrem in any one hour. Lower or 

104 
higher values could be used in specific circumstances. Lower values may be 

appropriate for more sensitive members of the population, such as residences 

of preschool age children or pregnant women. Higher values could be used if 

occupancy factors would be low or there were other restraints to relocation. 

For example, highways or parking lots could be used at higher dose levels 

based on low occupancy factors. Critical industrial processes could be 

maintained at higher dose levels based on economic needs, and the homes of 

people not wishing to relocate could be used based on the concept of informed 

consent. 

2.5.3 The Assessment Phase 

Soon after the release has ended, usually within ha]f an hour to an hour, 

the assessment phase of the emergency response should begin. The public will 

be concerned about whether the danger is over, whether they were exposed, 

whether they are contaminated with radioactivity, whether their homes are 

contaminated, with radioactivity," and what should they do. The news media will 

want to know what happened. Both the licensee and the NRC must be prepared to 

respond to such concerns promptly or suffer damaged reputations, ill-will, and 

possibly lawsuits. 

As discussed above, field measurements to locate ground contamination will 

be underway. 

The question of whether people were exposed can only be answered with 

abundant field measurements. Generally, direct measurements of radioactive 

material concentrations in the plume will not be possible because of a lack of 

time to prepare and a lack of ability_to locate and-follow the plume. Nevertheless, plant personnel should attempt such measurements with available instrumentation to the extent possible because if they are fortunate enough to get 

data, the data will be valuable. 

The answers of whether or to what degree people were exposed are likely to 

be best answered by measurements of ground contamination arid bioassay measurements. Ground contamination measurements will allow a direct determination of 

whether there was a release. They will also allow a quick rough order-ofmagnitude estimates of the time-integrated exposure to people in the areas. The 

time-integrated exposure is equal to concentration x in uCi/cm3 times time t in 

seconds: 

105 
P.9 

* GC X X t = j -

vd 

where GC is the ground contamination in uCi/cm2 and the deposition velocity 

V . may be taken to be 1 cm/s. This value for the deposition velocity should | 

normally provide conservative estimates of concentrations and dose, especially 

for distances within a few hundred meters of the release point. 

the effective dose equivalent H in rems to an individual during plume 

passage is: 

H = DCF x B x x x t x IO-6 

where H = effective dose equivalent from inhalation, rems 

DCF = dose conversion factor from Table 13, rems/curie inhaled 

B = breathing rate, 266 cm3/s 

X x t = concentration times time, uCi's/cm3 

10-6 = a factor to convert uCi to curies. 

At this point tlpse can be written directly in-terms of the measured ground 

contamination GC. The dose conversion factor DCF for the specific radionuclide 

is given in Table 13: --"-.'.- - . _' ' .-' -

H = 2.66 x 10-4 x DCF; x GC 

This equation provides a fast method to roughly approximate a person's 

internal dose directly from measurements, of ground .contamination. 

The quantity of material released xan be roughly approximated from ground 

contamination measurements as. well, the equation can be derived by similar 

means. 

Admittedly, this approach will be subject to Very large uncertainties. 

However, there is not likely to be any better means, to estimate the size of 

the release for at least several days. 

In addition to the dose estimates described above, bioassay measurements 

may be desirable whether a release occurred or not because there may be no way 

to prove that there was no release. Nasal swipes and skin contaminations are 

106 
P.10 

the easiest to obtain and help answer the question of whether a particular 

person was exposed. Urine samples can provide more sensitive and quantitative 

estimates of dose. Fecal sampling and in vivo counting can supplement the 

above methods. 

2.5.4 the Food Pathway 

From experience we know that the contaminated area will be primarily near 

the plant and will rapidly drop off with distance, the job of a monitoring 

team would be to go wherever necessary. We can define the monitoring capability 

needed. We cannot define?in advance an actual area that could beaffected. 

3. VALUE/IMPACf 

3.1 Alternatives 

* 

Three alternatives have been identified: (1) adopting aregulation 

containing the proposed "requirements, (2) imposing the requirements by license 

condition rather than by regulation, and (3) imposed no new requirements with 

regard to emergency planning. The first two alternatives would have essentially 

the same value and costs. Those values and costs are discussed below. The 

third alternative, no new requirements, would have essentially no value or 

costs. 

3.2 Value of the Proposed Action 

Value can be expressed in terms of risk reduction. 

Consider a release in which the effective dose equivalent at a distance 

of 100 meters is 5 rems, assuming Class F atmosphere stability and a wind speed 

of 1 meter/sec. The area over which the 1 rem protective action guide would be 

exceeded would be 0.006 square miles. If this area contained people at the 

average population density of the continental United States (72 people/square 

mile) it would contain on the average about half a person. 

107 
However, the facilities under consideration are usually located in built-up 

areas. A survey of potentially affected licensees shows that typical population densities are about 3,000 people/square mile.* Thus perhaps about 20 people 

would typically be in the area. Generally about 80% of people are in the buildings so about 4 people would be outdoors and 16 would be indoors. " 

The average dose to a person outdoors in the area was calculated to be 

about 3 rems. (This value was calculated for an open field; doses in urban 

areas would be less, but we ignore that factor.) In addition, doses to people 

in buildings would be half this dose because of protection provided by the 

building . • , "."•'•• - . 

the total collective dose for the urban area thus might be 40 person-rems. 

Protective actions that would be available to these people are primarily 

evacuation and sheltering. Evacuation would be. the more effective if it could 

be done promptly before the plume arrived. Sheltering would often be more 

practical because it can be done faster and most people are already inside. 

Very roughly the dose savings, due to protective actions for this adverse 

meteorology could be put at 20 person-rem, about half the potential dose. . 

Assuming that the chance of death due to cancer due to doses of several rems 

is in the range between zero to 10-4 cancer deaths/rem. the expected.number 

of lives saved due to the protective actions would be less than 0.002 for 

adverse meteorology and roughly 0.00002 lives averaged over all meteorology. . 

Our estimate of the frequency of a major release is less than 10-4/year\. 

insurance statistics available from insurance companies dealing in commercial 

-structures.indicate fire losses in unsprinklered commercial and industrial 

facilities to be 0.006/year. Sprinkler failure rates are estimated to "be 

0.038 should a fire occur.** The fire loss rate for a sprinklered facility 

should thus be roughly 2 x l0-4/year. It is now assumed that additional site y 

specific factors will reduce the probability, of a release to roughly 10-4/year. 

*J. P. McBride, "Economic Consequences of Accidental Releases from Fuel 

Fabrication and Radioisotope Processing Plants," NUREG/CR-0222, 1979, 

Appendix A. 

**National Fire Protection Association, Fire Protection Handbook, 14th edition, 

Table 14-19, page 14-5. 

108 
Examples of such factors would be: material kept in fireproof storage thus 

preventing significant release, filter system does not fail thus preventing 

release, or firemen extinguish fire before radioactive material is heated hot 

epqugh for a significant release, and many more. Thus protective actions 

could be expected to save 0.00000002 lives per year per facility. 

If a life is given a value of ten-million dollars, the value of protective 

actions at a typical site in an urban area is $0.20 per facility per year or 

less. 

Now consider a release in which the effective dose equivalent at a distance 

of 1000 meters is 5 rems assuming Class F atmospheric stability and wind speed . 

of .1 meter/sec. At this level early injuries have stillbeen avoided. The 

area over which the ~1 rem protective action guide would be exceeded would be 

0.15 square miles. For a typical built up site this area would contain about 

450 people.y 

The average dbse to a person outdoors in the area was calculated to be 

about 3 rems and to a person indoors was calculated to be 1.5 rems. The total 

collective dose assuming some people are indoors as previously discussed might 

be 800 person-rem. The dose savings due to protective actions could be about 

half of this or 400 person-rem. Lives saved due to these protective actions 

Could be ap to 6.04 life for adverse meteorology and 0.04 life for average 

meteorology. 

With the frequency of a major release at 10-4/year, protective actions 

might perhaps save up to 0.000004 life per year per facility. 

_ If a life" is given a value of ten-million dollars, the value of protective 

actions in a-densely populated urban area is $4 or less per facility per year. 

3.3 . Cos t - •_"--•-.""•• " 

For the smaller class of accidents, those exceeding 1 rem offsite but not 

5 rems at 1000 meters, the licensee is considered to have a 50-page plan telling 

whathe would do in the event of emergencies such as fires. 

Cost data were obtained from two radiopharmaceutical manufacturers. Both 

licensees calculated the cost of the onsite contingency plans required by order. 

The manufacturer with a small program and limited facilities estimated the 

initial set-up cost $84,000. Annual operating costs were estimated to be 

109 
P. 13 

$18,000. Labor accounted for 1/2 to 2/3 of the cost in each category. Labor 

was given a value of $30/hour with no overhead charged. The main equipment 

costs were for radios, extra monitoring equipment for emergency use, and extra 

respirators. The largest annual expense is for training. Other operating 

expenses are for audits[, drills, and equipment replacement and maintenance. 

To place all expenses on an annual-basis the initial set-up cost was 

divided by iO, assuming a ten-year useful life of a plan. Thus annual costs 

are estimated to be $26,000/year/facility for this radiopharmaceutical manufacturer with a small program. 

A second radiopharmaceutical manufacturer with one of the largest programs 

that would be covered by the regulation reported that the cost of establishing 

their on-site radiological.contingency plan was more than $550,000. No annual" 

operating costs were given. 

Assuming a 10-year plan life and operating costs of $18,000/year (the 

estimate of the other manufacturer) the total annual cost is $73,000/year/ 

facility for large facilities. 

Costs to NRC to review and inspect plans have been estimated to be 

$4,000/year/facility. 

3.4 Value/Impact of Alternatives 

The costs of emergency preparedness are expected to exceed the benefits in 

terms of protecting public health and safety as shown below. 

Table 24. Comparison of Costs and Benefits of Special 

Emergency Preparedness 

Size of Licensee . V Cost Benefit 

Small - Possessing"5 times quantity in Table 13 $30,000/yr $0.20/yr 

Large - Possessing 50 times quantity in Table 13 $77,000/yr $4/yr 

110 
P.14 

4.0 STATUTORY CONSIDERATIONS 

4.1 NRC Regulatory Authority 

The Atomic Energy Act, gives NRC authority to adopt regulations for protecting public health and safety. This proposed rule would be justified under 

that authority. 

4.2 Agreement State 

The" question is whether NRC's Agreement States would adopt offsite emergency preparedness requirements similar to NRC's. 

The NRC's Office of State Programs intends to make this requirement a 

matter requiring compatability. Thus, NRC would require that Agreement States 

adopt requirements similar to NRC's. 

4.3 Environmental Impact Appraisal 

The NRC's regulations [10 CFR § 51.5(b)] require that substantive and 

significant amendments (jfrom the-standpoint of environmental impact)-of regulations require an environmental impact statement. 

To make the .finding that amendments are not substantive and significant 

from the standpoint of environmental impact, NRC regulations [10 CFR § 51.5(c)(1)] 

require the preparation of a negative declaration and an environmental impact 

appraisal. 

The environmental impact appraisal roust include [10 CFR § 51.7(b)]T 

(1) A description of the proposed "action. 

(2) A summary description of the probable impacts of the proposed action 

on the environment. 

(3) The basis for the conclusion that no environmental impact statement 

need be prepared. 

Ill 
The proposed action is a rule to require emergency procedures for offsite releases. A description of the proposed requirements is contained in 

Section 1.1, "Description of the Proposed Action." 

A summary description of the probable impacts of the proposed action on 

the environment is contained in Section 3.2, "Value of the Proposed Action." 

The basis for the conclusion that no environmental impact statement need 

be prepared is that the benefits to public safety are neither substantive 

'nor significant as described in Section 3.2, "Value of the Proposed Action," 

and summarized in Table 24. 

5. CONCLUSIONS 

The conclusion of this Regulatory Analysis is that accidents at fuel cycle 

and other radioactive material licensees pose a very small risk to the public. 

Serious accidents are infrequent and would generally involve relatively small 

radiation doses to few people located in small areas. 

This is not to_say that radiation doses large-enough to exceed guides for 

taking protective actions cannot occur. It may. be possible to have an accident 

at some licensed facilities which would cause offsite doses exceeding protective 

action guides. However, offsite radiation doses large enough to cause acute 

fatality or even early injury from an airborne release are not considered 

plausible. 

For a licensee possessing 5 times the amount of material in Table 13, we 

conclude that protective actions in an urban area-might save.up to 0.00000002 

lives per year per facility. Perhaps about 20 to 30 licensees have a possibility of such an accident or worse. For these iTacilities we recommend there 

should be notification of local authorities. However, no special facilities, 

equipment, or other resources for responding are considered necessary. 

For a licensee with 50 times as much releasable material as in Table 13, we 

conclude that protective actions in a built up area might save up to 0.0000004 

lives per year per facility. There may be 2 or 3 licensees with a capability of 

an accident this severe. 

The cost of this preparedness may not be justified in terms of protecting 

public health and safety. Rather we would justify it in terms of the intangible 

benefit of being able to reassure the public that if an accident happens local 

authorities will be notified so they make take appropriate actions. 

112 
PJ6 

xM__m_____mmmxWxmmm, 

Although emergency preparedness for fuel cycle and other radioactive 

material licensees cannot be shown to be cost effective, the NRC feels that 

such preparedness represents a prudent step which should be taken in line with 

the NRC's philosophy of defense-in-depth, to minimize the adverse effects 

which could result from a severe accident at one of its facilities. 

113