1. BASIC PROFESSIONAL TRAINING COURSE Module II Radiation protection in nuclear facilities Case Studies
Version 1.0, May 2015
This material was prepared by the IAEA and co-funded by the European Union.  1

2. INTRODUCTION
Lecturer should decide if each student works on all cases or each student is assigned a certain number of cases of each topic.
If more examples are included in particular case, lecturer should distribute them among participants.
Each study case solution should be presented on board, with the procedure and proper interpretation.

3. Case 1
For following sources of radiation write down proper characteristics (natural/artificial source, ionising/nonionizing radiation):
GSM phone,
microwave oven,
cathode tube TV set,
radioactive level gauge in cement industry,
uranium ore in the ground,
uranium mine where (natural!) uranium ore is dig,
granite rock,
load of soil on the truck,
medical X-ray tube,
400 kV power line,
electric shaver,
kitchen blender,
Sun,
Universe,…

4. Case 2
For each of the following sources, describe interaction/interactions of emitted radiation with matter: 60Co, 137Cs, 90Sr, 3H, 241Am, 40K, 99mTc, fresh nuclear fuel (UO2), uranium ore, reactor core (operating), reactor core (after reactor shutdown).
List type/types of emitted radiation (alpha, beta, gamma, neutrons) for particular source;
Describe mechanism through which radiation interacts with matter (if more types of radiation are emitted, a participant should explain et least one type);
State whether the radiation is directly or indirectly ionising;
Explain the meaning of range or half-value layer (for alpha, beta, gamma, whichever is applicable);
For neutron radiation explain how interactions are different from interactions of other radiations and how interactions depend on neutron energy.

5. Case 3
Describe the construction of a coaxial gas detector and consequences of interaction of ionising radiations with detector’s internals;
Draw the current-voltage characteristic of a gas detector;
Mark ionisation chamber region, G-M region, and proportional region;
Describe characteristics of operation in a particular region;
Explain the need for detector window;
Explain the meaning of dead time;
Discuss the possibility of recognising signals from different radiation types in a gas detector;
Discuss the usability of particular types of gas detectors for measuring high gamma fields/low gamma fields/surface contamination with beta/gamma radionuclides.

6. Case 4
Describe the construction of a scintillation detector and functions of detector’s construction parts;
Describe the timeline of events occurring after entrance of radiation in the scintillation detector;
Explain the importance of the fact that different types of interaction for gamma radiation are possible inside of the scintillation part of the detector (i.e. the photoelectric effect, Compton scattering, pair production). Is there any practical use of these facts?
Considering the fact that scintillation part of a detector could be made of solid crystals or special plastic materials, what would be the main advantages of scintillation detectors over gas detectors?

7. Case 5
Thermoluminescent and optically stimulated luminescence detectors are small “pills” of special material that can “record” radiation.
How we extract the information from these detectors?
What information about radiation is available (and what information is not available) when these detectors are used?

8. Case 6
Usual gas detectors are not sensitive to neutron radiation (why?). What approach is used to circumvent this deficiency and to make gas detectors sensitive to neutrons (thermal neutrons in particular?)
Thermal neutrons entering the BF3 detector have energy below 1 eV, and gamma rays have much higher energy (usually much higher than 100 keV). But neutron pulses from BF3 detector are much higher than gamma pulses. What is the reason for this difference? What can we say about neutron energy if we analyse the neutron signals?

9. Case 7
What would be effective dose in an hour from 1 kBq of 131I in the thyroid?
(Consider that the mass of the thyroid is 20 g and average energy of emitted beta particles is 191 keV. For the sake of simplicity, consider that all beta particles are absorbed in the thyroid, and all gamma rays escape from the body.)

10. Case 8
A stainless steel screw (SS304, mass 10 g) has fallen in research reactor pool and rested on reflector belt near the reactor core. The screw was discovered an hour before end of the shift and possibility of removal was discussed. It was estimated (assumption: 4 h irradiation at 1011 cm-2s-1, 2 h delay), that the activities of gamma emitting radionuclides in the steel are following:
51Cr: MBq (27.7 d, Γair = 5.45·10-15 Gy·m2 ·h-1·Bq-1);
56Mn: 1.1 GBq (2.58 h, Γair = 2.2·10-13 Gy·m2 ·h-1·Bq-1);
59Fe: MBq (44.53 d, Γair =1.57·10-13 Gy·m2 ·h-1·Bq-1);
65Ni: 5.3 MBq (2.52 h, Γair = 8.0·10-14 Gy·m2 ·h-1·Bq-1); 
Immediate removal would imply manipulation of the activated screw with tongs for 30 seconds.

11. Case 8 (Cont.)
What would be the absorbed dose rate in the air 𝐷 at 25 cm from the screw and 75 cm from the screw and corresponding absorbed dose rates in the tissue?
What would be equivalent gamma dose to the skin of a worker handling the screw with 25 cm tongs?
What would be effective gamma dose considering that distance between the screw and the body is approximately 75 cm?
What would be effective gamma dose for the same operation, if operators postpone screw recovery until next morning (i.e. for additional 16 hours?)
Considering that 56Mn, 65Ni, and 59Fe are also strong beta emitters (in fact skin doses from beta are much higher than doses from gamma), what additional personal protective equipment would you recommend for the recovery of the screw?

12. Case 9
Dose rate measured at the distance 5 m from an industrial radiography source was 15 mSv/h.
What would be dose rate measured at 10 m, 1 m, and 0.1 m?
What would be the distance where dose rate would be 10 µSv/h?
What would be the distance where the dose rate from the source is equal to natural background (i.e. 0.1 µSv/h)?
How long would it take to exceed effective dose 5 mSv at the distance 10 meters?
How long would it take to exceed annual effective dose limit at 1 m (for occupationally exposed workers)?
How long would it take to exceed annual equivalent dose limit to the lens of the eye (for occupationally exposed workers) at 1 m?
How long would it take to exceed annual equivalent dose limit to skin (for occupationally exposed workers) at 10 cm?

13. Case 10
Given the dose coefficient for inhalation of 131I (133I, 134Cs, 137Cs, 90Sr, 58Co, 60Co, 239Pu) and considering 2000h / 1.2m3/h exposure:
Calculate ALI (Annual Limit of Intake);
Calculate DAC (Derive Air Concentration);
Radionuclide
h(g) [Sv/Bq]
131I
7.6 · 10-9
133I
1.5 · 10-9
134Cs
6.8 · 10-9
137Cs
4.8 · 10-9
90Sr
1.5 · 10-7
58Co
2.0 · 10-9
60Co
2.9 · 10-8
239Pu
4.7 · 10-5

14. Case 10 (Cont.)
If concentration of 131I (133I, 134Cs, 137Cs, 90Sr, 58Co, 60Co, 239Pu) in the air is 1000 Bq/m3, what would be required Protection factor (PF) for respiratory protection equipment to achieve concentration lower than 0.1 DAC in inhaled air?
If there is a mixture of radionuclides 131I, 133I, 134Cs, 137Cs, 90Sr, 58Co, 60Co, and 239Pu in the air, and each of these has concentration 100 Bq/m3, what would be required Protection factor for the case?
For each of the listed radionuclides calculate the committed effective dose acquired in an hour, if the concentration of radionuclide in the air is exactly 1 DAC.

15. Case 11
One of methods for reduction gamma dose rate is also introduction of shielding. For a certain gamma ray energy, effectiveness of shields is strongly dependant on shielding material, and thickness of shielding. As a rule, gamma rays with higher energy are more penetrating and require more shielding material for the same percent of dose rate reduction than gamma rays with lower energy For example, approximate tenth-value layers for 60Co (1170 keV keV) and 137Cs (661 keV) for lead are 3.6 cm and 1.9 cm, and for iron 5.4 cm and 3.9 cm. These are conservative values, which also take account of scattered radiation.

16. Case 11 (Cont.)
Calculate half-value layers for gamma rays from 60Co and 137Cs for lead and iron;
What would be the thickness of lead (iron) shielding required to decrease dose rate from 60Co (137Cs) to 0.1 % of unshielded value?
What would be the thickness of lead (iron) shielding required to decrease dose rate from 60Co (137Cs) to 1/16th of unshielded value?
What would be the thickness of lead (iron) shielding required to decrease dose rate from 60Co (137Cs) for 99%?
If we have lead (iron) shielding with thickness 10 cm (9 cm) how many times (approximately!) will be dose rate from 60Co (137Cs) lower then without shielding?

17. Case 12
Lead blankets are made of lead wool and are used as portable shielding. For one of these products manufacturer declares that a blanket achieves 34% reduction of dose rate for 60Co and 53% of dose rate reduction for 137Cs.
What would be dose rate reduction for 60Co (137Cs), if we use 2 and 3 layers of blanket?

18. Case 13 Exposure situations are divided into three categories:
planned exposure situations,
emergency situations,
and existing situations.
State at least five examples for each category in fuel cycle facilities, research facilities, and medical institutions.

19. Occupationally exposed workers
Case 14
Fill in the following table for planned exposures:
Occupationally exposed workers
Apprentices
Members of the public
Annual effective dose limit
Annual equivalent dose limit for the lens of the eye
Annual equivalent dose limit for extremities
Annual equivalent dose limit for skin

20. Case 15 One of the valves on primary system sampling line is leaking.
What radionuclides you would expect to be measured in the air in the vicinity of the leak? What radionuclides would you expect in the leaking primary water? 
Valve was changed during the outage. Before transport to decontamination facility, removable contamination on the valve was measured. What radionuclides you would expect to be detected?

21. Case 16
Reactor TRIGA Mk II is small pool-type research reactor. Core is at the bottom of the pool under six meters of deionised water and consists of approximately fifty fuel rods. Rods are stainless steel tubes with the mixture of uranium and zirconium hydride inside. Water in the pool is constantly circulated through filter, mixed-bed demineraliser, and heat‑exchanger, which is cooled with water pumped from the well and discharged to a river. Reactor building is normal industrial hall with ventilation exhaust over filters. Discharged water and air are being constantly monitored for radionuclides.
Reactor building is a part of research facility, which is surrounded with crop fields. The nearest village is 400 m away.

22. Case 16 (Cont.)
Identify important exposure pathways to population from TRIGA Mk II reactor;
Outline basic environmental monitoring programme for TRIGA Mk II reactor: list environmentally monitored constituents and necessary frequency of samplings and measurements.
The views expressed in this document do not necessarily reflect the views of the European Commission.