AERB SAFETY GUIDE NO.

AERB/SG/O-5

RADIATION PROTECTION
DURING
OPERATION OF NUCLEAR POWER
PLANTS

Issued in August 1998

Atomic Energy Regulatory Board

Mumbai – 400 094
 Price :

Orders for this guide should be addressed to:

The Administrative Officer,

Atomic Energy Regulatory Board
Niyamak Bhavan,
Anushaktinagar,
Mumbai - 400 094

ii
 FOREWORD

Safety of public, occupational workers and the protection of environment should be

assured while activities for economic and social progress are pursued. These activities
include the establishment and utilisation of nuclear facilities and use of radioactive sources.
They have to be carried out in accordance with relevant provisions in the Atomic Energy
Act 1962 (33 of 1962).

Assuring high safety standards has been of prime importance since the inception of nuclear
power programme in the country. Recognising this aspect, the Government of India
constituted the Atomic Energy Regulatory Board (AERB) in November 1983 vide standing
order No. 4772 notified in Gazette of India dated 31.12.1983. The Board has been
entrusted with the responsibility of laying down safety standards and to frame rules and
regulations in respect of regulatory and safety functions envisaged under the Atomic
Energy Act of 1962. Under its programme of developing safety codes and guides, AERB
has issued four codes of practice covering the following topics:

Safety in Nuclear Power Plant Siting

Safety in Nuclear Power Plant Design
Safety in Nuclear Power Plant Operation
Quality Assurance for Safety in Nuclear Power Plants.

Safety guides are issued to describe and make available methods of implementing specific
parts of the relevant codes of practice as acceptable to AERB. Methods and solutions other
than those set out in the guides may be acceptable if they provide at least comparable
assurance that nuclear power plants can be operated without undue risk to the health and
safety of general public and plant personnel.

The codes and safety guides may be revised as and when necessary, in the light of
experience as well as relevant developments in the field. The appendices included in the
document are considered to be an integral part of the document, whereas, foot-notes and
bibliography are to provide information that might be helpful to the user.

The emphasis in the codes and guides is on the protection of site personnel and public from
undue radiological hazard. However, for aspects not covered in the codes and guides,
applicable and acceptable national and international codes and standards shall be followed.
Industrial safety shall be assured through good engineering practices and through
compliance with the Factories Act 1948 as amended in 1987 and the Atomic Energy
(Factories) Rules, 1996.

The Code of Practice on Safety in Nuclear Power Plant Operation states the minimum
safety requirements to be met during the operation of thermal neutron reactor based power
plants in India. This safety guide provides guidance for radiation protection during the
operation of such power plants.

i
This safety guide has been prepared by the staff of AERB and other professionals. In
drafting this guide, they have used extensively the relevant guides of International Atomic
Energy Agency (IAEA) developed under Nuclear Safety Standards (NUSS) programme,
specially the IAEA Safety Guide on "Radiation Protection during Operation of Nuclear
Power Plants" (No.50-SG-05,1980) and recommendations of the International
Commission on Radiation Protection (ICRP-60) issued in 1990.

This safety guide has been reviewed by experts and vetted by the AERB Advisory
Committees before issue. AERB wishes to thank all individuals and organisations who
reviewed the draft and finalised this safety guide. The list of persons who have participated
in the committee meetings, alongwith their affiliations, is included for information.

(P. Rama Rao)

Chairman, AERB

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 DEFINITIONS

The following definitions apply to the purposes of this guide.

Absorbed Dose*
The fundamental dosimetric quantity D, defined as:

D = d∈/dm

where d∈ is the mean energy imparted by ionising radiation to matter in a volume element
and dm is the mass of matter in the volume element. The energy can be averaged over any
defined volume, the average dose being equal to the total energy imparted in the volume
divided by the mass in the volume. The SI unit of absorbed dose is the joule per kilogram
(J.kg-1), termed the gray (Gy).

Accident
Any unplanned event resulting in (or having the potential to result in) injury or
damages to equipment, which could cause release of unacceptable quantities of radioactive
material.

Accident Conditions
Substantial deviations from Operational States which could lead to release of
unacceptable quantities of radioactive materials. They are more severe than anticipated
operational occurrences and include design basis accidents and severe accidents 1.

Activity*
The quantity A for an amount of radionuclide in a given energy state at a given
time, is defined as:

A=dN/dt

where dN is the expectation value of the number of spontaneous nuclear transformations

from the given energy state in the time interval dt. The SI unit of activity is the reciprocal
second (s-1), termed the becquerel (Bq).

Annual Limit on Intake (ALI)*

The intake by inhalation, by ingestion or by absorption through the skin of a given
radionuclide in a year by Reference Man which would result in a committed dose equal to
the relevant dose limit. The ALI is expressed in units of activity.

1
A substantial deviations may be a major fuel failure, a Loss of Coolant Accident (LOCA) etc.
Examples of engineered safety features are: an Emergency Core Cooling System (ECCS) and containment.

iii
Atomic Energy Regulatory Board (AERB)
National authority designated by Government of India having the legal authority
for issuing the regulatory consents for various activities related to a facility and to perform
safety and regulatory functions including enforcement for the protection of the public and
operating personnel against radiation.

Collective Dose*
An expression of the total radiation dose incurred by a population, defined as the
product of the number of individuals exposed to a source and their average radiation dose.
The collective dose is expressed in man-sieverts (man.Sv).

Collective Effective Dose*

The total effective dose S to a population, defined as:

S = ∑ Ei .N i
i

where Ei is the average effective dose in the population subgroup i and Ni is the number of
individuals in the subgroup.

Committed Effective Dose*

It is the time integral of the whole body effective dose rate following an intake of
a radionuclide. The quantity E(τ), defined as:

E( τ ) = ∑ w .H T (τ )
T T

where HT(τ) is the committed equivalent dose to tissue T over the integration time τ. When
τ is not specified, it will be taken to be 50 years for adults and to age 70 years for intakes
by children.

Committed Equivalent Dose*

It is the tie integral of the equivalent dose rate in an organ or tissue following an
intake of a radionuclide. The quantity HT(τ), defined as:

t 0 +τ
H (τ ) = ∫ H (τ ) dt
T t0 T

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where t0 is the time of intake, HT(τ) is the equivalent dose rate at time t in an organ or tissue
T and τ is the time elapsed after an intake of radioactive substance. When τ is not defined it
will be taken to be 50 years for adults and to age 70 years for intakes by children.
Contamination*
The presence of radioactive substances in or on a material or in the human body
or other place in excess of quantities specified by the Competent Authority.

Controlled Area*
A delineated area to which access is controlled and in which specific protection
measures and safety provisions are, or could be, required for:

(a) controlling normal exposures or preventing the spread of contamination

during normal working conditions; and
(b) preventing potential exposures or limiting their extent should they occur.

Countermeasure*
An action aimed at alleviating the consequences of an accident.

Decontamination*
The removal or reduction of contamination by physical or chemical process.

Derived Air Concentration (DAC)*

It is that activity concentration of the radionuclide in air (Bq/m3) which, if
breathed by Reference Man for a working year of 2000 hours under conditions of light
physical activity (breathing rate 1.2 m3/h), would result in an inhalation of one ALI, or the
concentration which for 2000 hours of air immersion would lead to irradiation of any organ
or tissue to the appropriate annual dose limit.

Derived Intervention Level (DIL)*

Derived Intervention Levels are quantities that can be directly measured, such as
exposure rate from ground deposited activity and activity concentration in foodstuff and
water, at which intervention in the form of countermeasures should be initiated.

Design Basis Accident (DBA)

Design basis accidents are a set of hypothesised accidents which are analysed to
arrive at conservative limits on pressure, temperature and other parameters which are then
used to set specifications that must be met by plant structures, systems and components,
and fission product barriers.

Dose*
A measure of the radiation received or 'absorbed' by a target. The quantities
termed absorbed dose, organ dose, equivalent dose, effective dose, committed equivalent
dose or committed effective dose are used, depending on the context. The modifying terms
are used when they are not necessary for defining the quantity of interest.

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Effective Dose*
The quantity E, defined as a summation of the tissue equivalent doses, each
multiplied by the appropriate tissue weighting factor:

E = ∑ wT .H T
T

where HT is the equivalent dose in tissue T and wT is the tissue weighting factor for tissue
T. The unit of effective dose is J.kg-1, termed the sievert (Sv).

Emergency Situation
A situation which endangers or is likely to endanger safety of the Nuclear Power
Plant (NPP), the site personnel or the environment and the public.

Emergency Plan
A set of procedures to be implemented in the event of an accident.

Equivalent Dose*
It is the absorbed dose in an organ or tissue multiplied by the relevant radiation
weighting factor.

Exposure*
The act or condition of being subject to irradiation. Exposure can be either
external (irradiation by the sources outside the body) or internal (irradiation by sources
inside the body). Exposures can be classified as normal exposure or potential exposure;
either occupational, medical or public exposure; and in intervention situations, either
emergency exposure or chronic exposure. The term exposure is also used in radiodosimetry
to express the amount of ions produced in air by ionising radiation.

Intervention*
Any action intended to reduce exposure or the likelihood of exposure to sources,
which are not part of controlled practice or which are out of control as a consequence of
an accident.

Intervention Level (IL)*

A level of avertable dose at which a specific protective action or remedial action
is taken in an emergency exposure situation or chronic exposure situation.

Medical Exposure*
Exposure incurred by patients as part of their own medical or dental diagnosis or
treatment; by persons, other than occupationally exposed knowingly while voluntarily
helping in the support and comfort of patients; and by volunteers in a programme of
biomedical research involving their exposure.

Normal Operation

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 Operation of a plant or equipment within specified operational limits and
conditions. In the case of nuclear power plants this includes start-up, power operation,
shutting down, shutdown state, maintenance, testing and refuelling.

Nuclear Power Plant (NPP)

A thermal neutron reactor or reactors together with all structures, systems and
components necessary for safety and for the production of power, i.e., electricity.

Occupational Exposure*
All exposures of personnel incurred in the course of their work.

Off-site Emergency
Accident condition/emergency situation involving excessive release of
radioactive materials/hazardous chemicals from the plant into the public domain calling for
intervention.

Operational Limits and Conditions

Limits on plant parameters, and set of rules on the functional capability and the
performance level of equipment and personnel, approved by the Regulatory Body, for safe
operation of the facility.

Plant Management
The members of site personnel who have been officially delegated responsibility
and authority by the Operating Organisation for directing the operations of the plant.

Potential Exposure*
Exposure that is not expected to be delivered with certainty, but that may result
from an accident at a source or owing to an event or sequence of events of a probabilistic
nature, including equipment failures and operating errors.

Practice*
Any human activity that introduces additional sources of exposure pathways or
extends exposure to additional people or modifies the network of exposure pathways from
existing sources, so as to increase the exposure or the likelihood of exposure of people or
the number of people exposed.

Prophylaxis*
Prophylaxis is the intake of specific stable chemical compounds, which have a
reducing or blocking effect on the uptake of certain radionuclides. E.g., the use of stable KI
or KIO3 to reduce the uptake of radioiodines (particularly I-131) in thyroid gland.

Public Exposure*
Exposure incurred by members of the public from radiation sources, excluding
any occupational or medical exposure and the normal local natural background radiation

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but including exposure from authorised sources and practices and from intervention
situations.

Radiation Worker
Any person who is occupationally exposed to radiation and which in the opinion
of the Regulatory Body should be subjected to radiation surveillance.

Reference Level*
Action level, intervention level, investigation level or recording level. Such levels
may be established for any of the quantities determined in the practice of radiation
protection.

Reference Man*
An idealised adult Caucasian human defined by the ICRP for the purpose of
radiation protection assessment.

Regulatory Constraints*
Restrictions on radiation protection parameters specified by the Regulatory Body.

Responsible Organisation (RO)

The organisation having overall responsibility for siting, design, construction,
commissioning, operation and decommissioning a facility.

Site Emergency
Accidental condition/emergency situation in the plant involving radioactivity
transgressing the plant boundary but confined to the site, or involving release of hazardous
chemicals/explosion, whose effects are confined to the site, with off-site consequences
expected to be negligible.

Site Personnel
All persons working on the site, either permanently or temporarily.

Supervised Area*
An area not designated as controlled area but for which occupational exposure
conditions are kept under review even though specific protective measures and safety
provisions are not normally needed.

Technical Specification for Operation

A document submitted on behalf of or by the Responsible Organisation covering
Operational Limits and Conditions, Surveillance and Administrative Control requirements
for the safe operation of the facility and approved by Regulatory Body.

viii
ix
 CONTENTS

FOREWORD ………………………………………………………………… i
DEFINITIONS ……………………………………………………………… iii

1. INTRODUCTION ………………………………………………….. 1
1.1 General ……………………………………………………. 1
1.2 Objectives …………………………………………………. 1
1.3 Scope ……………………………………………………… 1

2. ORGANISATIONAL ASPECTS ………………………………..…. 3

2.1 General ……………………………………………………. 3
2.2 Duties and Responsibilities of the Operating
Organisation and Plant Management …………………….. 3
2.3 Duties and Responsibilities of Health Physics Unit ………. 5
2.4 Duties and Responsibilities of Radiation Workers ……….. 6
2.5 Health Surveillance and Medical Examination …………... 7

3. SYSTEM OF RADIATION PROTECTION ………………………. 8

3.1 Basic Principles …………………………………………… 8
3.2 Radiation Protection Programme …………………………. 8
3.3 Dose Limits ……………………………………………….. 8
3.4 Exposure Control Scheme ………………………………… 9
3.5 Measures for Control of Exposure to Site Personnel …….. 9
3.6 Exposure Control of Temporary Workers ……………….. 11
3.7 Reference Levels ………………………………………….. 11
3.8 Control of Exposure to the Members of Public …………… 12

4. IMPORTANT ASPECTS OF IMPLEMENTATION OF

RADIATION PROTECTION PROGRAMME …………………….. 13
4.1 Design Aspects ……………………………………………. 13
4.2 Operational Aspects ………………………………………. 13
4.3 Dose Determination ………………………………………. 14
4.4 Classification of Areas ……………………………………. 14
4.5 Zone Classification ……………………………………….. 15
4.6 Work in Controlled Areas ………………………………… 16
4.7 Work Planning ……………………………………………. 17
4.8 Radiation Work Permits …………………………………. 18
4.9 Use of Protective Clothing and Equipment ………………. 19
4.10 Decontamination ………………………………………….. 20
4.11 Area Monitoring and Surveys …………………………….. 20

5. DOSE ASSESSMENT ……………………………………………… 23

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 5.1 External Dose Monitoring ……………………………….. 23
5.2 Internal Dose Monitoring …………………………………. 24

6. MANAGEMENT OF RADIOACTIVE WASTES ………………… 25

6.1 General ……………………………………………………. 25
6.2 Monitoring of Solid Wastes ………………………………. 25
6.3 Monitoring of Liquid Wastes ……………………………… 25
6.4 Monitoring of Gaseous Wastes ……………………………. 26
6.5 Control of Releases ……………………………………….. 26

7. STORAGE, HANDLING AND TRANSPORTATION OF

RADIOACTIVE MATERIALS …………………………………. 28
7.1 General ……………………………………………………. 28
7.2 Radiological Control Measures …………………………… 28

8. ENVIRONMENTAL SURVEILLANCE ………………………….. 29

9. FUNCTIONAL INFRASTRUCTURE AND EQUIPMENT ………. 30

9.1 Health Physics Facilities …..……………………………… 30
9.2 Equipment …………………………………………………. 31

10. TRAINING AND QUALIFICATION ……………………………… 32

11. EMERGENCY PREPAREDNESS …………………………………. 33

11.1 Emergency Plan …………………………………………… 33
11.2 Facilities and Equipments provided at Site for
Handling Emergency ……………………………………… 33
11.3 Training and Exercises …………………………………… 34
11.4 Intervention Levels ……………………………………….. 34
11.5 Limitation of Exposure in Emergency …………………… 34
11.6 Assessment and Monitoring ……………………………… 35

12. DOCUMENTATION ……………………………………………….. 36

12.1 General ……………………………………………………. 36
12.2 Types of Records ………………………………………….. 36
12.3 Criteria for Preparing and Maintaining Records …………. 36

APPENDIX-I Explanatory notes on Radiation Protection

Concepts and Terms used in this Safety Guide … 38

APPENDIX-II Dose limits and Constraints …………………….. 44

APPENDIX-III Dose Limit for Occupational Exposures

As approved by AERB in March 1994 ………….. 46

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APPENDIX-IV Annual Limit of Intake (ALI inhalation)
values for radionuclides of common interest …… 47

REFERENCES …………………………………………………….. 48

LIST OF PARTICIPANTS OF WORKING GROUP .…………….. 49

ADVISORY COMMITTEE ON CODES, GUIDES AND

ASSOCIATED MANUALS FOR SAFETY IN OPERATION
OF NUCLEAR POWER PLANTS (ACCGASO) ………………….. 50

ADVISORY COMMITTEE ON NUCLEAR SCIENCES

(ACNS) ……………………………………………….…………….. 51

PROVISIONAL LIST OF SAFETY GUIDES ON

OPERATION OF NUCLEAR POWER PLANTS ……………..….. 52

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 1. INTRODUCTION
1.1 General

1.1.1 This Safety Guide is prepared as a part of Atomic Energy Regulatory Board's
(AERB) programme for establishing Codes of Practice and Safety Guides relating
to Nuclear Power Plant (NPP) Operation. The purpose of this Guide is to provide
practical guidance for establishing and maintaining a Radiation Protection
Programme.

1.1.2 This Guide deals with the protection of site personnel and the general public from
exposure to ionising radiation while still allowing necessary activities from which
radiation exposure might result. It outlines how radiation protection at the NPP
site can be implemented, in compliance with the guidelines established by the
AERB, for all operational states of the plant.

1.1.3 In the preparation of this guide, guidelines from IAEA Safety Guide on Radiation
Protection During Operation of Nuclear Power Plants [1], International Basic
Safety Standards for Protection Against Ionising Radiation and for the Safety of
Radiation Sources [2], Recommendations of International Commission on
Radiological Protection [3] and AERB Safety Manual on Radiation Protection for
Nuclear Facilities [4] have been taken into consideration.

1.2 Objectives

1.2.1 The objectives of this Safety Guide are as follows:

(1) To provide guidelines to the Plant Management for establishing

Radiation Protection Programme to be carried out efficiently and
effectively to achieve protection of the occupational workers, members
of the public and the environment from adverse effects of radiation
while at the same time allowing the justified activities.

(2) To focus on the need for a high degree of commitment on the part of all
levels of management of Operating Organisation and the plant personnel
to follow the exposure control measures during all operational states and
accident conditions at the plant site.

1.3 Scope

1.3.1 This Guide deals with the Radiation Protection Programme for NPPs. It outlines
the basic principles of radiation protection, organisational aspects, the
responsibilities of different groups of personnel and the practical aspects of
establishing and implementing a Radiation Protection Programme during the
operation of NPPs.

1
1.3.2 This Guide covers the requirements for a Radiation Protection Programme for all
operational states of the NPP. It also provides limited guidelines for accident
conditions.

1.3.3 This Guide outlines the principles of dose limitation to plant personnel and to the
public, but it does not give the methodology for the assessment of exposures to
members of public. It also covers the principal steps in environmental monitoring
programme, but does not include details on environmental surveys.

1.3.4 This Guide does not cover detailed information on handling of radiation sources
nor does it include guidelines on the transport of irradiated fuel and other
radioactive materials. Separate Guides on these topics are available and they are
referred to at appropriate places.

2
 2. ORGANISATIONAL ASPECTS
2.1 General

2.1.1 The Operating Organisation # shall be responsible for the establishment and
implementation of the Radiation Protection Programme.

2.1.2 A health physics unit shall be established at the NPP at an appropriate time in
order to ensure that the Radiation Protection Programme is implemented by the
operating organisation.

2.1.3 The position of the Station Health Physicist, who is in-charge of the health
physics unit, will be such that he can readily advise plant management on all
aspects related to Radiation Protection Programme. He shall have access to the
levels of the management which have authority to establish and to enforce
radiation protection procedures.

2.1.4 The plant management shall provide appropriate training and retraining to all
levels of personnel so as to have a continuing degree of competence.

2.1.5 All site personnel have an individual responsibility for practising exposure control
measures specified in the Radiation Protection Programme.

2.1.6 The operating organisation shall make adequate reviews and audits of the
implementation and the effectiveness of the Radiation Protection Programme.

2.2 Duties and Responsibilities of the Operating Organisation and Plant

Management

2.2.1 A strong commitment to ensure radiation safety should come from all levels of
management, including the Responsible Organisation, Design Group, Plant
Management and Operations and Maintenance Groups. Systematic efforts should
be made to induce safety awareness and propagate safety culture amongst staff on
continual basis.

2.2.2 Starting at the design stage, the responsible organisation shall review the design
provisions to determine if they are adequate for the successful implementation of
the Radiation Protection Programme. The aspects to be reviewed should include:

(1) Radiation shielding;

(2) Ventilation and air cleaning;
(3) Radioactivities in active systems;
#
Where Operating Organisation is not instituted, this will be the responsibility of Responsible
Organisation (R.O) or Plant Management as delegated by R.O.

3
 (4) Containment integrity;
(5) Instrumentation for radiation, contamination and personnel dose
monitoring and control under operational states and accident conditions;
(6) Equipment layout and access;
(7) Radiation and contamination zoning arrangements;
(8) Access controls, with particular attention to the number of persons
employed in controlled areas;
(9) Adequacy of access, and of working space for maintenance;
(10) Elements of design that allow quick and remote handling for dismantling
and reassembling of components, shielding, etc.;
(11) Radioactive waste storage, handling and equipment;
(12) Availability of protective equipment, protective wear and respirators, and
decontamination and monitoring rooms;
(13) Effluent Control and monitoring;
(14) Facilities for sampling during accident conditions;
(15) Availability of up-to-date set of drawings, supplemented by photographs,
and those operating and maintenance manuals and instructions that are
required for planning procedures to minimise exposure; and
(16) Radiological data acquisition, processing and retrieval.

Details of the design aspects of radiological protection during design are dealt
with in the AERB Safety Guide “Design aspects of Radiological Protection for
PHWR based NPPs" (AERB/SG/D-12) [5].

2.2.3 The plant management is responsible for all aspects of station operation. They
shall ensure compliance with radiation protection procedures by all the station
personnel in order to maintain occupational exposures within the specified limits
and As Low As Reasonably Achievable (ALARA).

2.2.4 The plant management shall prepare, in consultation with the Station Health
Physicist, Operating Manual for Radiation Protection Procedures and Radiation
Emergency Procedures for the plant giving details of the methods of controlling
the radiation exposure to site personnel and members of the public before
commencement of operation.

2.2.5 Adequate training in plant radiation protection and emergency procedures shall be
imparted to personnel at all levels.

2.2.6 The plant management shall provide adequate equipment and facilities for
personnel protection, for monitoring of external and internal exposure, for
monitoring the radiological conditions in the plant, and for monitoring the effluent
and the environment.
2.2.7 The plant management shall arrange for routine and non-routine medical
examinations of all radiation workers in the plant in accordance with the
requirements of the AERB.

4
2.2.8 The plant management shall establish procedures for periodic review of the
implementation of the Radiation Protection Programme, and shall ensure that its
objectives are achieved. Special attention shall be given to radiation incidents,
exposures beyond authorised limits and effluent discharge control. Periodic audits
shall also be arranged to ensure effective compliance with established procedures.

2.2.9 The plant management shall arrange for proper collection, storage, handling and
disposal of the wastes that are generated in the plant as per approved procedures.

2.2.10 The plant management shall supply all necessary information to assist the health
physicist in carrying out investigations in any radiation related unusual occurrence
or incident. The plant management should take effective measures to implement
the suggestions and recommendations given by the health physicist for
improvement of safety and rectify the deficiencies on priority.

2.2.11 All the data pertaining to radiological safety shall be maintained and periodic
reports shall be sent to the AERB as per the requirement [4].

2.2.12 The plant management shall maintain complete and up-to-date records of medical
and occupational histories of every radiation worker. The plant management shall
preserve the occupational health records along with the dose records for the
working life of the individual, and thereafter as specified by AERB [4].

2.2.13 The plant management shall arrange for appropriate monitoring of both liquid and
gaseous effluents to be discharged from the plant to the environment and also for
maintenance of relevant records of the same.

2.2.14 Compliance with applicable requirements of the AERB and with the radiation
protection aspects of the policies and procedures of the operating organisation
shall be demonstrated appropriately.

2.3 Duties and Responsibilities of Health Physics Unit

The duties and responsibilities assigned to health physics unit include the
following:

(1) To advise plant personnel/management regarding measures to be taken for

effective exposure control on the basis of assessment of the radiological
status in the plant areas;
(2) To arrange for and implement personnel monitoring for external and
internal exposures, site radiation surveys and other related services;
(3) To specify procedures and protective equipment for radiation work and to
issue radiological work permits prescribing the same;
(4) To prepare radiation protection procedures and data forms;

5
 (5) To specify radiation protection procedures for handling and storage of
radioactive material on site in accordance with the guidelines stipulated by
the regulatory body;
(6) To maintain adequate stocks of instruments for personnel dosimetry,
radiation survey, counting and analyses ;
(7) To classify and delineate radiation and contamination zones and to survey
and monitor them periodically;
(8) To monitor compliance by site personnel with plant radiation protection
procedures;
(9) To check that the active or contaminated waste or components sent off site
conform to the appropriate transport regulations with respect to the
radiological aspects;
(10) To notify plant management of any individual dose in excess of specified
limits, which shall be investigated by appropriately constituted committees;
(11) To participate in investigations of radiation exposures exceeding
prescribed levels for personnel and of abnormal radiological conditions and
emergency conditions;
(12) To prepare and maintain all records relevant to the Radiation Protection
Programme;
(13) To collaborate in the preparation of and implementation of emergency
plans and procedures;
(14) To specify, and equip the NPP with, special instrumentation and equipment
for radiation protection that is adequate to cope with emergency situations;
(15) To select and train adequate health physics personnel; and
(16) To train all site personnel in radiation protection measures as appropriate
to their duties as prescribed in station radiation protection procedures and
to prepare radiological training materials.

2.4 Duties and Responsibilities of Radiation Workers

2.4.1 The radiation worker shall follow the approved radiation protection and radiation
emergency procedures and shall refrain from any wilful act that could be
detrimental to self or to his co-workers or to the plant or to the environment.

He shall further:

(1) provide information about his past radiation work if any;

(2) make proper use of the protective equipment, radiation monitors and
dosimetric devices provided;
(3) inform forthwith the supervisor/health physicist of any unusual incident or
potentially unsafe situation that may come to his notice;
(4) comply with the requirements of health surveillance and dose assessment
programme;
(5) inform the health physicist regarding loss or damage to his personal dose-
monitoring device.

6
2.4.2 A female worker shall, on becoming aware that she is pregnant, notify the plant
management and health physicist in order that appropriate dose limits be applied.

2.5 Health Surveillance and Medical Examination

2.5.1 A health surveillance programme for radiation workers shall be established on the
general principles of occupational health. Health surveillance has the following
objectives:

(1) To assess the health status of the individual;

(2) To help in ensuring initial and continuing compatibility between the
health of the individual and the conditions of their work; and
(3) To provide base-line information useful in the case of occupational or
accidental radiation exposure.

2.5.2 The surveillance shall be done for all radiation workers. This shall consist of pre-
employment medical examination, which shall include family and personal
history, previous occupational history, previous radiation therapy and clinical
investigations.

2.5.3 All radiation workers shall be medically examined at specified intervals.

Appropriate information on individuals, who would require a change or restriction
in the job allocation on the basis of their health status, will be included in the
medical report submitted to the plant management.

2.5.4 When the life time dose exceeds a level as prescribed by AERB, further exposure
of such individuals shall be allowed only after reviewing the health status by
medical authorities as per approved procedures.

2.5.5 In case of an individual receiving or suspected to have received exposures

exceeding the specified values, medical investigations like chromosome
aberration analysis should be carried out and appropriate medical treatment, if
required, should be given.

2.5.6 Plant management should utilise the services of a physician who has had special
training in industrial and radiation medicine and who is competent to advise on
and supervise the medical examinations and treatment of any person involved in a
radiation accident.

7
 3. SYSTEM OF RADIATION PROTECTION

3.1 Basic Principles

For the radiation protection of site personnel and members of the public, a system
of radiation protection shall be established based on the following principles and
objectives.

(1) The effective dose or equivalent dose to individuals shall not exceed the
applicable dose limits stipulated by AERB.

(2) All exposures, individual, collective and potential shall be kept as low as
reasonably achievable (ALARA), economic and social factors being taken
into account.

3.2 Radiation Protection Programme

3.2.1. The above objectives should be accomplished by means of a Radiation Protection

Programme of precautionary measures, radiation and dose monitoring, safety
culture, management and surveillance.

3.2.2 The common objective of radiation protection measures in the operation of NPPs
is to protect the personnel and the environment from the deleterious effects of
radiation while at the same time allowing the operation and maintenance activities
as envisaged by the design.

3.2.3 The Radiation Protection Programme is structured to ensure that the exposures of
all site personnel, and the public are maintained consistent with the stipulations
issued by the AERB in this regard.

3.3 Dose Limits

3.3.1 The dose limits are given in Appendix-II. These dose limits do not include
radiation doses due to natural background radiation and those resulting from
medical exposure.

3.3.2 The safety directives and other requirements issued by the AERB shall be
translated into provisions or procedures that individuals can apply and shall be
incorporated in the radiation protection procedures of the plant.

3.3.3 The dose limits apply to the sum of the relevant doses from external exposure in
the specified period and the committed doses from intakes of radioactive
substances in the same period.

8
3.3.4 There is no separate occupational dose limit for women in general. However, once
pregnancy is declared or diagnosed the equivalent dose limit to the surface of the
abdomen shall be as given in Appendix-II (1.1.4).

3.3.5 No person below the age of 18 years shall be employed as radiation worker except
with prior permission of AERB in writing.

3.3.6 In the event of any site personnel receiving exposures exceeding the prescribed
levels it shall be investigated promptly and reported to AERB and remedial
actions taken to prevent the recurrence of the same.

3.3.7 The plant management should establish a dose reduction and ALARA programme
at the station and make all efforts to adhere to the guidelines on the collective
dose for a particular NPP prescribed by AERB.

3.3.8 In case collective dose exceeds the values specified by AERB, the same should be
reviewed and reported to AERB along with measures being taken to control the
same.

3.4 Exposure Control Scheme

3.4.1 The exposure control scheme should consist of a three-tier system, namely,
application of primary dose limits, regulatory constraints and operational
restrictions.

3.4.2 Operational restrictions shall be established on quantities such as dose rate,

effluent release rate and other operational parameters so that the exposure of site
personnel and members of the public does not exceed any relevant dose limit or
authorised limits.

3.4.3 The regulatory constraints stipulated by the AERB shall be followed for
controlling the exposures so as to keep the total exposure of personnel below the
specified dose limits. Individual exposures exceeding the regulatory constraints
shall be investigated and reported to the AERB.

3.5 Measures for Control of Exposures to Site Personnel

3.5.1 Exposures to site personnel should be limited by an appropriate combination of

radiation protection measures such as:

(1) Restriction of external exposure by means of shielding, remote operation,

source control and minimising the exposure time;
(2) Restriction of internal exposure by means of isolation, ventilation,
cleanliness and the use of protective clothing and respiratory equipments;
(3) Proper layout of plant areas including zone classification of areas;

9
 (4) Training of personnel;
(5) Review of work procedures, planning and dose budgeting;
(6) By appropriate mock up arrangements for rehearsing work plans for
special jobs;
(7) Monitoring of individuals and work areas.

3.5.2 The plant management shall in consultation with health physics unit prepare a
Collective Dose Estimate for the year including that for special jobs as per
approved procedures. The actual collective dose expenditure should be reviewed
by the plant management towards achieving exposures ALARA.

3.5.3 Use of techniques such as computerised dose tracking will aid in better
management of the overall radiological control process.

3.5.4 Proper work culture should be promoted among all levels of personnel for
reducing individual and collective doses. Workforce required for a specific job
should be carefully deployed keeping in view the special skills required and also
the budgeted collective dose for the job.

3.5.5 Adequate auxiliary lighting and a comfortable working environment can increase
the efficiency of the worker and thus reduce the time spent in the higher radiation
zones and thereby minimise the dose received.

3.5.6 Inspections and other works in high radiation areas should be carefully scheduled
taking advantage of reduction of radiation sources due to decay and other
processes during the reactor shut down period and absorption or exhaust of air
contaminants so that exposures can be minimised. Survey data and previous
experience can be made use of in planning and scheduling of specific tasks.

3.5.7 Adequate supervision and radiation protection surveillance should be provided,

while personnel work in radiation areas, to ensure that the appropriate procedures
are followed, that proper protection is taken and that potential problems that
might develop during the operation are addressed to in a timely manner.

3.5.8 There should be adequate communication systems such as telephones/ headphone

sets, walkie-talkie, CCTV between personnel working in high radiation zones and
those who monitor the work from other locations. This can permit exchange of
information and avoid unnecessary exposures to personnel.

3.5.9 Ventilation balancing, integrity of the ventilation system and proper airflow
pattern should be ensured to minimise the external exposure contribution due to
radioactive inert gases.

3.5.10 Ventilation of work areas and system integrity should be such that airborne
radioactive concentration in the work areas are maintained as low as possible and

10
 below the prescribed limits. Airborne radioactivity concentrations in excess of the
prescribed limits will require work restrictions, use of respiratory protection and
other protective equipment. This may also be required to be followed up by bio-
assay/whole body monitoring.

3.5.11 Auxiliary ventilation system as appropriate should be employed to augment the

permanent system in case there is substantial scope for potential airborne
contamination.

3.5.12 Uptake of radioactive material should be minimised by ensuring that adequate

protective equipment is properly worn, removed, stored, laundered and surveyed.
These physical controls in conjunction with training and administrative
requirements such as, prohibition of eating, drinking and smoking in controlled
areas and decontamination techniques ensure that potential ingestion of
radioactive material is minimised.

3.5.13 In order to limit internal exposures, procedures should be evolved regarding

restrictions on access of individual to radiation areas when his intake crosses a
certain level.

3.6 Exposure Control of Temporary Workers

Besides ensuring compliance with the applicable exposure control measures as in

section 3.5.1 above, the temporary workers employed for working in the
controlled areas should meet the following requirements.

(1) They should undergo pre-employment medical check up and training in

elementary radiation protection procedures applicable at the plant.
(2) They should be closely supervised by an appropriately qualified person
during their work to ensure that the safety procedures including use of
personnel monitoring devices and protective equipments are being
followed and they are guided during normal and emergency situations in
the plant.
(3) The plant management should obtain the previous dose data, if any. The
dose received by temporary workers shall be within the limits prescribed
by the AERB. The dose limit for this category is normally lower than
that for the radiation workers.
(4) Persons below the age of 18 years shall not be employed for radiation
work.

3.7 Reference Levels

In order to facilitate the administrative control of exposures to personnel, plant

management should establish values of measurable quantities called reference
levels that are lower than the limits prescribed. Whenever these levels are

11
 exceeded, a review or investigation should be carried out by plant management.
Reference levels may be applicable for dose for periods of less than one year,
dose rates or contamination levels in an area.

3.8 Control of Exposure to the Members of the Public

3.8.1 The doses to the members of the public due to operation of NPPs shall not exceed
the limits prescribed by the AERB. The control of public exposure in all normal
situations shall be exercised by application of controls at source rather than in the
environment. The dose limits for members of the public do not include
occupational exposures, exposures from the natural environment and medical
exposures.

3.8.2 The authorised limits for discharge of effluents to the environment should be
established based on the dose apportionment assigned to different facilities at the
site for different routes of releases and different radionuclides as approved by
AERB.
3.8.3 The amount and methods by which radioactive materials are released to the
environment shall be controlled in accordance with the Safety Directives issued
by the AERB.

3.8.4 Effluent monitoring shall be performed to demonstrate compliance with the

prescribed discharge limits. Environmental surveillance shall be carried out in
order to assess the radiological impact on the environment and to estimate the
public exposures to confirm compliance with the regulatory requirements.

12
4. IMPORTANT ASPECTS FOR IMPLEMENTATION OF
RADIATION PROTECTION PROGRAMME

4.1 Design Aspects

4.1.1 The design philosophy established for NPPs should strive to maintain
occupational radiation exposures ALARA and should be in compliance with
applicable regulations.

4.1.2 In order to achieve the above objectives, working procedures and methods should
be examined with regard to the possibility of reducing doses resulting from these
activities. The main methods of dose reduction are:

(1) Reducing the radiation levels in work areas achieved by proper design,
plant layout, and provision of shielding;
(2) Reducing surface and airborne contamination by design;
(3) Reducing the personnel exposure time by component selection, choice of
work methods and proper training; and
(4) Source control by proper selection of materials/components.

Details on the design aspects of radiological protection during design are dealt
with in the AERB Safety Guide on "Design Aspects of Radiological Protection"
(AERB/SG/D-12) [5].

4.1.3 The spread of airborne contamination within the station should be limited by
maintaining air pressure gradients and air flows from areas of low potential
airborne contamination to areas of higher potential contamination. Periodic
checks would ensure that the design pressure differentials are being maintained.

4.2 Operational Aspects

During the operating stage of NPPs, the following methods should be used for
achieving the objectives of radiation protection programme:

(1) Radiation and contamination control procedures;

(2) Adherence to approved operating and maintenance procedures;
(3) Implementing training programme;
(4) Appropriate use of automatic or remote controlled equipment;
(5) The use of temporary shielding;
(6) Provision of local ventilation with adequate filteration and
(7) Source control by ensuring leak tightness of components and systems.

13
4.3 Dose Determination

A complete dosimetry programme should be developed, documented and

implemented. Dose monitoring shall comply with the requirements established by
the AERB. It should include such aspects as:

(1) Personnel monitoring for external and internal exposures;

(2) Criteria for determining type and periodicity of monitoring;
(3) Procedures to be followed by radiation workers to comply with
dosimetry or monitoring requirements;
(4) Methods and procedures for ensuring quality assurance in the dosimetry
programme;
(5) Methodology for assigning dose;
(6) Establishment of recording levels, investigation levels and other levels
which may be appropriate and
(7) Recording and retention of dosimetry and other related data.

4.4 Classification of Areas

4.4.1 The plant areas should be classified into two types of areas, namely controlled
areas and supervised areas.

4.4.2 Controlled Area: Any area in which specific protection measures or safety
provisions are or could be required for controlling normal exposures or preventing
the spread of contamination during normal working conditions and preventing or
limiting the extent of potential exposures shall be designated as controlled area.

4.4.3 Requirements of Controlled Areas:

(1) delineate controlled areas by suitable means such as doors, distinctive

painting and signboards, etc.;
(2) restrict access to controlled areas by means of administrative procedures
such as the use of work permits, and by physical barriers which could
include locks or interlocks; the degree of restriction being commensurate
with the magnitude and likelihood of the expected exposure;
(3) provide individual personnel monitoring;
(4) provide protective clothing and equipment, monitoring equipment and
suitable storage for personal clothing, as appropriate, at entrances to
controlled areas;
(5) provide contamination monitors, washing or shower facilities and
suitable storage for contaminated protective clothing and equipment, as
appropriate, at the exits from controlled areas.

14
4.4.4 Supervised Area: Any area not already designated as a controlled area but where
occupational exposure conditions need to be kept under review even though
specific protection measures and safety provisions are not normally needed, shall
be designated as supervised areas.

4.4.5 Requirements of Supervised Areas:

(1) delineate the supervised areas by appropriate means;

(2) display approved signs at appropriate access points to supervised areas.

4.4.6 The radiation status of such supervised areas should be confirmed by regular area
monitoring. Routine individual personal monitoring is not needed while
confirmatory monitoring may be carried out once in a while.

4.5 Zone Classification

4.5.1 General

4.5.1.1 In order to restrict contamination in various areas to minimum and also to

facilitate control of spread of contamination potential, the entire plant area is
divided into various distinct zones.

4.5.1.2 Each zone is clearly demarcated and provided with interzonal barriers with
monitors such as hand and foot monitors/body monitor (frisker) for checking
personnel and equipment while passing from higher contaminated zone to lower
contaminated zone.

4.5.2 Typical Zoning Classification

4.5.2.1 Typically Supervised areas and Controlled Areas are assigned the following
zoning classification based on contamination potential.

(1) Supervised Area: Zone-1

(2) Controlled Areas: Zone-2

Zone-3
Zone-4

4.5.2.2 The various zones have the following features:

(1) Zone-1

(a) This zone will contain no radioactive equipment and will be

kept free of contamination at all times.

15
 (b) Typically it may comprise Administrative Buildings, Control
Room, DG Room, Turbine Building, Switch Yard etc.

(2) Zone-2

(a) This zone contains no radioactive equipment and should not

normally become contaminated. However some contamination
may, though inadvertently get into this area with movement of
personnel and equipment from zone-3. Contamination in zone-2
will be cleared as soon as it is discovered.
(b) Typically this zone will include inactive shops (mechanical,
electrical, control), laboratories, change room, wash room etc.
(c) For zone 2, Radiological Work Permit (RWP) is not generally
called for even though this is controlled area since no radiation
related activities are expected to be carried out normally. This
zone is a sort of buffer zone, which may get contaminated
occasionally.

(3) Zone-3

(a) It generally has only “closed” radioactive sources/ systems.

These zones will normally be contamination free unless the
equipment/systems are opened. This zone includes the service
areas for entire equipment and materials that are potential
sources of contamination. Hence this zone is likely to get
contaminated at times. Equipment lay out and work procedures
will be planned to keep contamination localised and loose
contamination will be cleared up whenever it occurs.
(b) Shops and laboratory areas handling contaminated equipment
frequently come under this zone. Reactor Auxiliary Building,
some parts of Decontamination Centre, laundry, and waste
management facility are also covered by this zone.
(4) Zone-4

(a) This zone normally contains sources of high radiation and hence
can be expected to remain contaminated.
(b) Contamination in this zone is kept localised and under control
by routine clean up operations to maximum extent feasible; but
some parts will remain contaminated.
(c) Whole of Reactor Building, some parts of Decontamination
Centre, Waste Management Plant, Spent Fuel Storage Bay
come under zone-4.

16
4.5.2.3 Use of TLD in zone-2 and beyond is mandatory. Additional dosimeters as
appropriate should be used in higher zones.

4.6 Work in Controlled Areas

4.6.1 The controlled areas and the different zones inside these areas shall be delineated
by suitable means. Contamination monitors should be located on authorised routes
between zones. Warning signs and entry requirements shall be posted at the
entrance of each zone. Procedures should be laid down for the movement of
personnel and equipments through different zones. Exits from controlled areas
shall be clearly marked.

4.6.2 Plant personnel who are required to enter the controlled areas shall be duly
authorised. Fulfilment of radiation protection requirement shall be a pre-requisite
for entry. The entry authorisation for a particular area of the plant may be for a
specified period, or it may be withdrawn if the situation warrants. The access to
controlled area should preferably be through a single checkpoint in order to limit
the spread of contamination.

4.6.3 Requirements for access to a controlled area may include:

(1) Possession of an authorisation issued by plant management and radiation

work permit;
(2) Availability of support personnel, equipment and personnel monitoring
devices;
(3) Protective clothing and equipments and
(4) The required qualification in radiation protection.

4.6.4 Rubber area procedures should be employed to control the spread of

contamination, within a particular zone, in locations where floor contamination
already exists or where it is anticipated due to the nature of work to be done. The
rubber area protective equipment should consist of rubber over-shoes, booties
(shoe covers) or gumboots, gloves and cover-alls or lab coats. Personnel and
equipment shall be checked for contamination and follow up action ensured at the
interzonal cross-point and final exit.

4.6.5 All items shall be monitored for radiation and loose contamination before they are
removed from the controlled area. Appropriate radiation protection measures such
as decontamination, adequate shielding and labelling etc shall be employed before
permitting these items outside controlled and/or supervised areas.

4.7 Work Planning

4.7.1 Major radiological work, particularly during plant shut down, should be planned
well in advance and the plans will include steps to minimise radiation exposures.

17
 The different sections of personnel including work units and health physics
personnel involved in a job should participate in the advance planning.

4.7.2 Detailed written plans should be prepared early enough to provide time to identify
all potential radiation hazards. Job planning should include use of special tools
and techniques, collective dose targets, comparison with similar jobs and
simulated mock up operations as appropriate to increase job efficiency and keep
radiation exposure ALARA.

4.7.3 There should be clear identification of responsibilities and co-ordination among

different agencies participating in the job. It should be ensured that personnel,
tools and equipment are available at the work place before starting the work, and
the personnel are skilled and trained in carrying out the job. Preparation of the
area such as cordoning off and posting warning signs, laying down temporary
shielding, local ventilation exhaust and contamination monitors, may be required
during certain jobs.

4.7.4 Radiological work associated with very high dose rates must be subject to
rigorous planning in consultation with the health physicist and approved by
appropriate levels of plant management.

4.7.5 Written reports, including the radiation protection aspects should be prepared and
reviewed subsequent to any major radiological work.

4.8 Radiation Work Permits (RWP)

4.8.1 The practical implementation of work planning includes issuance of a radiation

work permit to authorise any work, which requires specific radiological
precautions.

4.8.2 For each job that requires radiological precautions to be taken a Radiation Work
Permit (RWP) should be prepared, issued and terminated in an approved manner.
Such job situations generally include:

(1) Entry into any area having high radiation levels;

(2) Any maintenance work, which involves opening of any system that,
contains or could potentially contain any radioactive fluid;
(3) Any job that involves removal or closure of shielding from any
equipment containing radioactive solid material;
(4) Any maintenance of contaminated or potentially contaminated equipment
using methods involving abrasion, cutting, machining or welding, and
(5) Handling of radioactive materials such as spent resin filters/ strainers,
spent fuel, etc..

18
4.8.3 The RWP should contain brief description of the job, location and starting time.
The issuer should be of the level of a shift health physicist duly qualified in
radiation protection and authorised by the Station Health Physicist. The radiation
protection personnel should evaluate the radiological conditions associated with
the work to be performed based on radiation surveys and specify appropriate
protective clothing/devices, dosimeters, special samples, surveys, procedures and
precautions to be taken, time limit and the duration of validity of the permit. On
completion of the job, after noting down the work details and recording the doses
received, the supervisor gets the RWP terminated. If the work is of longer
duration the RWP should be revalidated in each shift and fresh RWP should be
obtained every day.

4.8.4 The person in whose name the RWP is issued, remains responsible for the
radiation protection of the individuals covered by the permit until the surrender of
the permit and must advise the personnel of time limitation, correct use of
protective equipments and dosimeters, bioassay sample submission and of
changes in the radiological conditions.

4.8.5 The RWP should be retained as a record. The RWP procedure provides a
mechanism for collection and evaluation of data pertaining to radiation exposures.

4.9 Use of Protective Clothing and Equipment

4.9.1 Appropriate protective clothing and equipment shall be used in areas where the
airborne and/or surface contamination will involve exposure to internal
contamination.

4.9.2 Protective clothing is used as a barrier to prevent contamination from reaching the
body surface. The higher the level of contamination, the greater the degree of
protection that will be required. This can be achieved by wearing protective
clothing like plastic suits over the coveralls, gloves and shoe covers. When the
individual leaves the contamination zone after work, the outer set of protective
clothing should be removed and placed in bins or bags provided for this purpose.

4.9.3 Ventilated plastic suit shall be used in case it requires complete and effective
isolation of the individual from the environment due to the nature of airborne
contamination prevailing in the work area. This is appropriate for protection
against high airborne tritium contamination and work involving high levels of
loose surface contamination. In special circumstances, plastic suit together with a
self-contained breathing apparatus could also be used as necessary.

4.9.4 Respiratory protective equipment shall be used in areas where airborne

contamination exceeds or is likely to exceed the prescribed limits. The respirators
used should be of approved types only. They should be suitable for the specific
radionuclide of concern. The equipment shall be specifically selected to provide

19
 the degree of protection (protection factor) that is needed for the radiological
conditions.

4.9.5 Reusable protective clothing, shoe covers, gloves etc. used in contaminated areas
should be collected separately and surveyed before sending them to laundry.
They should be surveyed again before reissue to ensure that the contamination
levels are within the prescribed limits. Respirators should be checked for
contamination prior to cleaning and disinfection. The protective equipment
should be routinely inspected, tested and checked for contamination before issue.

4.9.6 Personnel should be trained in the appropriate use of protective clothing and
protective equipment. Special training should be given in the handling of Self-
Contained Breathing Apparatus (SCBA).

4.9.7 The medical compatibility of the individual for the use of specific protective
equipment should also be considered.

4.10 Decontamination

4.10.1 The objective of decontamination is to remove the radioactive materials from

surfaces where it is not desired and to reduce radiation exposure. Adequate
decontamination facility and equipments shall be provided in the controlled area
for personnel as well as equipment decontamination.

4.10.2 All the contaminated equipment should be decontaminated to appropriate levels

before commencement of any maintenance work.

4.10.3 The floor area should be decontaminated periodically on completion of the job so
as to minimise the spread of contamination.

4.10.4 While decontaminating, following a liquid spillage, attention to airborne

contamination should be given and appropriate protective equipment should be
used.

4.10.5 On detection of contamination on any part of the body of the radiation worker, the
decontamination procedures as laid down by the health physicist shall be
followed. If contamination still persists above the prescribed levels medical
advice should be sought through the health physicist.

4.11 Area Monitoring and Surveys

4.11.1 General

20
4.11.1.1 The objectives of area monitoring and surveying are to provide information
on radiological conditions throughout the plant and to ensure that the zone
designation remains valid.

4.11.1.2 Surveys related to specific operations and maintenance jobs should be

performed before the start of a job as well as through the operation of the job.
Non-routine surveys should be conducted as and when necessary.

4.11.1.3 All areas designated as controlled areas and supervised areas shall be
monitored regularly. Workplace monitoring should be done for (a) external
monitoring (b) monitoring for air contamination and (c) monitoring for
surface contamination.

4.11.1.4 The nature and frequency of monitoring will be determined by the likelihood
of changes in the radiological conditions and the potential for exposure.

4.11.1.5 There should be a periodic review of the data collected and remedial actions
taken. A review on implementation plan for improvement should be
conducted.

4.11.2 External Radiation Monitoring

4.11.2.1 External radiation monitoring of the plant areas should be carried out with a
frequency determined by consideration of expected changes in radiation
environment. Areas subject to variation in radiation levels or increased time
of occupancy should be surveyed on a more frequent basis. Radiation
monitoring of the areas should be done for beta-gamma and neutrons as
appropriate. The radiation field on contact of equipments and at a reference
distance also should be carried out. The results of monitoring should be
documented and displayed as necessary.

4.11.2.2 Work place monitoring for external radiation can be carried out automatically
by fixed area gamma and neutron monitors. However this should be
supplemented by periodic manual monitoring by Health Physics personnel.
The area monitors should have recording facilities and shall be calibrated and
checked at appropriate intervals. Audio-visual alarms for these monitors shall
be available in control room in addition to the local area. Whenever area
monitor gives an alarm, the area should be thoroughly surveyed to ascertain
the radiological status.

4.11.3 Air Monitoring

4.11.3.1 An air-monitoring programme should be designed to meet the following

objectives:

21
 (1) To monitor airborne contamination in the workplace and to assess the
need for appropriate protective equipments;
(2) To help to detect unexpected airborne contamination, if any, to enable
timely protective action; and
(3) To assess the potential intake of radioactive materials through
inhalation and to plan for individual internal monitoring.

4.11.3.2 Airborne radioactivity within the plant areas should be kept below prescribed
level and every effort should be made to minimise it. Air activity
measurements should be carried out from different areas of the plant for
tritium in air, radioactive particulates and radioiodines. Methods for the
collection of air samples, assessment of air activity, frequency of sampling
should be available in the Radiation Protection Procedures. Reference levels
and the protective measures to be taken in case the air activity exceeds the
same should be included. There should be provision for identification of the
radionuclides by spectrometry.

4.11.3.3 A centralised system for collection and analysis of air samples from different
areas of the plant could be employed. Spot samples also shall be collected
for tritium measurements and high volume air samplers for particulates. In
addition, the air activity measurements can be carried out through continuous
air activity monitors for beta-gamma (particulates) and tritium.

4.11.4 Monitoring for Surface Contamination

4.11.4.1 Surface contamination monitoring should have the following objectives:

(1) Assessment of contamination levels in the plant areas and on

equipments and personnel;
(2) Preventing spread of contamination; and
(3) Restricting contamination to approved levels.

4.11.4.2 A system of periodic contamination monitoring of the controlled areas and

supervised areas of the plant should be established to meet the above
objectives. Contamination surveys are normally performed to establish gross
beta-gamma contamination level, but if need be the samples may be
processed for specific type of radiation or specific radionuclides using
spectrometry.

22
 5. DOSE ASSESSMENT

5.1 External Dose Monitoring

5.1.1 Routine individual monitoring shall be done for all personnel entering the
controlled areas.

5.1.2 Dose to site personnel shall be assessed by monitoring of individuals and working
areas by appropriate means. When an individual receives only external radiation,
uniformly distributed over his body, the reading of a properly located personal
dosimeter may be taken to represent the effective dose. When it is known or
suspected that the exposure has been significantly non-uniform, a special
assessment of the dose to individual parts of the body should be done with
additional dosimeters worn at appropriate body locations.

5.1.3 The dosimeters should be used for a specific period and the readings taken to
assess the integrated equivalent dose over the specified period. However, in
suspected or known case of higher exposures, arrangements shall be made to
evaluate the dose from personnel monitoring device such as Thermoluminescent
Dosimeter (TLD) as soon as practicable. Direct Reading Dosimeter (DRD) should
be issued for all jobs under Radiation Work Permit (RWP). DRD readings help to
control day to day exposure of personnel.

5.1.4 The issue of TLD badge shall be authorised by appropriate authorities in the plant
management. All occupational radiation workers shall be allotted a permanent
number for monitoring purposes and the TLD should have the photograph of the
individual.

5.1.5 The visitors should be accompanied by a person duly qualified in Radiation

Protection during their visit in the plant areas. In the case of visitors, group
monitoring may be done instead of individual monitoring.

5.1.6 Any loss of a personnel monitoring device should be promptly reported. Duplicate
shall be issued only after due authorisation and shall be clearly marked. If a
dosimeter, which has recorded a person's long-term exposure is lost before the
exposure has been recorded or if the information it provides is suspect for some
reason, it will be necessary to estimate the external radiation dose by using other
available information such as:

(1) Summing the DRD records for the same period;

(2) Calculating the dose based on the known radiation field in work location
and the exposure period;
(3) Assessing the dose based on the dose received by co-workers; and
(4) Using chromosome aberration analysis wherever applicable.

23
 A record of all these actions should be maintained for future reference.
5.1.7 In the case of personnel working in an area where neutron exposure is likely,
special neutron badges shall also be used to assess the neutron exposure.

5.2 Internal Dose Monitoring

5.2.1 Internal dose monitoring, to assess the committed effective dose (CED) or
committed equivalent dose in the case of uptakes in certain organs, shall be
carried out for all personnel working in controlled areas. This monitoring shall
be performed as per procedures laid down. The type of internal dose monitoring
and periodicity that may be needed depend on the radiological conditions in the
work place.

5.2.2 The internal contamination shall be assessed by the use of methods such as bio-
assay, whole body counting or counting of particular organs of the body. For this,
adequate facilities such as bio-assay laboratory and whole body or single organ
counters should be available. The periodicity of bio-assay sampling should be
such that internal dose can be evaluated properly.

24
 6. MANAGEMENT OF RADIOACTIVE WASTES

6.1 General

6.1.1 All operations in the plant shall be carried out in such a way that the amount of
wastes generated is minimised by appropriate methods and management practices.
The plant management shall develop an integrated strategy for the management of
radioactive wastes. This should include collection, characterisation, segregation,
handling, treatment, conditioning, storage (with due precautions to fire, flood,
earthquake and other site characteristics), transport and disposal.

6.1.2 The plant management shall ensure that gaseous and liquid effluents from NPPs
shall be controlled, monitored and recorded to ensure that the authorised limits are
not exceeded.

6.1.3 Monitoring data should be suitably documented, reported and retained in

accordance with the policies of the operating organisation and the requirements of
the AERB. These data should be routinely analysed to establish trends in relation
to radioactive releases and may be used for optimisation purposes.

6.1.4 The instruments used should be appropriate for the radiation and energies to be
measured and should be routinely calibrated for radiation, energy and effluent
flow rate. The monitoring system should be maintained to provide high
reliability. There should be back up methods for assessing the release in the event
of system failure.

6.1.5 The radiation protection requirements in the handling of radioactive wastes shall
be in accordance with AERB's Safety Guide "Operational Management of
Radioactive Effluents and Wastes Arising in NPPs" (AERB/SG/O-11) [6].

6.1.6 The infrastructure available at site shall be compatible with the scheme of
management of radioactive wastes.

6.2 Monitoring of Solid Wastes

Radioactive solid waste should be characterised, segregated, packed, monitored,

labelled and transported in accordance with the regulations.

6.3 Monitoring of Liquid Wastes

The liquid wastes from the plant shall be segregated depending on their
radioactive characteristics and chemical composition and monitored at
appropriate location to assess the amount of radioactive release. It should be
done by on-line monitoring and by representative sampling. The on line monitors

25
 should have provision for alarms in the control room, in addition to local alarms,
to ensure that the set discharge limits are not exceeded.

6.4 Monitoring of Gaseous Wastes

6.4.1 Gaseous radioactive effluents released through the stack (the preferred single
point exit route) shall be continuously monitored. The radionuclides that are to be
monitored should be identified. The stack monitoring system should provide
plant operators with information on airborne release rates that will assist them in
the control of releases.

6.4.2 The stack monitoring system should provide an accurate measure of the quantity
released for each category of radionuclide that is monitored. These releases shall
be in compliance with regulatory limits.

6.5 Control of Releases

6.5.1 In order to comply with the specified dose limits for the members of the public,
prescribed limits for discharge for gaseous and liquid effluents and release rates
should be established, with the approval of AERB. Reference levels appropriate
for the plant to meet the regulatory requirements also shall be established.

6.5.2 The discharge of gaseous and liquid effluents shall be controlled taking into
account the characteristics of the local environment and the possible bio-
accumulation of the radionuclides therein. Release of radioactive effluents into
the environment should be kept under control by the following means:

(1) Development of reference levels for releases/release rates from the plant;
(2) Proper sampling of the effluents and analysis of the constituent
radionuclides and their amounts at the discharge points;
(3) Development of approved procedures for releasing radioactive materials,
for monitoring the releases and for recording data on each release; and
(4) Assurance of required dilution in the case of liquid wastes.

6.5.3 In the case of gaseous release, equipment such as the following should be used for
source control as appropriate.

- High Efficiency Particulate Air (HEPA) filters for the removal of particulate
activity
- Impregnated charcoal filter beds for the removal of radioactive iodines
- Dryer systems to remove tritiated water vapour from air in Pressurised
Heavy Water Reactors (PHWR)

26
 They must be regularly tested and maintained to ensure optimum performance.
The testing of filters should include in-situ testing for integrity and efficiency of
the filter bank as a whole.

6.5.4 If an uncontrolled release occurs, plant management shall:

(1) Estimate the amount of the release by taking appropriate steps such as
sampling, monitoring, etc., in a timely manner appropriate to the severity
of the case;
(2) If required, initiate emergency actions, as specified in AERB's Safety
Guide “Preparedness of the Operating Organisation for Emergencies at
NPPs” (AERB/SG/O-6) [7];
(3) Report to AERB in accordance with applicable requirements; and
(4) Investigate the event, identify and implement suitable corrective actions.

27
7. STORAGE, HANDLING AND TRANSPORTATION OF
RADIOACTIVE MATERIALS

7.1 General

7.1.1 Radiological control measures shall be established for the storage and handling of
radioactive materials including wastes, within the plant and for their transportation
from the plant in order to prevent the spread of contamination and to limit the
exposure of site personnel and the general public. This should be in accordance
with the provisions of Safety Code for Safe Transport of Radioactive Materials
(AERB/SC/TR-1), Safety Guide on Procedures for Forwarding, Transport,
Handling and Storage of Radioactive Consignments (AERB/SG/TR-3) and Safety
Guide "Operational Management of Radioactive Effluents and Wastes Arising in
NPPs" (AERB/ SG/O-11) [8,9,6].

7.1.2 The personnel involved in the transport of radioactive materials should have
appropriate training to handle an emergency situation during handling or transport
as per Codes and Guides mentioned above.

7.2 Radiological Control Measures

7.2.1 The control measures developed or adopted should be appropriate to the type of
radiation and quantity of radioactive material involved, to its physical and
chemical form.

7.2.2 The following radiation control measures should be used in the safe handling of
radioactive materials at the plant site:

(1) Radiation monitoring to control personnel exposures;

(2) Radiation warning signs, labels and tags;
(3) Separate storage room for radioactive materials and use of approved
containers with due precautions against floods, fire and earthquake;
(4) Procedures for storage and handling;
(5) Procedures for handling emergency situations while storage or transport;
(6) Maintaining an inventory of radioactive materials;
(7) Records of radioactive shipments;
(8) Gate monitors at the exit point, to detect movement of unshielded
radioactive materials if any; and
(9) Review and audits to establish compliance with Radiation Protection
Programme.

28
 8. ENVIRONMENTAL SURVEILLANCE

8.1 An environmental monitoring programme shall be established and implemented in

accordance with the requirements of AERB to assess the dose to the members of
the public due to plant operation. This programme should include comprehensive
monitoring of radionuclide contents in different environmental matrices from
aquatic, atmospheric and terrestrial domains to obtain activity distribution pattern
and to estimate intake of radionuclides by man through inhalation and ingestion
routes towards assessing dose to the member of the public due to plant releases.

8.2 The environmental surveillance shall include the collection and analyses of
various samples such as vegetation, sediment, fish, milk, air, water and other
environmental matrices. The samples should be collected routinely from specified
locations and analysed for any radioactivity which may be attributable to
radioactive releases from the plant.

8.3 A Micro Meteorological Laboratory (MML) also should be established for

conducting the pre-operational studies to establish base line data and for
continued meteorological surveillance. The MML make measurements of
meteorological parameters such as wind speed and direction, temperature, relative
humidity, atmospheric pressure, solar radiation and rain fall. On the basis of real-
time meteorological parameters, the MML can identify the worst affected sector
due to normal and accidental releases. MML data evaluate radiation dose to the
population from radioactive discharges from the plant through air route.

8.4 The staff of Environmental Survey Laboratory (ESL) and MML should undergo
an orientation course on station design features, potential accidents and their
scenarios, source terms, release fractions and environmental behaviour of released
activity.

8.5 Public exposure due to radioactive releases from the NPP shall be evaluated by
ESL at the site. Based on diet survey and the radioactivity data in various dietary
components both from terrestrial (air route) and aquatic sources (water route),
effective dose to local population due to radioactive releases shall be estimated.

8.6 A summary of the environmental surveillance including the public exposure

assessment due to the radioactive releases and micro meteorological
measurements for the year shall be submitted to AERB in addition to informing
the plant management. Based on this report the plant management should plan and
implement corrective measures, if necessary, and send compliance reports to
AERB.

29
 9. FUNCTIONAL INFRASTRUCTURE AND EQUIPMENT

9.1 Health Physics Facilities

9.1.1 Functional infrastructure and equipment such as health physics facilities,

instrumentation and protective equipment shall be provided by the plant
management for effective radiological control. Availability of adequate qualified
persons should also be ensured.

9.1.2 The health physics facilities with adequate space and ventilation that are required
to be provided should include the following:

(1) Health Physics Operations Office: Sufficient office space to

accommodate the health physics staff, permanent records and technical
literature;
(2) Personnel monitoring facilities including thermoluminescent dosimeter
(TLD) storage/issue room, whole body monitoring facilities and exit
monitoring facilities;
(3) Place for storage of TLD racks;
(4) Area for installation of portal surveillance monitors at the final exit
point;
(5) Counting Room: A low radiation background counting room for
performing analysis of air, water and swipe samples;
(6) Sample Preparation Laboratory with fume hood facility for preparation
of samples;
(7) Instrumentation Calibration Facility: Designed and located such that
radiation in the calibration area should not interfere with low level
monitoring or counting systems;
(8) Source Room with adequate shielding; and
(9) Computerised centre for data storage management, retrieval of station
radiological and personnel dose and other safety related data.

9.1.3 The other facilities that are required to be provided with adequate space and
ventilation shall include the following:

(1) Change Room: Change room with personnel decontamination facility,

lockers, receptacles for contaminated clothing etc.;
(2) Personnel Decontamination Facility with showers, wash basins and
contamination monitors;
(3) Equipment Decontamination Facility to clean, and decontaminate
equipment and hand tools;
(4) Transit waste storage room;
(5) Storage Facility for contaminated equipment/tools;
(6) Active Workshops;

30
 (7) First Aid Room;
(8) Laundry;
(9) Meteorological Data Collection Facility;
(10) Servicing & Testing of protective equipment; and
(11) Communication Facilities.

9.2 Equipment

9.2.1 The plant management shall provide equipment necessary for the Radiation
Protection Programme, such as various monitoring and analysis instruments, and
other protective equipment. The selection and quantity of instrumentation and
equipment should be based on the anticipated needs of the NPP during normal
operations, major outages and anticipated operational occurrences and accident
conditions.

9.2.2 The instruments and equipment provided shall include:

(1) Counting and analysis instruments;

(2) Portable radiation monitoring instruments;
(3) Personnel dose measuring instruments;
(4) Protective Clothing:
(a) Protective clothing such as coveralls, lab. coats, plastic suits
ventilated plastic suits gloves, shoe covers, gum boots etc.
(b) Respiratory protective equipment: Oro-nasal respirator, full face
respirator, air-line respirator, tritium respirator and self-
contained breathing apparatus.
(5) Contamination monitors for installation at rubber areas, interzonal and
final exit points;
(6) Standard sources for calibration of instruments; and
(7) Air sampling devices.

31
 10. TRAINING AND QUALIFICATION

10.1 Training in radiation protection is a basic requirement for personnel working in

the station. The access and the type of work permitted depend on the established
levels of plant system training, skills and qualification in radiation protection
courses. The training should include radiation protection procedures and
radiation emergency procedures.

10.2 Job oriented training is necessary to ensure that everyone attains and maintains the
level of competence required for his duties and for his level of responsibility. This
will help to minimise exposure times and to have better contamination control to
achieve lower individual and collective doses.

10.3 The operating organisation shall make the necessary arrangements to train its site
personnel. All plant personnel must attain and maintain an adequate level of
capability in radiation protection in accordance with the requirements of their job
functions.

10.4 The radiation protection training programme should also address the training
needs of temporary workers. The temporary workers must possess adequate
knowledge and skills for the work they are to perform. Only after pre-
employment medical examination and training in radiation protection, the
individuals should be permitted to work in controlled areas. They shall be directly
supervised by personnel qualified in Radiation Protection Procedures for jobs in
radiation areas.

10.5 Training courses should also be held periodically for doctors and para-medical
personnel covering all aspects of radiation protection and handling of
contaminated/exposed persons.

10.6 Training objectives shall be reviewed periodically to ensure that they take into
account any change in requirements. The training programme should reflect the
contemporary knowledge and techniques. This should be an on going programme.
Periodic refresher course shall be conducted to keep qualification up-to-date. The
record of the training and qualification status for all the personnel should be
maintained at the station.

32
 11. EMERGENCY PREPAREDNESS

11.1 Emergency Plan

11.1.1 The plant management shall develop its respective emergency preparedness plans to
handle any emergency arising out of a potential accident. The emergency plan
shall have provisions to take necessary measures for coping with such situation
and ensure readiness of persons, facilities and equipments and for effective co-
ordination between various groups at site and off-site including public authorities.
The emergency plan forms a part of the requirements to be satisfied for granting
operating licence by the AERB.

11.1.2 Separate but inter connected emergency plans shall be prepared to handle plant,
site and off-site emergencies. The plant and site emergency plan shall be
prepared by the plant management as per AERB Safety Guide "Preparedness of
the Operating Organisation for Emergencies at NPPs" (AERB/SG/O-6) [7].

11.1.3 In the case of multi-facility site an integrated emergency plan shall be available
which should include actions by personnel at different facilities in the event of an
accident at any plant.

11.1.4 The guidelines to prepare site emergency plan are given in AERB Safety Manual
"Site Emergency Plan for Nuclear Installations" (AERB/M/NISD-1) [10]. The
plan shall be periodically reviewed and updated. The procedures for handling off-
site emergencies are given in the AERB Safety Manual "Off-Site Emergency Plan
for Nuclear Installations" (AERB/M/NISD) [11].

11.2 Facilities and Equipment Provided at Site for Handling Emergency:

11.2.1 Adequate facilities and equipment for handling emergency at site as necessary
shall be provided by the plant management.

11.2.2 The facilities located within the site shall include a first-aid room and facilities for
decontamination of personnel, equipment and areas.

11.2.3 The personnel handling an emergency shall be provided with dosimeters and
appropriate protective clothing and equipment. Sufficient equipment should be
readily available as per approved list to meet the minimum requirements
envisaged during an emergency situation.

11.2.4 High-range radiation instruments, including direct reading dosimeters and self
powered air samplers shall be readily available for use during an emergency
situation. High-range fixed area gamma monitors with recorder facility also
should be provided at selected locations for rapid assessment of an emergency

33
 situation. Provisions should be available to obtain information on the radiological
status (both radiation field and airborne activity concentration) inside the
containment without entry into the containment in the case of accidental
situations.

11.2.5 A radiation emergency vehicle equipped with radiation monitoring equipments

and communication facility shall be available.

11.2.6 All equipment for emergency situations shall be periodically inspected and tested,
as necessary to ensure that they are functionally available on demand.

11.3 Training and Exercises

Emergency exercises shall be conducted at regular intervals as stipulated by

AERB. Provision shall be made for training of the personnel who would be
involved in implementation of the plan. There should be a post exercise critique
to review all the positive points and deficiencies observed during the exercise.
Records of all exercises along with deficiencies observed shall be maintained for
review and rectification.

11.4 Intervention Levels (IL)

11.4.1 The intervention levels as specified by AERB shall be applied for implementation
of appropriate counter measures such as, use of protective equipments,
introduction of prophylaxis, sheltering, or evacuation in the case of radiation
emergencies. The intervention levels (ILs) and derived intervention levels (DILs)
given in the AERB Safety Guide "Intervention Levels and Derived Intervention
Levels for Off-Site Radiation Emergency" (AERB/SG/HS-1)[12] shall be
incorporated in Radiation Emergency Procedures.

11.5 Limitation of Exposure in Emergency

11.5.1 Accident analysis of the plant shall demonstrate that the dose resulting from the
Design Basis Accident (DBA) at the site boundary is not in excess of the values
stipulated by the AERB in this regard.

11.5.2 In an emergency situation, some people would be exposed to doses exceeding the
prescribed limits for occupational workers, for saving life, preventing serious
injury or to prevent substantial increase in the scale or magnitude of the accident.
The exposures of such personnel should not be allowed to exceed the values
prescribed by AERB in this regard.

11.5.3 Such exposures shall be duly authorised and purely voluntary. The volunteers
should be informed about the risk involved in such exposures. They shall be
provided with adequate monitoring devices and protective equipments. The doses

34
 received by such personnel shall be entered in the dose records along with the
cross-reference to the report on emergency conditions.

11.6 Assessment and Monitoring

Radiological conditions in the plant site areas shall be assessed in the event of an
emergency as appropriate towards confirming the radiological conditions for the
purposes of exposure control. The monitoring shall include assessment of
radiation levels, airborne contamination, surface contamination and effluent
releases.

35
 12. DOCUMENTATION

12.1 General

Records shall be kept in order to maintain up-to-date information concerning

important aspects of the Radiation Protection Programme. The information
retained will be of great aid in reviewing and assessing the Radiation Protection
Programme.

12.2 Types of Records

The relevant documents be maintained should include:

(1) Health physics reports;

(2) Personnel dose and previous radiation history;
(3) Radiation surveys and monitoring;
(4) Radiation Work Permits;
(5) Instrument calibration;
(6) Inventory of protective equipment and their periodic verification and test
results;
(7) Inventory of sources and radioactive materials;
(8) Radioactive shipments;
(9) Radioactive liquid, gaseous and solid waste storage and disposal;
(10) Medical surveillance;
(11) Bio-assay and whole body counting;
(12) Over exposure and unusual occurrence reports;
(13) Emergency exercise records; and
(14) Training and qualification records.

12.3 Criteria for Preparing and Maintaining Records

12.3.1 Records should be prepared in a systematic manner and as per the requirement of
the AERB.

12.3.2 The dose records of the temporary workers should be maintained separately.
Exposures received prior to employment should be supported by valid documents.

12.3.3 The dose and health records of all the occupational workers shall be maintained
by the plant management for the duration of the working life of each worker and
afterwards until the worker attains or would have attained the age of 75 years, and
for not less than 30 years after the termination of the work involving occupational
exposure whichever is later.

36
12.3.4 The records classified as long term, medium term and short term shall be kept in a
safe and protected environment for specified periods. The long term records shall
be maintained for the life of the plant, medium term for a period of 10 years and
short term for a period of 3 years.

37
 APPENDIX-I

EXPLANATORY NOTES ON RADIATION PROTECTION CONCEPTS AND

TERMS USED IN THIS GUIDE

I-1 Introduction

I-1.1 The radiation protection concepts and terms used in this Guide are based on the
"Recommendations of the International Commission on Radiological Protection"
(ICRP-60, 1990) [3] and the International Basic Safety Standards for Protection
against Ionising Radiation and for the Safety of Radiation Sources (IAEA Safety
Series No. 115, 1996) [2].

I-1.2 Practices and Intervention

Any human activity which increases the overall exposure to radiation is called a
"practice". Any human action intended to reduce or avert exposure to sources
which are not part of controlled practices or which are out of control as a
consequence of an accident is called "intervention".

I-1.3 Radiation protection is concerned with the protection of the individuals, their
progeny and mankind as a whole, while still allowing necessary activities from
which radiation exposure might result.

I-2 Biological Aspects of Radiological Protection

The harmful biological effects of exposure to ionising radiation have been divided
into two types (1) Deterministic (Non-stochastic) effects and (2) Stochastic
effects. The aim of radiation protection is to limit the probability of stochastic
effects to an acceptable level and to avoid deterministic effects by setting dose
limits below the thresholds for such effects.

I-3 Dosimetric Quantities used in Radiation Protection

I-3.1 Absorbed Dose (DT)

The fundamental dosimetry quantity D defined as D = d∈/dm, where d∈ is the

mean energy imparted by ionising radiation in a volume element and dm is the
mass of matter in the volume element. The energy absorbed per unit mass in a
tissue or an organ is called the absorbed dose of the tissue. The SI unit of the
absorbed dose is joule per kilogram, termed the gray (Gy).

1 Gy = 1 J.kg-1

38
I-3.2 Equivalent Dose (HT,R)

The quantity HT,R, defined as:

HT,R = DT,R.wR

where DT,R is the absorbed dose delivered by radiation type R averaged over a
tissue or organ T and wR is the radiation weighting factor for radiation type R.

When the radiation field is composed of different radiation types with different
values of wR, the equivalent dose is:

HT = ∑ wR.DT,R
R

The unit of equivalent dose is J.kg-1, termed the sievert (Sv).

1 Sv = 1 J.kg-1

I-3.3 Radiation Weighting Factor (wR)

The probability of biological effect of radiation is found to depend not only on

the absorbed dose, but also on the type and energy of the radiation causing the
dose. This is taken into account by weighting the absorbed doses by a factor
called radiation weighting factor wR, related to the quality of the radiation
(previously known as quality factor). The values of wR for different types of
radiation are given in Table-1.

Definition: Multipliers of absorbed dose used for radiation protection purposes to

account for the relative effectiveness of different types of radiation in inducing
health effects.

I-3.4 Effective Dose

The quantity E, is defined as a summation of the tissue equivalent doses, each

multiplied by the appropriate tissue weighting factor:

E = ∑ w .H
T T T

where HT is the equivalent dose in tissue T and wT is the tissue weighting factor
for tissue T. From the definition of equivalent dose it follows that:

E = ∑ w .∑ w .D
T T R T T .R

39
 where wR is the radiation weighting factor for radiation R and DT,R is the average
absorbed dose in the organ or tissue T. The unit of effective dose is J.kg-1, termed
the sievert (Sv).

I-3.5 Tissue Weighting Factor (wT)

The probability of stochastic effect due to a given dose differs from one organ or
tissue to another. In order to combine the effect of doses to different tissues in a
way which will likely correlate well with the total of the stochastic effects, a
weighting factor called tissue weighting factor, wT is used. This factor represents
the relative contribution of the organ or tissue to the total detriment resulting from
uniform irradiation of the whole body. The values of wT are given in Table-2.

Definition: Multipliers of the equivalent dose to an organ or tissue used for

radiation protection purposes to account for the different sensitivities of different
organs or tissues to the induction of stochastic effects of radiation.

I-3.6 Committed Equivalent Dose HT(τ)

Tissue irradiation from incorporated radionuclides is spread out in time, energy

deposition occurring as the radionuclide decays. The time integral of the
equivalent dose rate in an organ or tissue T, following intake of a radionuclide is
called committed equivalent dose.

The quantity H(τ), defined as:

t0+τ
HT(τ) = ∫ HT(τ) dt
t0

where t0 is the time of intake , HT(τ) is the equivalent dose rate at time t in an
organ or tissue T and τ is the time elapsed after an intake of radioactive
substances. When τ is not defined it will be taken to be 50 years for adults and to
age 70 years for intakes by children.

I-3.7 Committed Effective Dose E(τ)

The time integral of the effective dose rate following an intake of a radionuclide is
called the committed effective dose.

40
 The quantity E(τ), defined as:

E(τ) = ∑wT.HT(τ)
T
where HT(τ) is the committed equivalent dose to tissue T over the integration time
τ. When τ is not specified, it will be taken to be 50 years for adults and to age 70
years for intakes by children.

I-3.8 Collective Effective Dose (S)

The total effective dose of a group of persons on which the total biological effect
in that group would depend is called the collective effective dose.

The total effective dose S to a population, defined as:

S = ∑ E .N
i i i

where Ei is the average effective dose in the population subgroup i and Ni is the
number of individuals in the subgroup.

41
 Table 1

Radiation Weighting Factors (wR)

Type and energy range wR

Photons, all energies 1

Electrons, and muons, all energies 1

Neutrons, energy

< 10 keV 5

10 keV to 100 keV 10

> 100 keV to 2 MeV 20

> 2 MeV to 20 MeV 10

> 20 MeV 5

Protons, other than recoil protons energy > 2 MeV 5

Alpha particles, fission fragments, heavy nuclei 20

42
 Table 2

Tissue Weighting Factors (wT)

Tissue or organ wT

Gonads 0.20
Bone marrow 0.12
Colon 0.12
Lung 0.12
Stomach 0.12
Bladder 0.05
Breast 0.05
Liver 0.05
Oesophagus 0.05
Thyroid 0.05
Skin 0.01
Bone surface 0.01
*Remainder 0.05

43
 APPENDIX-II

DOSE LIMITS AND CONSTRAINTS

II-1 Occupational Dose Limits

II-1.1 Primary Limits

II-1.1.1 Limit on Life-time Dose

(a) The cumulative life-time occupational effective dose of a worker shall

not exceed 1 Sv.
(b) Dose Constraint: Medical review shall be undertaken at cumulative
occupational effective dose of 0.5 Sv to the workers.

II-1.1.2 Effective Dose Limit: The cumulative effective dose constraint for five years
from January 1, 1994 to December 31, 1998 is 100 mSv for individual
radiation workers. The annual effective dose to individual workers in any
calendar year during the five-year block shall not exceed 30 mSv. This is
given in Appendix-III.

II-1.1.3 Dose Limits for the lens of eye, skin and hands and feet: For the lens of the
eye and the skin the dose limit shall be 150 mSv and 500 mSv of equivalent
dose in a year respectively. The equivalent dose limit for hands and feet also
shall be 500 mSv in a year. The skin dose shall be averaged over 1 cm2, if
the exposed area is ≤10 cm2. It shall be averaged over the actual exposed
area if it is >10 cm2.

II-1.1.4 Dose Limit for Female Workers: There is no special occupational dose limit
for women in general. However, once pregnancy is declared the equivalent
dose limit to the surface of the woman's abdomen shall be 2 mSv for the
remainder of the pregnancy and the limit on intake of radionuclides shall be
1/20 of the Annual Limit on Intake (ALI).

II-1.1.5 Dose Limit for Apprentices and Trainees: Apprentices and trainees in
nuclear installations, if any, shall be of age not less than 16 years. For the
apprentices of age between 16 and 18 years the annual dose limits shall be an
effective dose of 6 mSv, an equivalent dose of 50 mSv to the lens of the eye
an equivalent dose of 150 mSv to the extremities or skin.

II-1.1.6 Dose Limit for Temporary Workers: For temporary workers employed, the
dose received shall not exceed the limit stipulated by the AERB. Temporary
workers shall be employed as per approved procedures.

44
II-1.2 Dose Limits (Internal Exposure)

II-1.2.1 Annual Limit on Intake (ALI): The internal exposures are to be controlled
by applying the secondary limit called annual limit on intakes (ALI). This
would correspond to an intake which would result a committed effective dose
of 20 mSv. The limit on internal exposures for individual radionuclides shall
be 1 ALI for each year. If the intake is from more than one radionuclide,
total intake during any one year shall be limited such that :

∑I
i i
≤1
( ALI )
i

where Ii refers to intake of i th radionuclide to which a worker is exposed and

(ALI)i is the annual limit on intake for radionuclide i . The limits on some
specific radionuclides as prescribed by AERB are given in Appendix-IV.

II-1.3 Derived air concentration (DAC): DAC values of individual radionuclides

should be used for controlling the internal exposures due to inhalation. DAC
of a radionuclide in Bq m-3 of air is given by

ALI of the given radionuclide

DAC =
2.4 × 103

where 2.4×103 m3 is the volume of the air inhaled by ICRP reference man in a
working year.

However in the case of HTO, the DAC value has to be obtained by

multiplying the above value by 2/3 for accounting 50% absorption through
skin.

II-2 Dose Limits for Members of the Public

II-2.1 Control of public exposures under all normal operations of the plant shall be
exercised by application of controls at source rather than in the environment.
Limits are applied to doses incurred as a result of practices. The public
exposure limits do not include occupational exposure and exposures from the
natural environment and medical exposures.

II-2.3 The limit of public exposure shall be an effective dose of 1 mSv in a year for
external exposure. If the exposures are both external and internal, the limit of

45
1 mSv in a year shall be applicable to the sum of the effective dose from
external exposure and 70 year committed effective dose from the intake
incurred during the year.

46
 APPENDIX-III

DOSE LIMIT FOR OCCUPATIONAL EXPOSURES AS APPROVED BY AERB

IN MARCH 1994.

I. Effective Dose Limits

(i) The cumulative effective dose constraint for five years from January 1, 1994
to December 31, 1998 will be one hundred milliSievert (100 mSv) for
individual radiation workers.

(ii) The annual effective dose to individual workers in any calendar year during
the five-year block shall not exceed the limit of thirty milliSievert (30 mSv).

II. Investigation Levels

(i) Individual effective dose exceeding twenty milliSievert (20 mSv) in a year
shall be investigated by a committee to be constituted by Chairman, AERB
for this purpose. The committee shall ensure that the five-year constraint of
not exceeding one hundred milliSievert (100 mSv) is met in all cases.

(Ref.: AERB Safety Directive 6/1994 dated March 10. 1994)

47
 APPENDIX-IV

ANNUAL LIMIT OF INTAKE (ALI) (INHALATION) VALUES FOR

RADIONUCLIDES OF COMMON INTEREST

Nuclide Aerosol class ALI (1991)

(Bq)
H-3 (tritiated water vapour) D 1 x 109
Co-60 W 2 x 106
Y 4 x 105
Sr-90 D 4 x 105
Y 6 x 104
I-131 D 1 x 106
Cs-137 D 2 x 106
Th-232 W 9 x 101
Y 9 x 101
U-238 D 9 x 104
W 1 x 104
Y 6 x 102
Pu-239 W 3 x 102
Y 3 x 102
Am-241 W 3 x 102

Ref.: AERB Safety Directive 2/91 dated July 22, 1991

Note: D, W and Y represent the retention half-times of inhaled radioactive materials from
the pulmonary region

D < 10 days
W 10-100 days
Y > 100 days

48
 REFERENCES

1. INTERNATIONAL ATOMIC ENERGY AGENCY, Radiation Protection during

Operation of Nuclear Power Plants; IAEA Safety Series No.50-SG-05. (1983)

2. INTERNATIONAL ATOMIC ENERGY AGENCY, International Basic Safety

Standards for Protection Against Ionising Radiation and for the Safety of
Radiation Sources; IAEA Safety Series 115. (1996)

3. INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,

Recommendations of the International Commission on Radiological Protection;
ICRP Publication 60. (1990)

4. ATOMIC ENERGY REGULATORY BOARD, Radiation Protection for Nuclear

Facilities; AERB Safety Manual. (1996)

5. ATOMIC ENERGY REGULATORY BOARD, Design Aspects of Radiological

Protection for PHWR based Nuclear Power Plants; AERB Safety Guide No.
AERB/SG/D-12 (under preparation)

6. ATOMIC ENERGY REGULATORY BOARD, Operational Management of

Radioactive Effluents and Wastes Arising in NPPs; AERB Safety Guide No.
AERB/SG/O-11 (under preparation).

7. ATOMIC ENERGY REGULATORY BOARD, Preparedness of the Operating

Organisation for Emergencies at NPPs; AERB Safety Guide No. AERB/SG/O-6
(under preparation).

8. ATOMIC ENERGY REGULATORY BOARD, Safe Transport of Radioactive

Materials; AERB Safety Guide No. AERB/SG/TR-1.

9. ATOMIC ENERGY REGULATORY BOARD, Procedures for Forwarding,

Transport, Handling and Storage of Radioactive Consignments; AERB Safety
Guide No. AERB/SG/TR-3.

10. ATOMIC ENERGY REGULATORY BOARD, Site Emergency Plan for Nuclear
Installations; AERB Safety Manual No. AERB/M/NISD-1.

11. ATOMIC ENERGY REGULATORY BOARD, Off-Site Emergency Plan for

Nuclear Installations; AERB Safety Manual No. AERB/M/NISD.

12. ATOMIC ENERGY REGULATORY BOARD, Intervention Levels and Derived

Intervention Levels for Off-Site Radiation Emergency; AERB Safety Guide No.
AERB/SG/HS-1.

49
50
 LIST OF PARTICIPANTS OF WORKING GROUP
Dates of meeting: October 18 & 19, 1993
April 04 & 05, 1994
August 02 & 03, 1994

Shri M. Sundaram (Convenor) Head, PPSS, HPD, BARC

Shri R.M. Sharma Health Physicist-in-Charge, RAPS
Shri S.K. Fotedar Directorate of Operations, NPC
Shri K.K. Narayanan (Co-opted) PPSS, HPD, BARC
Shri S.A. Hussain (Co-opted) Health Physics Unit (BARC), KAPP
Shri George Thomas (Member-Secretary) HSD, AERB

51
 ADVISORY COMMITTEE ON CODES, GUIDES AND
ASSOCIATED MANUALS FOR SAFETY IN OPERATION OF
NUCLEAR POWER PLANTS (ACCGASO)

Dates of Meeting: February 23 & 24, 1995

September 19 & 20, 1995
November 27 & 28, 1995

Shri G.V.Nadkarni (Chairman) Formerly Director E & PA, NPCIL

Shri V.S.Srinivasan NPCIL
Shri Y.K.Joshi RAPS, NPCIL
Shri Ravindranath TAPS, NPCIL
Shri V.V.Sanathkumar MAPS, NPCIL
Shri R.S.Singh AERB
Shri Ram Sarup AERB
Shri S.T.Swamy (Co-opted) AERB
Shri S.K.Warrier (Member-Secretary) AERB

52
 ADVISORY COMMITTEE ON NUCLEAR SCIENCES (ACNS)

Dates of Meeting: January 11, 1997

February 22, 1997

Shri S.K.Mehta (Chairman) Formerly Director RG, BARC

Shri S.M.C.Pillai President, Nagarjuna Group
Prof. M.S.Kalra IIT, Kanpur
Prof. U.N.Gaitonde IIT, Mumbai
Shri S.K.Goyal BHEL
Shri Ch.Surendar NPCIL
Shri S.K.Sharma BARC
Dr. V.Venkatraj BARC
Shri V.K.Chaturvedi NPCIL
Shri M.S.Kumra BARC
Shri S.P.Singh Formerly Head, NSD, AERB
Shri G.K.De AERB
Smt. Usha A Menon AERB

53
 PROVISIONAL LIST OF SAFETY GUIDES ON OPERATION OF NUCLEAR
POWER PLANTS

Safety Series No. Provisional Title

AERB/SG/O-1 Training and Qualification of Operating Personnel of NPPs
AERB/SG/O-2 In-Service Inspection of NPPs
AERB/SG/O-3 Operational Limits and Conditions for NPPs
AERB/SG/O-4 Commissioning Procedures for NPPs
AERB/SG/O-5 Radiation Protection during Operation of NPPs
AERB/SG/O-6 Preparedness of the Operating Organisation for Emergencies at
NPPs
AERB/SG/O-7 Maintenance and Modifications of NPPs
AERB/SG/O-8 Surveillance of Items Important to Safety in NPPs
AERB/SG/O-9 Management of NPPs for Safe Operation
AERB/SG/O-10 Core Management and Fuel Handling for NPPs
AERB/SG/O-11 Operational Management of Radioactive Effluents and Waste
Arising in NPPs

54