SUB-CHAPTER: L.2
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
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CHAPTER L: RADIATION PROTECTION
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SUB-CHAPTER L.2. DEFINITION OF RADIOACTIVE
SOURCES IN THE PRIMARY CIRCUIT
During normal plant operation (stabilised operation or shutdown transient conditions), the dose
rates in the vicinity of plant systems and components arise from activity within the relevant area
and/or surface contamination.
The specific activities are explained in detail in Chapter K.1. In radioprotection, the following two
source terms are used:
-
operational source term data is used to calculate worker dose estimates and for defining
the facility Pressurised Nuclear Equipment (ESPN) classification (see Chapter C.2),
-
the biological protection design basis source term is used as a design parameter for EPR
buildings, systems and shielding provisions.
Surface deposited activity, which represents a major contributor to worker dose due to ionising
radiation, is detailed in Section 2 within Sub-chapter L.2.
Design basis values, operational data, as well as compliance with layout and design principles for
complex system, all play a part in the optimisation of worker radioprotection.
Moreover, peak values are not generally used for shielding calculations, with the exception of the
primary coolant purification system and reactor shut-down cooling system filters. For other
systems, the peak has only a relatively low impact on the shielding design because of the short
life span of the nuclides involved. Experience shows that its influence on worker exposure is
negligible.
1.
SELECTION OF NUCLIDES FOR RADIOPROTECTION AND
ESPN CLASSIFICATION
Because of their radioactive characteristics, some of the radionuclides considered in discharge or
accident studies have no significant impact on radioprotection.
The radiological impact of radionuclides with a short radioactive half-life such as nitrogen-16 (7.3
s) and nitrogen-17 (4.2 s) becomes totally negligible after a few minutes simply due to radioactive
decay.
The radiation emitted by the pure beta emitting radionuclides such as carbon-14 or nickel-63
have sufficiently low energy that they can be stopped by the structures designed to protect
against gamma emitting radionuclides.
The primary circuit inventory of radionuclides which may have an effect on the protection of
workers against ionising radiation, is given in table L.2 TAB 1 for stabilised operation and in table
L.2 TAB 2 for power transients (load reduction for fission products and oxygenation peak for
corrosion products).
In order to cover the most recent operational feedback from German and French units, chrome51, silver-110m, antimony-124 and antimony-122 will be included in the pre-operational safety
report.
SUB-CHAPTER: L.2
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
CHAPTER L: RADIATION PROTECTION
2.
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ACTIVATED CORROSION PRODUCT DEPOSITION
During the unit operating life, radioactive materials are deposited on the internal surfaces of the
pipework, vessels, tanks, etc, and can accumulate to form a contamination deposit (mobile
and/or fixed). This accumulation of contamination is a continuous process which depends mainly
on the physical and chemical condition of the primary circuit (RCP) [RCS] water in different
reactor states (full power and shutdown states).
The distribution of corrosion products in the water depends on the reactor state.
Operational plant data for Co-58 and Co-60 in the primary circuit of French power stations are
shown below.
Operational data for activity deposited in French power stations
Power station
Hot leg
Cross-over leg
(Bq/m2)
(Bq/m2)
Nuclide
900 MWe
Blayais 4
1300 MWe
Penly
Co-58
3.3E+09
3.3E+09
Co-60
1.6E+09
2.1E+09
Co-58
2.1E+09 – 2.6E+09
4.3E+09
Co-60
6.0E+08 – 7.0E+08
1.1E+09
Specific concentrations and deposited activity of important nuclides in the main loops have been
identified following an analysis of measurements taken in French power stations and German
Konvoi power stations. The deposited values are given for the primary loop in table L.2 TAB 3.
The deposited values for Cobalt-58 and Cobalt-60 are used for dose rate calculations at different
locations on the station. The values obtained are adequately representative, as these nuclides
are responsible for the largest share of dose rates at shutdown.
To estimate dose rate on other pipework or equipment, existing dose rates are scaled using
linear interpolation using the Cobalt activity deposited in the primary loops. For certain
components, variations in the shielding thickness (steel, water) are also taken into account.
In order to get the most recent operational feedback from German and French units, deposited
activity from chrome-51 will be covered in the pre-operational safety report.
3.
SPECIFIC ACTIVITIES AND ACTIVITY INVENTORY
3.1.
PRIMARY CIRCUIT
Radioactive materials in the primary circuit and connected systems, arise from:
SUB-CHAPTER: L.2
FUNDAMENTAL SAFETY OVERVIEW
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VOLUME 2: DESIGN AND SAFETY
CHAPTER L: RADIATION PROTECTION
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fission products released through defects in fuel rod cladding during operation of the unit.
-
residual contamination with uranium oxide originating from the scattering of fissile material
during preceding campaigns and/or from the manufacturing process.
-
corrosion products activated by the neutron flux in the core,
-
primary coolant activation products, for example, 3H (tritium) or 16N (nitrogen 16).
For the primary coolant, two types of activity values have been selected to characterise normal
full power operating conditions:
-
the average expected value (“best estimate” conditions) or “operational data” (see note 1),
-
95% confidence values or “Design basis” values (pessimistic values chosen to bound
95% of operational data from French and German 900 and 1300 PWRs).
The specific activities of the fission and of corrosion products in the primary coolant are given in
tables L.2 TAB 1 and L.2 TAB 2.
The peak shutdown factors for corrosion and fission products used to define the specific transient
activities are also given for information.
Note 1 : The operational data is all-encompassing scenarios at more than 95% of the values
measured over about twenty cycles for the RCP of the N4 plant series and at 100% for
the other systems.
3.2.
PRIMARY CIRCUIT CHEMICAL AND VOLUME CONTROL
Table L.2 TAB 1 gives the concentrations of activity in the RCV [CVCS] discharge line, with the
exception of Nitrogen-16 (they are identical to those of the primary system).
The rate of purification in normal operating conditions is 36 Mg/hr.
For a short period before shutdown (72 hr) the purification rate may increase up to 72 Mg/hr.
3.3.
PRIMARY COOLANT PURIFICATION SYSTEM
The concentration of radioactivity (ionic or particulate species) in the primary coolant is reduced by
running the demineraliser continuously at a purification rate of 10%/hr or 20%/hr.
3.4.
PRIMARY COOLANT DEGASIFICATION SYSTEM
The primary coolant degasification system is mainly used to eliminate hydrogen from the primary
circuit before the unit is shut down and to reduce radioactive gases in the primary coolant during
normal operating conditions and particularly before shutdown.
SUB-CHAPTER: L.2
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
CHAPTER L: RADIATION PROTECTION
3.5.
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PROCESSING AND STORAGE OF PRIMARY EFFLUENTS
The fluid extracted from the primary circuit to compensate for the injection of demineralised water
and boric acid for the long-term control of reactivity is transferred to the primary effluents
processing and storage system. In the primary effluent processing system, demineralised water
and 4% boric acid are recovered by evaporation of the borated fluid. If necessary, the boric acid
and the demineralised water recovered in this way may be re-injected into the primary system to
increase or reduce the boron concentration as required.
3.6.
GASEOUS WASTE PROCESSING SYSTEM
The liquid storage tanks, the pressuriser expansion tank, the volume control tank (only during
shutdown), the deaerators, the boric acid tanks, the evaporators and several bleed tanks are
connected to the gaseous waste processing system (TEG) [CSTS]. This system comprises two
functional parts:
-
the bleed unit,
-
the decay unit,
The Gaseous waste processing system (TEG) [CSTS] system is mainly made up of nitrogen. The
quantity of radioactive isotopes in these units depends on whether the primary system is
degassed.
3.7.
SAFETY INJECTION/ REACTOR HEAT REMOVAL AT SHUTDOWN
To shut down the plant under normal operating conditions, the reactor is placed in cold shut down
conditions using the reactor heat removal system (RIS/RRA) [SIS/RHRS].
When the SIS/RHR is in use, a section of this system conveys primary cooling water with the
specific activity of primary coolant. Under other normal operating conditions, the system contains
fluid from the in-containment refuelling water storage tank (IRWST).
For design purposes, the activity concentrations applicable when the reactor vessel is opened
are assumed.
3.8.
SPENT FUEL POOL WATER TREATMENT AND HEAT REMOVAL
Radioactive impurities in the spent fuel pool water and in the pool heat removal system result
from:
-
the release of fission products from failed fuel rods,
-
the deposition of activated corrosion products on the surface of the stored fuel rods,
-
the transport of small quantities of primary coolant via the transfer tube during fuel transfer
operations.
SUB-CHAPTER: L.2
FUNDAMENTAL SAFETY OVERVIEW
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CHAPTER L: RADIATION PROTECTION
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The pool water is purified by the pool water purification and cooling system (PTR) [FPCS]. The
resulting activity must be maintained at as low a level as possible, so that dose rates on the
operating floors access routes (Reactor Building and Fuel Building) comply with a green zoning
classification.
3.9.
MAIN STEAM PIPEWORK
Normally, secondary systems do not contain any radioactivity, as they are completely segregated
from the active primary cooling system within the steam generators.
Activity may only enter the secondary system and hence the main steam system of connecting
systems, following a leak in a steam generator. During normal operations, a small leak of 3 L/hr
per Steam Generator has been considered to calculate design basis activity in the main steam
system.
SUB-CHAPTER: L.2
UK-EPR
FUNDAMENTAL SAFETY OVERVIEW
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STANDARD
TAB 1:
SPECIFIC CONCENTRATIONS OF RADIONUCLIDES IN THE PRIMARY
CIRCUIT AFFECTING RADIOPROTECTION :
NORMAL OPERATION
NUCLIDE
Mn-54
SPECIFIC ACTIVITY (Bq/Mg)
Biological
Actual plant data
protection devices
design basis
2.0E+06
4.0E+06
Co-58
8.0E+06
1.6E+07
Fe-59
5.0E+05
1.0E+06
Co-60
5.0E+05
1.0E+06
Cr -51
X
X
Ag -110m
X
X
Sb -124
X
X
Sb -122
X
X
Ar -41
3.0E+08
1.0E+09
Kr-85m
2.0E+08
5.5E+09
Kr-85
1.9E+07
5.2E+08
Kr-87
3.6E+08
1.0E+10
Kr-88
5.0E+08
1.4E+10
Xe -133m
1.1E+08
1.7E+09
Xe -133
5.0E+09
8.0E+10
Xe -135
1.1E+09
1.8E+10
Xe -138
8.5E+08
1.4E+10
Sr-89
3.0E+05
4.9E+06
Sr-90
1.9E+03
3.0E+04
I -131
1.0E+08
1.6E+09
I -132
1.8E+08
2.3E+09
I -133
3.1E+08
4.9E+09
I -134
1.1E+08
1.7E+09
I -135
2.0E+08
3.3E+09
Cs-134
4.0E+07
3.2E+08
Cs-136
X
X
Cs-137
4.0E+07
3.2E+08
Cs-138
8.5E+08
1.4E+10
H-3
3.7E+10
3.7E+10
X: Values to be included later using most recent operational feedback.
SUB-CHAPTER: L.2
FUNDAMENTAL SAFETY OVERVIEW
UK-EPR
VOLUME 2: DESIGN AND SAFETY
CHAPTER L: RADIATION PROTECTION
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TABLE
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STANDARD
TAB 2:
SPECIFIC CONCENTRATIONS OF RADIONUCLIDES IN THE PRIMARY
CIRCUIT AFFECTING RADIOPROTECTION:
SHUTDOWN TRANSIENT
NUCLIDE
SPECIFIC ACTIVITY (Bq/Mg)
Biological protection
Actual Plant data
devices design basis
Peak factor
Mn-54
6.0E+08
1.2E+09
300
Co-58
8.0E+10
1.6E+11
10000
Fe-59
1.5E+08
3.0E+08
300
Co-60
Cr -51
2.5E+08
X
5.0E+08
X
500
X
Ag -110m
X
X
X
Sb -124
X
X
X
Sb -122
X
X
X
Ar -41
3.0E+08
1.0E+09
1
Kr-85m
4.6E+08
1.3E+10
2.3
Kr-85
1.9E+07
5.2E+08
1
Kr-87
8.3E+08
2.3E+10
2.,3
Kr-88
1.2E+09
3.2E+10
2.3
Xe -133m
2.5E+08
3.9E+09
2.3
Xe -133
9,5E+09
1.5E+11
1.9
Xe -135
1,5E+09
2.5E+10
1.4
Xe -138
2,5E+09
4.1E+10
2.9
Sr-89
3.0E+05
4.9E+06
1
Sr-90
1.9E+03
3.0E+04
1
I -131
2.3E+09
3.7E+10
23
I -132
2.2E+09
3.4E+10
12
I -133
2.4E+09
3.7E+10
7.6
I -134
1,5E+09
2.4E+10
14
I -135
1.4E+09
2.3E+10
7.1
Cs-134
9.6E+08
7.7E+09
24
Cs-136
X
X
X
Cs-137
8.0E+08
6.4E+09
20
Cs-138
2,5E+09
3.7E+10
4.1E+10
3.7E+10
2.9
-
H-3
X: Values to be included later using most recent operational feedback.
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FUNDAMENTAL SAFETY OVERVIEW
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STANDARD
TAB 3:
CORROSION PRODUCTS RADIOACTIVE DEPOSITS IN THE PRIMARY
LOOPS (RCP 1, 2, 3, 4)
HOT/COLD LEG
STEAM GENERATORS
(Bq/m2)
(Bq/m2)
Mn-54
2.5E+08 – 4.0E+08
6.5E+07 – 1.3E+08
Co-58
3.0E+09 – 5.2E+09
2.5E+08 – 2.6E+09
Fe-59
7.0E+07 – 2.0E+08
5.0E+07 – 1.2E+08
Co-60
5.0E+08 – 9.8E+08
2.5E+08 – 5.0E+08
Cr-51
X
X
Ag-110m
X
X
Sb-124
X
X
Sb-122
X
X
NUCLIDE1
X: Values to be included later using most recent operational feedback.
Note 1:
In order to avoid exceptional pollution events (due to Ag-110m and Sb-124), recorded in
operational feedback from existing units, the design of the primary components in contact with
the primary cooling water aims to avoid use of metals containing the polluting elements.
With this objective in mind, the main improvements being pursued are:
-
a reduction in the use of helicoflex seals in favour of graphite seals,
-
increased use of antimony free bearings and stops with submerged rotor,
-
use of antimony free mechanical fittings on affected pumps.