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Gamma and dose fields around the primary circuit of a helium-cooled fusion power plant
Fusion
Eno!nn o Fusion Engineering and Design 36 (1997) 191-201 ELSEVIER
and Desugn
Gamma and dose fields around the primary circuit of a helium-cooled fusion power plant C.B.A. Forty UKAEA Fusion, Culham, Abingdon, Oxfordshire OXI4 3DB, UK
Received 20 December 1996; accepted 8 January 1997
Abstract
The Monte Carlo code MCNP-4 has been used to determine the photon flux spectra around two important regions of the primary circuit of the conceptual SEAFP Plant Model 1 (PM-1) fusion power plant. These fluxes have been converted to absorbed dose rates in human tissue by using dose rate conversion factors. Dose rates around a 'hot-leg' pipe were found to be low and not significantly above naturally occurring background levels. The dose fields surrounding a steam generator were found to be higher than the 'hot-leg' pipe, but still well within acceptable limits. Dose rates inside the steam generator shell are considerably higher and would probably require remote and robotic method for inspection and maintenance tasks. The annual occupational radiation exposure to all plant workers was estimated to be approximately 170 man-mSv year-1, which is similar to the levels experienced with modern PWRs, after decades of optimisation. © 1997 Elsevier Science S.A.
1. Introduction
The three year European Safety and Environmental Assessment o f Fusion Power project (SEAFP), has been completed and reported [1]. Its main aim was to provide a convincing demonstration of the good inherent safety characteristics of electricity generation by fusion power. The study investigated two conceptual fusion power plant options employing either long-term or nearterm technologies. The results from the study were, however, deemed generic enough to cover a broad range of other fusion plant concepts. The long-term 'advanced concept' option, referred to as the SEAFP Plant Model 1 (PM-1), was envisaged as having a helium-cooled first wall, blanket and divertor, a Li20 tritium generating blanket
material and a vanadium alloy for structural components and cooling pipes. Although much of the study was focused on the public impact o f fusion power, a key element in the overall safety assessment was the likely occupational radiation exposure (ORE) to plant personnel from normal operation. ORE was found to arise from three main sources [2]: 1. internal fl exposure from inhaled tritium; 2. external ~-field exposure from activated structure and ancillary plant; 3. external ~,-field exposure from the primary cooling loops. Of these, the most significant source of exposure is that due to the activation of the primary circuit cooling loops. F o r the PM-1, neutron interaction with blanket cooling pipes results in radioactive
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C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201
ions being ejected into the coolant by the mechanism of daughter nucleus recoil sputtering [3]. This process occurs when a target nucleus in the near-surface of a cooling pipe captures a neutron and forms a short-lived compound nucleus. This state is then assumed to decay isotropically by emission of a particle, simultaneously providing the remaining radioactive daughter nucleus with sufficient recoil momentum to propel it through the pipe lattice. If the trajectory of the recoil nucleus crosses an inner pipe surface, it is lost into the coolant. Once entrained in the helium coolant, the nucleus rapidly attains the correct electronic charge configuration, becoming a radionuclide atom. It may then become transported to other regions of the cooling loop where deposition can occur. The details of these mass transfer processes are discussed in an accompanying paper [4]. Consequently, the entire cooling loop becomes activated over time. Those contaminated parts of each circuit outside the biological shield, notably the main 'hot-leg' pipe, steam generators and helium circulators, may then contribute dominantly to the ORE. A simplistic ORE assessment was made for the SEAFP PM-1 [2], using an idealised pipe geometry. The 7-radiation field associated with a monoenergetic internally distributed photon source was calculated at a point outside the pipe with the RANKERN [5] computer code. Based on this result, the annual ORE was estimated. RANKERN is a random sampling point kernel code, ideal for relatively simple problems involving known build-up factors [6]. However, some important features missing from point kernel methods are the inadequacies in the following areas. 1. The inability to model very geometrically complex systems, such as some heat exchangers, pipe manifolds and pumps. 2. The assumption that scattering does not degrade the photon energy and therefore attenuate the spectrum. 3. Systems involving multi-layer attenuation shielding, where build-up methods fail. The adoption of a point kernel method in SEAFP was justified in that, whilst being a simplified
approach, nothing better was available at the time and it nevertheless gives a fairly robust 'first stab' estimate of likely y-fields near some features of interest. The purpose of the present work, is therefore to repeat the calculations, taking account of the additional features enumerated above. To do this, we thus require the use of a more sophisticated technique. A review of various gamma transport methods and a set of systematic benchmark calculations of increasing complexity [7], was performed with this aim in mind. Of the methods reviewed, three dimensional Monte Carlo was found to be the most flexible and accurate technique available. Further calculations have therefore been made using the MCNP-4 [8] Monte Carlo code. Two regions of the cooling circuit have been modelled to assess the 7-dose fields and the resulting absorbed dose rate in human tissue. The first and simplest problem, repeats the SEAFP analysis in slightly more detail. The object for study is a straight, large diameter, thick-walled pipe transporting the coolant to the heat exchanger. The second, more complex geometry problem examines the y- and dose-fields at several locations around a steam generator. An estimate for the annual ORE is made based on these results.
2. SEAFP PM-1 cooling circuits Design information regarding the machine, blanket and cooling circuit designs are presented in great detail elsewhere [9-11]. In the following description, a simplified view is presented. The SEAFP PM-1 design is modular. It consists of 16 blanket segments which combine to give the toroidal geometry of the tokamak. These are surrounded by poloidal and toroidal field coils and the cryostat. Each segment consists of two inboard and three outboard blanket canisters comprising first wall, reinforcing struts and back plate all fabricated from a vanadium alloy. Every blanket canister contains 35 compartments arranged along the poloidal length, each of which contains 20 coolant tubes also made from the vanadium alloy. The compartments are filled with LizO
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191 201
spheres for tritium and heat generation. A series of manifolds connects together the pipes in the compartments to common canister lines and finally from two segments into a single cooling loop containing a heat exchanger and helium circulator. Plasma purity and exhaust control is achieved using divertors situated at the bottom of the machine. The helium cooling circuits for the SEAFP PM-1 consist of eight primary circuit loops, which cool the first wall/blanket modules, and two independent divertor loops. Fig. I shows a schematic of a single primary circuit loop. For any one primary circuit loop, the main pipe work and cooling plant items of interest in this study are the 'hot-leg' pipe which runs from the blanket outlet manifolds to the steam generator and the steam generator itself. These two areas are among the most contaminated circuit plant regions where man access is permitted and thus potentially the most significant risk areas for ORE. Each 'hotleg' is a long, large diameter low-carbon steel pipe with a relatively thick pipe wall to withstand the
I •
1
'hotleg'
F __
i i ) I
coolant
steam generator
t
h©lium circulator
J__, blaiaket p i p e w o r k
193
high pressures and temperatures of the helium coolant it contains. In the steam generator, helium flows downwards over inconel tubes before exiting at the bottom. It then passes through the helium circulator and is returned to the blanket modules.
3. Monte Carlo simulation There are a variety of techniques available for determining photon fluxes at a position in space. Three of the most commonly employed, are: solution by analytical formulae; numerical point kernel summation; and Monte Carlo simulation [7]. For problems involving complex geometry, variable energy source spectrum and multi-layer shielding, the most reliable and accurate method is 3-D Monte Carlo simulation. The code adopted for this study is the general purpose, continuousenergy, generalised-geometry, time-dependent coupled neutron/photon Monte Carlo transport code, MCNP-4 [8]. It solves neutral particle transport problems and may be used in neutron only mode, photon only mode or combined mode. The code is used by submitting an input file and receiving the output. The input file contains all details of the geometry of the system described in terms of volume cells bounded by surfaces. By combining simple geometries together, it is possible to describe arbitrarily complex systems. The input file also specifies the relevant material data, the location and characteristics of the neutron or photon source, the output requirements ('tallies'), the use of any variance reduction techniques necessary for more efficient particle tracking and finally the desired method for terminating the calculation.
i I )
4. MCNP input geometry 4.1. 'Hot-leg" pipe geometry
'cold leg'
Fig. 1. Schematicrepresentation of a primary coolingloop in the SEAFP PM-1.
Each 'hot-leg' is a straight pipe of length l = 50 m, with inside diameter of d~= 1.25 m and wall thickness of t = 19 cm [11]. The low carbon steel (Fe-0.1% C) pipe is assumed to be located in air
194
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201
/
\\
!
",
P1
\ \\
++
--'-, \'x,]
/\\~c2
\
",1, " , v
./ ]
[x,,~
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The uniformly distributed internal radioactive pipe wall deposit was simulated by defining a gamma emitting source (described in Section 5.1) located on the inner cylinder surface, C1. The photons were emitted at all solid angles from this surface at positions chosen at random. Each photon was tracked before and after an interaction with the various attenuation media, until its energy either fell below 1 keV or it was lost to the 'outside world', Secondary recoil electrons and their resulting bremsstrahlung photon emissions were also followed, but their contribution to the calculated flux at the detector was found to be negligible. 4.2. Steam generator geometry
/
Fig. 2. Geometric construction used in the M C N P - 4 Monte Carlo 'hot-leg' calculations.
at STP and is filled with helium coolant at Tpc = 833 K and at Pr+ = 9 MPa pressure. This geometry is simulated by defining the volume cells enclosed by three concentric cylindrical surfaces and two plane surfaces as shown on Fig. 2. The coolant cell was defined by the volume inside the innermost cylindrical surface, C1 bounded at both ends by plane surfaces PI and P2. This cell was assigned the appropriate elemental and density properties for pressurised helium. The volume between the two innermost cylinders, C1, C2 and the end planes, P1 and P2, was assigned the material properties for low carbon steel and defines the pipe wall. The air space around the pipe was defined by the volume bounded by the outermost cylindrical surface, C3, the second cylindrical surface, C2, and end planes, P1 and P2. The surface of the outermost cylinder marks the interface at which particles are qost' from the simulation. The detector position (i.e. the spatial location where the photon flux is calculated) is situated halfway along the pipe length and comprises an annular ring 3.63 m in diameter - 1 m from the pipe surface. A detector with ring geometry was chosen in order to improve counting statistics.
Each steam generator consists of a low carbon steel shell approximately 20 m in height, through which the helium flows downwards over helically wound inconeI alloy tube banks carrying pressurised water/steam. A central tube made from low carbon steel and passing through the tube bundle windings allows a variable fraction of helium to bypass the heat transfer surfaces to cope with fluctuations in steam demand. A schematic diagram of the geometric approximation to the steam generator is shown in Fig. 3. The helically wound SG tube banks were modelled by using a simplifying approximation for the MCNP-4 geometric construction. For the purposes of the calculation, a unit cell was defined as having a repeating structure that is then implemented in the model. The unit cell was taken to consist of an axial slice cut through the steam generator at a position where the steam tubes are located. Rather than model the tubes in their helical array, a simplified geometry was employed, whereby they are represented by closed loop toroidal pipes. Each unit cell, thus consisted of a staggered array of 21 tori of circular cross section, as shown in Fig. 3. All tori had a minor inner radius ri = 10 mm, minor outer radius ro = 12.5 mm and varying major radius. The inner steam generator bypass tube and outer shell were also built into the unit cell model. This unit cell was repeated over the tube bank height of h t = 10.5 m,
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201
to thus represent the heat transfer surfaces. The remaining upper and lower portions of the steam generator shell and bypass tube were modelled similarly, but without the steam tubes. Finally, the steam generator ends were modelled using hemispherical domes, with inlet and outlet pipes. Each toroidal steam tube was assumed to be made from an inconel alloy (20%Fe-20°/oCr 60%Ni) of density Pinc --- 8000 kg m - 3 and is filled with pressurised water/steam at T ~ = 533 K, Ps¢ = 13 MPa and p~ = 800 kg m -3. The pressurised water/steam constitutes the secondary coolant and is used to drive turbines. The inner helium bypass tube was defined as being made from low carbon steel and having a radius of R b = 0 . 5 2 m, wall thickness of t b = 0.13 m and height of h b = 19.2 m. The steam generator outer shell was also assumed to be made from low carbon steel with dimensions: inner radius of Ro = 1.47 m, shell thickness of ts = 0.13 m, total SG height of hs = 22.2 m with a domed top and bottom. Pressurised helium at Tp~ = 833 K, Pp~ = SGinlet
repeat unit cell
shell
tube bank height
195
9 MPa and density of Pr~--5.26 kg m -3 was assumed to fill the remaining space inside the shell. The outer airspace was again at STP, with the ring detector placed at several locations inside and outside of the shell.
5. The source term and surface emission source density
Daughter nucleus recoil sputtering from the vanadium alloy blanket cooling pipes was identified as the principle source of active contamination of the PM-1 primary circuits [3]. Transport and deposition of this activity was calculated using the H E T R A N code [4]. Although the bulk coolant also contains active species, it is the deposited radionuclides that constitute the dominant source term. These surface deposited nuclide concentrations are found to be made up from many nuclides each of which possesses its own characteristic X-ray and y-ray emissions. While the data on photon yield, photon energy, the existence of possible isomeric states, etc. are available in the literature [12-14], the most convenient source of data is the decay library of the European Activation File (EAF-4) [15]. This contains all the essential nuclear data for 1602 active species starting from the light nuclide 3H to the heavy actinide 248Cm inclusive. To further ease the handling of these data, the neutron activation code FISPACT-4.1 [16] was used to process the photon emission data provided by EAF-4 into 24 discrete energy groups from which the rate of photon emission per disintegration, Ry, may be obtained for each nuclide. Deposited activity concentrations are therefore converted to surface emission flux spectra by multiplying with R~. 5.1. The "hot-leg" source term and surface emission source density
SG outlet
Fig. 3. Schematic representation of the steam generator (SG) geometry used in the MCNP-4 Monte Carlo calculations.
The variation in concentration of deposited activity from the inlet end of the 'hot-leg' pipe to the outlet end, 50 m downstream is less than 5%. The whole inner surface of the pipe can, therefore, be assumed to have a uniform distribution of de-
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191 201
196
Table 1 Surface concentrations of deposited nuclides in the 'hot leg' pipe and steam generator Nuclide
13N 16N 24Na 46Sc 47Sc 48Sc 51Ti 181W Total
Average deposited concentration (Bq m -2) 'Hot-leg'
Steam generator
2.05 3.92 1.61 1.51 1.55 6.44 2.78 2.86 5.63
1.27 x 3.39 x 1.76 x 1.70 x 1.99 x 8.70 x 4.20 x 3.36 x 6.70 x
x x x x x x x x x
105 102 108 108 108 107 106 107 108
105 102 l0 s 108 108 107 106 107 108
posited activation products. H E T R A N calculations [4] indicate that this total concentration, which we define as the source term, is 5.63 x 108 Bq m - 2, made up from the partial concentrations of the important nuclides (with penetrating photon emissions) listed in Table 1. The photon emissions associated with each deposited nuclide is obtained using R~. Photon emissions are placed into the 24 energy group structures as shown in Table 2 (note, that only some of the energy groups have any significant contribution). The sum of all energy group contributions is defined as the surface emission source density (SESD). 5.2. The steam generator source term and surface emission source density The steam generator is immediately downstream from the 'hot-leg' pipe and therefore has a similar nuclide deposition concentration per unit area at the inlet. There is however, an increased deposition, relative to the 'hot-leg', of non-volatile species (i.e. all nuclides except the gaseous 13N and 16N) within the steam generator leading to an axial deposition concentration through the shell and tube banks. For simplicity, the deposition concentrations at the mid-height position have been used in the calculations. Because of the large surface area of pipe-work available for deposition, the source term is in fact, considerably higher per unit length of steam generator compared with the 'hot-leg'. The large number of inconel tube sur-
faces plus the shell and helium bypass surfaces result in a source term approximately ten times greater than was the case for the simple 'hot-leg' pipe surface. This increased source term only occurs over the 10.5 m of tube bank height where the heat transfer surfaces are concentrated. The upper and lower shell portions have only the deposits on the shell and bypass walls and the source term at these locations is therefore considerably lower. The resulting unit area deposition concentrations are listed in Table 1, for the most significant photon emitters. Table 2 gives the conversion from source term to SESD. It is noteworthy that the deposition concentrations and therefore SESD values are generally higher in the steam generator compared with the 'hot-leg' pipe. However, the reduced SESD value in the steam generator relative to the 'hot-leg' pipe at the gamma energy interval 0.4-0.6 MeV is found to be due to the lower flux of annihilation photons arising from 13N decay.
6. Photon flux calculation and results
6.1. The 7-field around the "hot-leg'pipe The total photon emitting source strength is defined as the SESD (9.71 × 108 7 m - 2 s - l ) times the total internal surface area of the 'hot-leg' pipe given by, A = ndil, thus yielding a total source strength of 1.91 x 1011 7 s-1, assumed to be distributed uniformly. This was partitioned into the partial contributions made from photons in each of the 24 energy groups. As can be seen from Table 2, only about half of the energy groups provide any contribution. The calculation was run for 15 million particle histories, by which time the counting statistics had become acceptably accurate. The flux spectrum at the ring detector was discretised into the same 24 group structure as the source. The total photon flux (sum of all energies) was calculated to be 6.13 × 105 7 m - 2 s-~ (-+ 4.8%). The photon spectrum at the detector is compared with the 'raw' surface emission source density on the pipe inner surface in Fig. 4. The missing energy groups in the source spectrum are
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191 201
197
Table 2 E n e r g y dependent photon surface emission source density (SESD) emitted at the inner 'hot leg' and steam generator surfaces
Gamma group no.
Energy interval (MeV)
1
S E S D (7 m
0.00-0.01 0.05-0.1 0.10 0.20 0.30-0.40 0.40-0.60 0.60-0.80 0,80-1.00 1.00-1.22 1.22-1.44 1.66-2.00 2.50-3.00 3.00-4.00
4 5 7 8 9 10 11 12 14 16 17
Total photon density
clearly evident in Fig. 4, particularly for the 0.20.3 and 1.66-1.83 MeV group ranges. Major peaks are observed in the group ranges defined at 0.1-0.2 MeV (due mainly to 47Sc); 0.8-1.0 and 1.0-1.22 MeV (due to 46Sc and 48Sc) and 2.5-3.0 MeV (due to 24Na). The detector flux spectrum is distinctly different from the 'raw' spectrum, being less peaky and shifted to low energy groups as expected. The smoothing of the detector spectrum 1012
........
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........
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SESD flux
detector
101o 10 9
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107 x
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1
It
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1I t7
i
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1 photon
I
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I
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energy
10
(MeV)
Fig. 4. Photon spectra associated with the 'hot-leg' cooling
pipe.
2
S
I)
'Hot-leg'
Steam generator
1.49 X 1 0 7 1.19 x 108 2,40 x 106 4.18 x 105 2.98 x 104 2.20 x 108 2.12 x 108 2.29 x 108 5.16 x 103 1.61 x 108 9.25 × 104 9.71 x 108
1.39 1.75 1.53 3.63 2.59 4,50 2,63 2,54 2.67 4.07 1.76 1.01 1.15
× lO 7 x 107
x x x x x x x × x x ×
108 10 '5 105 lO4 108 10 s t08 104 10 s 105 109
is mainly due to the multiple Compton scattering events at the intermediate energies and the production of bremsstrahlung photons at lower energies. 6.2. The ?,-field around the steam generator The MCNP-4 input file was modified to take into account all the geometry and materials specifications described in Section 4.2. The photon emitting source strength was defined as the SESD (1.15 × 10 9 ?' m -2 S 1) times the total surface area of the steam generator. Calculating the internal area of the steam generator is not as straight forward as the earlier cylindrical pipe problem, mainly because of the spatial variation in available surfaces, i.e. the source distribution is variable in radial and axial co-ordinates. For simplicity, the steam generator was divided into an upper and lower shell section and a tube bank section each weighted by the available deposition areas available. The total source strength was finally calculated from the summation of each section and yielded 3.0 x 1012 y s - 1 . This was again subdivided into the partial contributions made from photons in the 24 energy groups. Flux tally results were requested in the 24 gamma group energy bins for a range of locations around and inside the steam generator as shown
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201
198
0.0
1.6
2.6
I 22.2 20.7 - -
"ql
14.5 - -
:q
D
• ql
C
F
\ 9.25 - -
\
strength becomes lower. The flux inside the shell, at the tube bank mid-height at position F, is observed to be many orders of magnitude higher than outside the shell thus giving an indication of the attenuation in the shell wall itself. Statistical variances are relatively poor compared with results obtained for earlier examples, but could be improved given longer computational running times and the use of more sophisticated variance reduction techniques. To examine the changes in spectral shape, the photon spectrum at the various locations A to F were compared with the 'raw' surface emission source density spectrum on the inner surfaces of the steam generator. With the exception of the magnitude of the group values, the spectral shapes were very similar to those shown for the 'hot-leg' in Fig. 4 and are therefore not repeated here.
7. Dose calculation and results 4.0 - -
1.5
-
0.0
-
-
:,l
A
-
Fig. 5. Schematic of the steam generator showing detector locations for flux tallies. All dimensions in metres and not drawn to scale,
in the schematic diagram in Fig. 5. Locations A to E are all external to the shell at various heights and at 1 m from the shell surface. Location F, is at the mid tube bank height at a position between the inner shell and outer tube bank windings. The MCNP-4 calculation was run allowing for the tracking of 150 thousand particle histories. The total fluxes at locations A to F together with their statistical standard deviations are given in Table 3. The photon flux is seen to peak opposite the mid-height position of the tube banks, location C, where the local source strength is highest due to the large surface areas present. It diminishes in the axial direction towards the top and the bottom of the steam generator as the source
Once the photon flux spectrum (discretised into 24 energy groups) has been calculated at the detector position, the dose rate in human tissue is computed. This is tissue (organ) dependent and therefore critically dependent on the orientation of the body with the photon flux. To simplify the analysis, energy dependent dose conversion factors, v, which provide a weighted average of tissue types and relative body orientations are used [8]. Over the 5 KeV- 14 MeV photon energy range the dose conversion factors are shown in Fig. 6. These data are fitted with a high order polynomial from which the values at mid-point energies in the 24 group structure can be determined. Table 3 Photon fluxes (with statistical standard deviation) around the steam generator Location
Total photon flux, ~b (7 m - 2 s - I )
~r ( + %)
A B C D E F
8.42 2.65 4.50 2.80 3.68 1.93
22.1 9.9 8.7 10.0 20.6 8.0
× × × × × ×
106 107 107 107 106 10 l°
C.B.A. Forty~Fusion Engineering and Design 36 (1997) 191-201 14x10-12
.... r
........
i
........
i
........
7.2. Dose rate around the steam generator
i
12x10-12
1 0 x l 0 .12
~7 8 x 1 0 -j2
6x10 J2
4x10-t2
8 2x10-12
0xl0 °
0.01
0.l
l
10
photon energy (MeV)
Fig. 6. Dose rate conversion factors, v, at discrete energies plus high order regression fit.
Having obtained the dose conversion factors in the same group structure as the detector fluxes, all that remains is to determine the dose rate in human tissue, D, by: dD
24
¥ = t~"1"=~ivi
199
(1)
where ~bi, is the group energy partial photon flux. 7.1. Dose rate near the 'hot-leg' pipe
The dose rate and photon flux are shown together in Fig. 7, at the mid-length position and at 1 m from the 'hot-leg' pipe surface. Both flux and dose spectra exhibit broadly similar features with peaks occurring at the same energies and the fall off at the top and bottom energy ranges. Most of the absorbed dose occurs in the energy range 0.2-3.0 MeV. The lower energy dose rate arises from the multiply scattered photons, while the majority of the dose rate at 2.5 MeV is due to uncollided 24Na photons (E~ = 2.754 MeV). The total dose rate is of order 2.2 laSv h - t , which is low enough to pose no important radiological risks. In fact, this figure is only about six to ten times larger than the average natural background radiation levels found at sea level (which gives an annual dose of around 2-3 mSv year- t, i.e. approximately 0.23-0.34 pSv h -1 [17]).
Photon fluxes in the 24 energy group structure obtained for locations A to F were converted to dose rates by multiplying by the appropriate dose conversion factors, v. The summation at each location for each of the 24 groups finally yielded the total dose rates. Since the shape of the flux spectra and dose rate spectra are similar for all locations, only two locations are used to illustrate the behaviour, namely locations C and F. Fig. 8 shows the results for the 'internal' location, referring to the tube bank mid-height position inside the shell wall and the 'external' location referring to the same axial height position but 1 m from the outer shell wall. The 'smoothing' of the flux after attenuation through the shell wall is evident on comparison of the internal and external locations. A relative reduction in the peak at 2.75 MeV is also seen. As before, dose rate curves show similar spectral shapes to their corresponding flux spectra. The maximum external dose rate (summed over all energies) occurs at location C and has a value 166 ~tSv h - l , whereas the dose rate inside the shell is s o m e 10 3 times higher. The external gamma field around the outside of the steam generator, whilst moderately high, would still permit individual worker access for over 100 h, and is 106
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10 5
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i
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i
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i
i
i
10
photon energy (MeV)
Fig. 7. Comparison of photon flux and dose rate spectra for the 'hot-leg'.
200
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201 1012
l0 t 'internal' - location F
10 LI
10 0
lOIO
10-t
109
10-2 10.3
,~
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10 3
lO-S •
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-
101
dose rate
i
10-9
photon flux I
t
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11111
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i
1
estimate gave the cooling circuit ORE at about 80 man-mSv year-~ [1] which is an equivalent result bearing in mind the differing technique used in the flux calculation. This value is also similar to the 200 man-mSv year-1 reported for modern Konvoi PWR plants [19] which are currently regarded as the lowest ORE plants anywhere in the world [2], after decades of optimisation. Thus on this simple estimate, the SEAFP PM-1 ORE is expected to be equivalent to modern fission power plant, already considered to be well inside acceptable limits.
,
, , ,11
10-10 10
photon energy (MeV)
Fig. 8. Comparisonof photon flux and dose rate spectra for two locationsaround the steam generator. therefore well within acceptable levels. Any maintenance work required inside the steam generator shell (for example, pressure tube repair or inspection) would however, require very strict working practices to minimise dose exposure. Tasks of this nature would, in all likelihood, be performed remotely or robotically.
8. Annual ORE estimation
To estimate the annual collective dose to all plant personnel, a detailed analysis of all radiation fields around the plant, derived air concentrations (DACs) in contaminated rooms and the occupation task breakdown would be required. In relation to anticipated ITER OREs, Sandri [18] has developed an effective methodology for this type of estimation. The detail required for an accurate assessment is, however, well beyond the scope of the present work and is not attempted here. However, a simple estimate may be made as follows. Noting again that the cooling circuit doses dominate ORE and an approximate correlation that a dose rate of 1 mSv h - 1 is equivalent to 1 man-Sv y e a r - l [2], the annual ORE due to the cooling plant systems (dominated by the steam generator contribution) is estimated to be approximately 170 man-mSv year- 1. The earlier SEAFP
9. Conclusions
Near term magnetic confinement fusion power plants will be based on the D-T fuel cycle. The resulting high energy neutrons will sputter radioactive ions from the blanket pipe walls into the coolant. In the absence of other source mechanisms, such as corrosion, dissolution and erosion, this sputtering term is the principle means by which the cooling circuits become activated. This is likely to be the case for the conceptual heliumcooled power plant, SEAFP PM-1. Active material is transported around the loop and deposited on internal pipe wall surfaces and other plant equipment such as steam generators and pumps/ circulators. Much of this equipment lies outside the biological shield where man access is allowed, resulting in potential occupational radiation exposures (OREs) to plant personnel. This ORE has been assessed as follows: (A) the photon fluxes in space have been determined near to critical plant regions; (B) based on these fluxes, dose rate have been calculated; and (C) the annual ORE has been estimated from the dose rates. Two regions of the coolant loop were modelled, namely: 1. a 'hot-leg' pipe; 2. a steam generator. Average surface deposit concentrations were calculated at these locations for a range of important radionuclides using the HETRAN code. These activity concentrations were next converted to a surface emission source density (SESD) through the FISPACT/EAF-4 code/library system. The
C.B.A. Forty~Fusion Engineering and Design 36 (1997) 191 201
7-fields surrounding the cooling plant equipment were determined using the MCNP-4 Monte Carlo code. Partial fluxes were calculated in a 24 energy group structure format, allowing for easy conversion to absorbed dose rate in human tissue using dose rate conversion factors. Dose rates around the 'hot-leg' pipe were calculated to be extremely low ( ~ 2.2 laSv h - l ) and not significantly above naturally occurring background levels. Because of the thinner shell wall and much greater surface areas available for deposition, the steam generator dose fields were found to be several orders of magnitude higher at around 166 ~tSv h - 1 . Whilst higher, these dose levels are well within acceptable levels and would permit worker residence times in the region of such a field for as long as 100 h. Dose rates inside the steam generator shell are considerably higher still and any tube bank inspection or maintenance regime would probably require remote or robotic methods. Based on these dose rates and using a crude scaling factor, the annual collective ORE from the whole plant, without any attempt at optimisation, is estimated to be approximately 170 man-roSy h - ' , which is equivalent to the levels achieved with the best PWRs operating today, after decades of practical optimisation.
Acknowledgements I should like to express my thanks to Dr P.J. Karditsas and Dr N.P. Taylor, both of UKAEA Fusion, for their valuable help. The funding for this work was provided by the UK Department of Trade and Industry and Euratom.
References [1] J. Raeder et al., Safety and Environmental Assessment of Fusion Power, European Commission Report No, EURFUBRU XII-217/95, 1995.
201
[2] C.G. Kinniburgh, Occupational radiation exposure, Report No SEAFP/R-A10B(94), 1994. [3] C.B.A. Forty, A nuclear sputtering model for fusion applications, UKAEA Fusion/Strategic Studies Note, SSN 96/7 (1996). [4] P.J. Karditsas, C.B.A. Forty, Modelling and simulation of activation product transport in helium cooling loops, Fus. Eng. Des. 36 (1997) 203-217. [5] P.C. Miller, RANKERN Users Guide, ANSWERS (RANKERN) 2, UKAEA Winfrith Report, 1993. [6] R.G. Jaeger (Ed.), Engineering Compendium on Radiation Shielding, Vol. 1, Springer-Verlag, Berlin, 1968. [7] C.B.A. Forty, Gamma radiation fields around the cooling circuits of a fusion power plant, UKAEA Fusion Report No., UKAEA FUS 325 (1996). [8] J.F. Briesmeister (Ed.), M C N P - - a general Monte Carlo code for neutron and photon transport, Los Alamos Report No, LA-73396-M, Rev. 2, 1986. [9] F. Engelmann et al., Specifications of reactor parameters and plasma operation, SEAFP Task M1 Final Report, R-M1-Rev.3, 1994. [10] W. Daenner and G. Simbolotti, Blanket design, SEAFP Task M6 Final Report, RM6-Rev.1, 1994. [11] J. Wagner, G. Sable and F.J. Mustoe, Cooling system design--helium option, SEAFP Task M8 Final Report, R-MS.2-Rev.1, 1994. [12] ICRP, Recommendations of the International Commission on Radiological Protection, Publication 38. Ann. ICRP 11-13, Pergamon Press, Oxford, 1983. [13] E. Browne, R.B. Firestone, Tables of Radioactive Isotopes, Wiley, New York, 1986. [14] D.C. Kocher, Radioactive decay tables, US Department of Energy Report, DOE/TIC-I1026, 1981. [15] J. Kopecky, D. Nierop, The European activation file EAF-4, Petten Report No ECN-C-95-072, 1995. [16] R.A. Forrest, J.-Ch. Sublet, FISPACT 4.1 user manual, UKAEA Report No., UKAEA FUS 287, 1995. [17] UNSCEAR, Sources and effects of ionizing radiation, United Nations Scientific Committee on the effects of atomic radiation, Report to the General Assembly, with scientific annexes, 1993. [18] S. Sandri, EU Task SEA2: personnel safety, ITER Task No S81TD01/D7105, ENEA Report No, ERG-FUS/ TECN S + E 13/95, 1995. [19] G.H. Hecht, Occupational exposure in LWR plants built by Siemens/KWU, International workshop on new developments in occupational dose control and ALARA implementation at nuclear power plants and similar facilities, NUREG/CP-0110, Brookhaven National Laboratory, September 1989.
Eno!nn o Fusion Engineering and Design 36 (1997) 191-201 ELSEVIER
and Desugn
Gamma and dose fields around the primary circuit of a helium-cooled fusion power plant C.B.A. Forty UKAEA Fusion, Culham, Abingdon, Oxfordshire OXI4 3DB, UK
Received 20 December 1996; accepted 8 January 1997
Abstract
The Monte Carlo code MCNP-4 has been used to determine the photon flux spectra around two important regions of the primary circuit of the conceptual SEAFP Plant Model 1 (PM-1) fusion power plant. These fluxes have been converted to absorbed dose rates in human tissue by using dose rate conversion factors. Dose rates around a 'hot-leg' pipe were found to be low and not significantly above naturally occurring background levels. The dose fields surrounding a steam generator were found to be higher than the 'hot-leg' pipe, but still well within acceptable limits. Dose rates inside the steam generator shell are considerably higher and would probably require remote and robotic method for inspection and maintenance tasks. The annual occupational radiation exposure to all plant workers was estimated to be approximately 170 man-mSv year-1, which is similar to the levels experienced with modern PWRs, after decades of optimisation. © 1997 Elsevier Science S.A.
1. Introduction
The three year European Safety and Environmental Assessment o f Fusion Power project (SEAFP), has been completed and reported [1]. Its main aim was to provide a convincing demonstration of the good inherent safety characteristics of electricity generation by fusion power. The study investigated two conceptual fusion power plant options employing either long-term or nearterm technologies. The results from the study were, however, deemed generic enough to cover a broad range of other fusion plant concepts. The long-term 'advanced concept' option, referred to as the SEAFP Plant Model 1 (PM-1), was envisaged as having a helium-cooled first wall, blanket and divertor, a Li20 tritium generating blanket
material and a vanadium alloy for structural components and cooling pipes. Although much of the study was focused on the public impact o f fusion power, a key element in the overall safety assessment was the likely occupational radiation exposure (ORE) to plant personnel from normal operation. ORE was found to arise from three main sources [2]: 1. internal fl exposure from inhaled tritium; 2. external ~-field exposure from activated structure and ancillary plant; 3. external ~,-field exposure from the primary cooling loops. Of these, the most significant source of exposure is that due to the activation of the primary circuit cooling loops. F o r the PM-1, neutron interaction with blanket cooling pipes results in radioactive
0920-3796/97/$17.00 © 1997 Elsevier Science S.A. All rights reserved. PIIS0920-3796(97)00022-7
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C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201
ions being ejected into the coolant by the mechanism of daughter nucleus recoil sputtering [3]. This process occurs when a target nucleus in the near-surface of a cooling pipe captures a neutron and forms a short-lived compound nucleus. This state is then assumed to decay isotropically by emission of a particle, simultaneously providing the remaining radioactive daughter nucleus with sufficient recoil momentum to propel it through the pipe lattice. If the trajectory of the recoil nucleus crosses an inner pipe surface, it is lost into the coolant. Once entrained in the helium coolant, the nucleus rapidly attains the correct electronic charge configuration, becoming a radionuclide atom. It may then become transported to other regions of the cooling loop where deposition can occur. The details of these mass transfer processes are discussed in an accompanying paper [4]. Consequently, the entire cooling loop becomes activated over time. Those contaminated parts of each circuit outside the biological shield, notably the main 'hot-leg' pipe, steam generators and helium circulators, may then contribute dominantly to the ORE. A simplistic ORE assessment was made for the SEAFP PM-1 [2], using an idealised pipe geometry. The 7-radiation field associated with a monoenergetic internally distributed photon source was calculated at a point outside the pipe with the RANKERN [5] computer code. Based on this result, the annual ORE was estimated. RANKERN is a random sampling point kernel code, ideal for relatively simple problems involving known build-up factors [6]. However, some important features missing from point kernel methods are the inadequacies in the following areas. 1. The inability to model very geometrically complex systems, such as some heat exchangers, pipe manifolds and pumps. 2. The assumption that scattering does not degrade the photon energy and therefore attenuate the spectrum. 3. Systems involving multi-layer attenuation shielding, where build-up methods fail. The adoption of a point kernel method in SEAFP was justified in that, whilst being a simplified
approach, nothing better was available at the time and it nevertheless gives a fairly robust 'first stab' estimate of likely y-fields near some features of interest. The purpose of the present work, is therefore to repeat the calculations, taking account of the additional features enumerated above. To do this, we thus require the use of a more sophisticated technique. A review of various gamma transport methods and a set of systematic benchmark calculations of increasing complexity [7], was performed with this aim in mind. Of the methods reviewed, three dimensional Monte Carlo was found to be the most flexible and accurate technique available. Further calculations have therefore been made using the MCNP-4 [8] Monte Carlo code. Two regions of the cooling circuit have been modelled to assess the 7-dose fields and the resulting absorbed dose rate in human tissue. The first and simplest problem, repeats the SEAFP analysis in slightly more detail. The object for study is a straight, large diameter, thick-walled pipe transporting the coolant to the heat exchanger. The second, more complex geometry problem examines the y- and dose-fields at several locations around a steam generator. An estimate for the annual ORE is made based on these results.
2. SEAFP PM-1 cooling circuits Design information regarding the machine, blanket and cooling circuit designs are presented in great detail elsewhere [9-11]. In the following description, a simplified view is presented. The SEAFP PM-1 design is modular. It consists of 16 blanket segments which combine to give the toroidal geometry of the tokamak. These are surrounded by poloidal and toroidal field coils and the cryostat. Each segment consists of two inboard and three outboard blanket canisters comprising first wall, reinforcing struts and back plate all fabricated from a vanadium alloy. Every blanket canister contains 35 compartments arranged along the poloidal length, each of which contains 20 coolant tubes also made from the vanadium alloy. The compartments are filled with LizO
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191 201
spheres for tritium and heat generation. A series of manifolds connects together the pipes in the compartments to common canister lines and finally from two segments into a single cooling loop containing a heat exchanger and helium circulator. Plasma purity and exhaust control is achieved using divertors situated at the bottom of the machine. The helium cooling circuits for the SEAFP PM-1 consist of eight primary circuit loops, which cool the first wall/blanket modules, and two independent divertor loops. Fig. I shows a schematic of a single primary circuit loop. For any one primary circuit loop, the main pipe work and cooling plant items of interest in this study are the 'hot-leg' pipe which runs from the blanket outlet manifolds to the steam generator and the steam generator itself. These two areas are among the most contaminated circuit plant regions where man access is permitted and thus potentially the most significant risk areas for ORE. Each 'hotleg' is a long, large diameter low-carbon steel pipe with a relatively thick pipe wall to withstand the
I •
1
'hotleg'
F __
i i ) I
coolant
steam generator
t
h©lium circulator
J__, blaiaket p i p e w o r k
193
high pressures and temperatures of the helium coolant it contains. In the steam generator, helium flows downwards over inconel tubes before exiting at the bottom. It then passes through the helium circulator and is returned to the blanket modules.
3. Monte Carlo simulation There are a variety of techniques available for determining photon fluxes at a position in space. Three of the most commonly employed, are: solution by analytical formulae; numerical point kernel summation; and Monte Carlo simulation [7]. For problems involving complex geometry, variable energy source spectrum and multi-layer shielding, the most reliable and accurate method is 3-D Monte Carlo simulation. The code adopted for this study is the general purpose, continuousenergy, generalised-geometry, time-dependent coupled neutron/photon Monte Carlo transport code, MCNP-4 [8]. It solves neutral particle transport problems and may be used in neutron only mode, photon only mode or combined mode. The code is used by submitting an input file and receiving the output. The input file contains all details of the geometry of the system described in terms of volume cells bounded by surfaces. By combining simple geometries together, it is possible to describe arbitrarily complex systems. The input file also specifies the relevant material data, the location and characteristics of the neutron or photon source, the output requirements ('tallies'), the use of any variance reduction techniques necessary for more efficient particle tracking and finally the desired method for terminating the calculation.
i I )
4. MCNP input geometry 4.1. 'Hot-leg" pipe geometry
'cold leg'
Fig. 1. Schematicrepresentation of a primary coolingloop in the SEAFP PM-1.
Each 'hot-leg' is a straight pipe of length l = 50 m, with inside diameter of d~= 1.25 m and wall thickness of t = 19 cm [11]. The low carbon steel (Fe-0.1% C) pipe is assumed to be located in air
194
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201
/
\\
!
",
P1
\ \\
++
--'-, \'x,]
/\\~c2
\
",1, " , v
./ ]
[x,,~
A
"\ ~
The uniformly distributed internal radioactive pipe wall deposit was simulated by defining a gamma emitting source (described in Section 5.1) located on the inner cylinder surface, C1. The photons were emitted at all solid angles from this surface at positions chosen at random. Each photon was tracked before and after an interaction with the various attenuation media, until its energy either fell below 1 keV or it was lost to the 'outside world', Secondary recoil electrons and their resulting bremsstrahlung photon emissions were also followed, but their contribution to the calculated flux at the detector was found to be negligible. 4.2. Steam generator geometry
/
Fig. 2. Geometric construction used in the M C N P - 4 Monte Carlo 'hot-leg' calculations.
at STP and is filled with helium coolant at Tpc = 833 K and at Pr+ = 9 MPa pressure. This geometry is simulated by defining the volume cells enclosed by three concentric cylindrical surfaces and two plane surfaces as shown on Fig. 2. The coolant cell was defined by the volume inside the innermost cylindrical surface, C1 bounded at both ends by plane surfaces PI and P2. This cell was assigned the appropriate elemental and density properties for pressurised helium. The volume between the two innermost cylinders, C1, C2 and the end planes, P1 and P2, was assigned the material properties for low carbon steel and defines the pipe wall. The air space around the pipe was defined by the volume bounded by the outermost cylindrical surface, C3, the second cylindrical surface, C2, and end planes, P1 and P2. The surface of the outermost cylinder marks the interface at which particles are qost' from the simulation. The detector position (i.e. the spatial location where the photon flux is calculated) is situated halfway along the pipe length and comprises an annular ring 3.63 m in diameter - 1 m from the pipe surface. A detector with ring geometry was chosen in order to improve counting statistics.
Each steam generator consists of a low carbon steel shell approximately 20 m in height, through which the helium flows downwards over helically wound inconeI alloy tube banks carrying pressurised water/steam. A central tube made from low carbon steel and passing through the tube bundle windings allows a variable fraction of helium to bypass the heat transfer surfaces to cope with fluctuations in steam demand. A schematic diagram of the geometric approximation to the steam generator is shown in Fig. 3. The helically wound SG tube banks were modelled by using a simplifying approximation for the MCNP-4 geometric construction. For the purposes of the calculation, a unit cell was defined as having a repeating structure that is then implemented in the model. The unit cell was taken to consist of an axial slice cut through the steam generator at a position where the steam tubes are located. Rather than model the tubes in their helical array, a simplified geometry was employed, whereby they are represented by closed loop toroidal pipes. Each unit cell, thus consisted of a staggered array of 21 tori of circular cross section, as shown in Fig. 3. All tori had a minor inner radius ri = 10 mm, minor outer radius ro = 12.5 mm and varying major radius. The inner steam generator bypass tube and outer shell were also built into the unit cell model. This unit cell was repeated over the tube bank height of h t = 10.5 m,
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201
to thus represent the heat transfer surfaces. The remaining upper and lower portions of the steam generator shell and bypass tube were modelled similarly, but without the steam tubes. Finally, the steam generator ends were modelled using hemispherical domes, with inlet and outlet pipes. Each toroidal steam tube was assumed to be made from an inconel alloy (20%Fe-20°/oCr 60%Ni) of density Pinc --- 8000 kg m - 3 and is filled with pressurised water/steam at T ~ = 533 K, Ps¢ = 13 MPa and p~ = 800 kg m -3. The pressurised water/steam constitutes the secondary coolant and is used to drive turbines. The inner helium bypass tube was defined as being made from low carbon steel and having a radius of R b = 0 . 5 2 m, wall thickness of t b = 0.13 m and height of h b = 19.2 m. The steam generator outer shell was also assumed to be made from low carbon steel with dimensions: inner radius of Ro = 1.47 m, shell thickness of ts = 0.13 m, total SG height of hs = 22.2 m with a domed top and bottom. Pressurised helium at Tp~ = 833 K, Pp~ = SGinlet
repeat unit cell
shell
tube bank height
195
9 MPa and density of Pr~--5.26 kg m -3 was assumed to fill the remaining space inside the shell. The outer airspace was again at STP, with the ring detector placed at several locations inside and outside of the shell.
5. The source term and surface emission source density
Daughter nucleus recoil sputtering from the vanadium alloy blanket cooling pipes was identified as the principle source of active contamination of the PM-1 primary circuits [3]. Transport and deposition of this activity was calculated using the H E T R A N code [4]. Although the bulk coolant also contains active species, it is the deposited radionuclides that constitute the dominant source term. These surface deposited nuclide concentrations are found to be made up from many nuclides each of which possesses its own characteristic X-ray and y-ray emissions. While the data on photon yield, photon energy, the existence of possible isomeric states, etc. are available in the literature [12-14], the most convenient source of data is the decay library of the European Activation File (EAF-4) [15]. This contains all the essential nuclear data for 1602 active species starting from the light nuclide 3H to the heavy actinide 248Cm inclusive. To further ease the handling of these data, the neutron activation code FISPACT-4.1 [16] was used to process the photon emission data provided by EAF-4 into 24 discrete energy groups from which the rate of photon emission per disintegration, Ry, may be obtained for each nuclide. Deposited activity concentrations are therefore converted to surface emission flux spectra by multiplying with R~. 5.1. The "hot-leg" source term and surface emission source density
SG outlet
Fig. 3. Schematic representation of the steam generator (SG) geometry used in the MCNP-4 Monte Carlo calculations.
The variation in concentration of deposited activity from the inlet end of the 'hot-leg' pipe to the outlet end, 50 m downstream is less than 5%. The whole inner surface of the pipe can, therefore, be assumed to have a uniform distribution of de-
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191 201
196
Table 1 Surface concentrations of deposited nuclides in the 'hot leg' pipe and steam generator Nuclide
13N 16N 24Na 46Sc 47Sc 48Sc 51Ti 181W Total
Average deposited concentration (Bq m -2) 'Hot-leg'
Steam generator
2.05 3.92 1.61 1.51 1.55 6.44 2.78 2.86 5.63
1.27 x 3.39 x 1.76 x 1.70 x 1.99 x 8.70 x 4.20 x 3.36 x 6.70 x
x x x x x x x x x
105 102 108 108 108 107 106 107 108
105 102 l0 s 108 108 107 106 107 108
posited activation products. H E T R A N calculations [4] indicate that this total concentration, which we define as the source term, is 5.63 x 108 Bq m - 2, made up from the partial concentrations of the important nuclides (with penetrating photon emissions) listed in Table 1. The photon emissions associated with each deposited nuclide is obtained using R~. Photon emissions are placed into the 24 energy group structures as shown in Table 2 (note, that only some of the energy groups have any significant contribution). The sum of all energy group contributions is defined as the surface emission source density (SESD). 5.2. The steam generator source term and surface emission source density The steam generator is immediately downstream from the 'hot-leg' pipe and therefore has a similar nuclide deposition concentration per unit area at the inlet. There is however, an increased deposition, relative to the 'hot-leg', of non-volatile species (i.e. all nuclides except the gaseous 13N and 16N) within the steam generator leading to an axial deposition concentration through the shell and tube banks. For simplicity, the deposition concentrations at the mid-height position have been used in the calculations. Because of the large surface area of pipe-work available for deposition, the source term is in fact, considerably higher per unit length of steam generator compared with the 'hot-leg'. The large number of inconel tube sur-
faces plus the shell and helium bypass surfaces result in a source term approximately ten times greater than was the case for the simple 'hot-leg' pipe surface. This increased source term only occurs over the 10.5 m of tube bank height where the heat transfer surfaces are concentrated. The upper and lower shell portions have only the deposits on the shell and bypass walls and the source term at these locations is therefore considerably lower. The resulting unit area deposition concentrations are listed in Table 1, for the most significant photon emitters. Table 2 gives the conversion from source term to SESD. It is noteworthy that the deposition concentrations and therefore SESD values are generally higher in the steam generator compared with the 'hot-leg' pipe. However, the reduced SESD value in the steam generator relative to the 'hot-leg' pipe at the gamma energy interval 0.4-0.6 MeV is found to be due to the lower flux of annihilation photons arising from 13N decay.
6. Photon flux calculation and results
6.1. The 7-field around the "hot-leg'pipe The total photon emitting source strength is defined as the SESD (9.71 × 108 7 m - 2 s - l ) times the total internal surface area of the 'hot-leg' pipe given by, A = ndil, thus yielding a total source strength of 1.91 x 1011 7 s-1, assumed to be distributed uniformly. This was partitioned into the partial contributions made from photons in each of the 24 energy groups. As can be seen from Table 2, only about half of the energy groups provide any contribution. The calculation was run for 15 million particle histories, by which time the counting statistics had become acceptably accurate. The flux spectrum at the ring detector was discretised into the same 24 group structure as the source. The total photon flux (sum of all energies) was calculated to be 6.13 × 105 7 m - 2 s-~ (-+ 4.8%). The photon spectrum at the detector is compared with the 'raw' surface emission source density on the pipe inner surface in Fig. 4. The missing energy groups in the source spectrum are
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191 201
197
Table 2 E n e r g y dependent photon surface emission source density (SESD) emitted at the inner 'hot leg' and steam generator surfaces
Gamma group no.
Energy interval (MeV)
1
S E S D (7 m
0.00-0.01 0.05-0.1 0.10 0.20 0.30-0.40 0.40-0.60 0.60-0.80 0,80-1.00 1.00-1.22 1.22-1.44 1.66-2.00 2.50-3.00 3.00-4.00
4 5 7 8 9 10 11 12 14 16 17
Total photon density
clearly evident in Fig. 4, particularly for the 0.20.3 and 1.66-1.83 MeV group ranges. Major peaks are observed in the group ranges defined at 0.1-0.2 MeV (due mainly to 47Sc); 0.8-1.0 and 1.0-1.22 MeV (due to 46Sc and 48Sc) and 2.5-3.0 MeV (due to 24Na). The detector flux spectrum is distinctly different from the 'raw' spectrum, being less peaky and shifted to low energy groups as expected. The smoothing of the detector spectrum 1012
........
10 tl
I
.....
........
I
........
I
SESD flux
detector
101o 10 9
i--I I I
I0 8
107 x
I I I
106
~=
105
,=,
104
~1
I
....
1i
I
I I
I I I I I I~ j
1
i
I
10 3
I I
.... 10 2
I I I I
10 I
10o . . . . 0.01
....
t
1
It
I
I-- I
1I t7
i
0.1
1 photon
I
' 1I 'II I
I
I 1 -
energy
10
(MeV)
Fig. 4. Photon spectra associated with the 'hot-leg' cooling
pipe.
2
S
I)
'Hot-leg'
Steam generator
1.49 X 1 0 7 1.19 x 108 2,40 x 106 4.18 x 105 2.98 x 104 2.20 x 108 2.12 x 108 2.29 x 108 5.16 x 103 1.61 x 108 9.25 × 104 9.71 x 108
1.39 1.75 1.53 3.63 2.59 4,50 2,63 2,54 2.67 4.07 1.76 1.01 1.15
× lO 7 x 107
x x x x x x x × x x ×
108 10 '5 105 lO4 108 10 s t08 104 10 s 105 109
is mainly due to the multiple Compton scattering events at the intermediate energies and the production of bremsstrahlung photons at lower energies. 6.2. The ?,-field around the steam generator The MCNP-4 input file was modified to take into account all the geometry and materials specifications described in Section 4.2. The photon emitting source strength was defined as the SESD (1.15 × 10 9 ?' m -2 S 1) times the total surface area of the steam generator. Calculating the internal area of the steam generator is not as straight forward as the earlier cylindrical pipe problem, mainly because of the spatial variation in available surfaces, i.e. the source distribution is variable in radial and axial co-ordinates. For simplicity, the steam generator was divided into an upper and lower shell section and a tube bank section each weighted by the available deposition areas available. The total source strength was finally calculated from the summation of each section and yielded 3.0 x 1012 y s - 1 . This was again subdivided into the partial contributions made from photons in the 24 energy groups. Flux tally results were requested in the 24 gamma group energy bins for a range of locations around and inside the steam generator as shown
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201
198
0.0
1.6
2.6
I 22.2 20.7 - -
"ql
14.5 - -
:q
D
• ql
C
F
\ 9.25 - -
\
strength becomes lower. The flux inside the shell, at the tube bank mid-height at position F, is observed to be many orders of magnitude higher than outside the shell thus giving an indication of the attenuation in the shell wall itself. Statistical variances are relatively poor compared with results obtained for earlier examples, but could be improved given longer computational running times and the use of more sophisticated variance reduction techniques. To examine the changes in spectral shape, the photon spectrum at the various locations A to F were compared with the 'raw' surface emission source density spectrum on the inner surfaces of the steam generator. With the exception of the magnitude of the group values, the spectral shapes were very similar to those shown for the 'hot-leg' in Fig. 4 and are therefore not repeated here.
7. Dose calculation and results 4.0 - -
1.5
-
0.0
-
-
:,l
A
-
Fig. 5. Schematic of the steam generator showing detector locations for flux tallies. All dimensions in metres and not drawn to scale,
in the schematic diagram in Fig. 5. Locations A to E are all external to the shell at various heights and at 1 m from the shell surface. Location F, is at the mid tube bank height at a position between the inner shell and outer tube bank windings. The MCNP-4 calculation was run allowing for the tracking of 150 thousand particle histories. The total fluxes at locations A to F together with their statistical standard deviations are given in Table 3. The photon flux is seen to peak opposite the mid-height position of the tube banks, location C, where the local source strength is highest due to the large surface areas present. It diminishes in the axial direction towards the top and the bottom of the steam generator as the source
Once the photon flux spectrum (discretised into 24 energy groups) has been calculated at the detector position, the dose rate in human tissue is computed. This is tissue (organ) dependent and therefore critically dependent on the orientation of the body with the photon flux. To simplify the analysis, energy dependent dose conversion factors, v, which provide a weighted average of tissue types and relative body orientations are used [8]. Over the 5 KeV- 14 MeV photon energy range the dose conversion factors are shown in Fig. 6. These data are fitted with a high order polynomial from which the values at mid-point energies in the 24 group structure can be determined. Table 3 Photon fluxes (with statistical standard deviation) around the steam generator Location
Total photon flux, ~b (7 m - 2 s - I )
~r ( + %)
A B C D E F
8.42 2.65 4.50 2.80 3.68 1.93
22.1 9.9 8.7 10.0 20.6 8.0
× × × × × ×
106 107 107 107 106 10 l°
C.B.A. Forty~Fusion Engineering and Design 36 (1997) 191-201 14x10-12
.... r
........
i
........
i
........
7.2. Dose rate around the steam generator
i
12x10-12
1 0 x l 0 .12
~7 8 x 1 0 -j2
6x10 J2
4x10-t2
8 2x10-12
0xl0 °
0.01
0.l
l
10
photon energy (MeV)
Fig. 6. Dose rate conversion factors, v, at discrete energies plus high order regression fit.
Having obtained the dose conversion factors in the same group structure as the detector fluxes, all that remains is to determine the dose rate in human tissue, D, by: dD
24
¥ = t~"1"=~ivi
199
(1)
where ~bi, is the group energy partial photon flux. 7.1. Dose rate near the 'hot-leg' pipe
The dose rate and photon flux are shown together in Fig. 7, at the mid-length position and at 1 m from the 'hot-leg' pipe surface. Both flux and dose spectra exhibit broadly similar features with peaks occurring at the same energies and the fall off at the top and bottom energy ranges. Most of the absorbed dose occurs in the energy range 0.2-3.0 MeV. The lower energy dose rate arises from the multiply scattered photons, while the majority of the dose rate at 2.5 MeV is due to uncollided 24Na photons (E~ = 2.754 MeV). The total dose rate is of order 2.2 laSv h - t , which is low enough to pose no important radiological risks. In fact, this figure is only about six to ten times larger than the average natural background radiation levels found at sea level (which gives an annual dose of around 2-3 mSv year- t, i.e. approximately 0.23-0.34 pSv h -1 [17]).
Photon fluxes in the 24 energy group structure obtained for locations A to F were converted to dose rates by multiplying by the appropriate dose conversion factors, v. The summation at each location for each of the 24 groups finally yielded the total dose rates. Since the shape of the flux spectra and dose rate spectra are similar for all locations, only two locations are used to illustrate the behaviour, namely locations C and F. Fig. 8 shows the results for the 'internal' location, referring to the tube bank mid-height position inside the shell wall and the 'external' location referring to the same axial height position but 1 m from the outer shell wall. The 'smoothing' of the flux after attenuation through the shell wall is evident on comparison of the internal and external locations. A relative reduction in the peak at 2.75 MeV is also seen. As before, dose rate curves show similar spectral shapes to their corresponding flux spectra. The maximum external dose rate (summed over all energies) occurs at location C and has a value 166 ~tSv h - l , whereas the dose rate inside the shell is s o m e 10 3 times higher. The external gamma field around the outside of the steam generator, whilst moderately high, would still permit individual worker access for over 100 h, and is 106
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200
C.B.A. Forty/Fusion Engineering and Design 36 (1997) 191-201 1012
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estimate gave the cooling circuit ORE at about 80 man-mSv year-~ [1] which is an equivalent result bearing in mind the differing technique used in the flux calculation. This value is also similar to the 200 man-mSv year-1 reported for modern Konvoi PWR plants [19] which are currently regarded as the lowest ORE plants anywhere in the world [2], after decades of optimisation. Thus on this simple estimate, the SEAFP PM-1 ORE is expected to be equivalent to modern fission power plant, already considered to be well inside acceptable limits.
,
, , ,11
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Fig. 8. Comparisonof photon flux and dose rate spectra for two locationsaround the steam generator. therefore well within acceptable levels. Any maintenance work required inside the steam generator shell (for example, pressure tube repair or inspection) would however, require very strict working practices to minimise dose exposure. Tasks of this nature would, in all likelihood, be performed remotely or robotically.
8. Annual ORE estimation
To estimate the annual collective dose to all plant personnel, a detailed analysis of all radiation fields around the plant, derived air concentrations (DACs) in contaminated rooms and the occupation task breakdown would be required. In relation to anticipated ITER OREs, Sandri [18] has developed an effective methodology for this type of estimation. The detail required for an accurate assessment is, however, well beyond the scope of the present work and is not attempted here. However, a simple estimate may be made as follows. Noting again that the cooling circuit doses dominate ORE and an approximate correlation that a dose rate of 1 mSv h - 1 is equivalent to 1 man-Sv y e a r - l [2], the annual ORE due to the cooling plant systems (dominated by the steam generator contribution) is estimated to be approximately 170 man-mSv year- 1. The earlier SEAFP
9. Conclusions
Near term magnetic confinement fusion power plants will be based on the D-T fuel cycle. The resulting high energy neutrons will sputter radioactive ions from the blanket pipe walls into the coolant. In the absence of other source mechanisms, such as corrosion, dissolution and erosion, this sputtering term is the principle means by which the cooling circuits become activated. This is likely to be the case for the conceptual heliumcooled power plant, SEAFP PM-1. Active material is transported around the loop and deposited on internal pipe wall surfaces and other plant equipment such as steam generators and pumps/ circulators. Much of this equipment lies outside the biological shield where man access is allowed, resulting in potential occupational radiation exposures (OREs) to plant personnel. This ORE has been assessed as follows: (A) the photon fluxes in space have been determined near to critical plant regions; (B) based on these fluxes, dose rate have been calculated; and (C) the annual ORE has been estimated from the dose rates. Two regions of the coolant loop were modelled, namely: 1. a 'hot-leg' pipe; 2. a steam generator. Average surface deposit concentrations were calculated at these locations for a range of important radionuclides using the HETRAN code. These activity concentrations were next converted to a surface emission source density (SESD) through the FISPACT/EAF-4 code/library system. The
C.B.A. Forty~Fusion Engineering and Design 36 (1997) 191 201
7-fields surrounding the cooling plant equipment were determined using the MCNP-4 Monte Carlo code. Partial fluxes were calculated in a 24 energy group structure format, allowing for easy conversion to absorbed dose rate in human tissue using dose rate conversion factors. Dose rates around the 'hot-leg' pipe were calculated to be extremely low ( ~ 2.2 laSv h - l ) and not significantly above naturally occurring background levels. Because of the thinner shell wall and much greater surface areas available for deposition, the steam generator dose fields were found to be several orders of magnitude higher at around 166 ~tSv h - 1 . Whilst higher, these dose levels are well within acceptable levels and would permit worker residence times in the region of such a field for as long as 100 h. Dose rates inside the steam generator shell are considerably higher still and any tube bank inspection or maintenance regime would probably require remote or robotic methods. Based on these dose rates and using a crude scaling factor, the annual collective ORE from the whole plant, without any attempt at optimisation, is estimated to be approximately 170 man-roSy h - ' , which is equivalent to the levels achieved with the best PWRs operating today, after decades of practical optimisation.
Acknowledgements I should like to express my thanks to Dr P.J. Karditsas and Dr N.P. Taylor, both of UKAEA Fusion, for their valuable help. The funding for this work was provided by the UK Department of Trade and Industry and Euratom.
References [1] J. Raeder et al., Safety and Environmental Assessment of Fusion Power, European Commission Report No, EURFUBRU XII-217/95, 1995.
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[2] C.G. Kinniburgh, Occupational radiation exposure, Report No SEAFP/R-A10B(94), 1994. [3] C.B.A. Forty, A nuclear sputtering model for fusion applications, UKAEA Fusion/Strategic Studies Note, SSN 96/7 (1996). [4] P.J. Karditsas, C.B.A. Forty, Modelling and simulation of activation product transport in helium cooling loops, Fus. Eng. Des. 36 (1997) 203-217. [5] P.C. Miller, RANKERN Users Guide, ANSWERS (RANKERN) 2, UKAEA Winfrith Report, 1993. [6] R.G. Jaeger (Ed.), Engineering Compendium on Radiation Shielding, Vol. 1, Springer-Verlag, Berlin, 1968. [7] C.B.A. Forty, Gamma radiation fields around the cooling circuits of a fusion power plant, UKAEA Fusion Report No., UKAEA FUS 325 (1996). [8] J.F. Briesmeister (Ed.), M C N P - - a general Monte Carlo code for neutron and photon transport, Los Alamos Report No, LA-73396-M, Rev. 2, 1986. [9] F. Engelmann et al., Specifications of reactor parameters and plasma operation, SEAFP Task M1 Final Report, R-M1-Rev.3, 1994. [10] W. Daenner and G. Simbolotti, Blanket design, SEAFP Task M6 Final Report, RM6-Rev.1, 1994. [11] J. Wagner, G. Sable and F.J. Mustoe, Cooling system design--helium option, SEAFP Task M8 Final Report, R-MS.2-Rev.1, 1994. [12] ICRP, Recommendations of the International Commission on Radiological Protection, Publication 38. Ann. ICRP 11-13, Pergamon Press, Oxford, 1983. [13] E. Browne, R.B. Firestone, Tables of Radioactive Isotopes, Wiley, New York, 1986. [14] D.C. Kocher, Radioactive decay tables, US Department of Energy Report, DOE/TIC-I1026, 1981. [15] J. Kopecky, D. Nierop, The European activation file EAF-4, Petten Report No ECN-C-95-072, 1995. [16] R.A. Forrest, J.-Ch. Sublet, FISPACT 4.1 user manual, UKAEA Report No., UKAEA FUS 287, 1995. [17] UNSCEAR, Sources and effects of ionizing radiation, United Nations Scientific Committee on the effects of atomic radiation, Report to the General Assembly, with scientific annexes, 1993. [18] S. Sandri, EU Task SEA2: personnel safety, ITER Task No S81TD01/D7105, ENEA Report No, ERG-FUS/ TECN S + E 13/95, 1995. [19] G.H. Hecht, Occupational exposure in LWR plants built by Siemens/KWU, International workshop on new developments in occupational dose control and ALARA implementation at nuclear power plants and similar facilities, NUREG/CP-0110, Brookhaven National Laboratory, September 1989.