Radiation Protection at New Reactors

Transcription

1 Radiation Protection at New Reactors A.Brissaud EDF INDUSTRY Basic Design Department EDF- SEPTEN VILLEURBANNE Cedex FRANCE 1 - INTRODUCTION The knowledge concerning the sources, dose rates and doses in a PWR does not come out of thin air. Up to the mid 70s the knowledge of EDF concerning actual sources of doses in PWRs was scarce. At this time, it was thought that the major source consisted in fission products (FP). Corrosion products (CP) deposits that are now recognised to be the main origin of occupational radiation exposure were not quantifided. Thus, while 50 PWR units were ordered by EDF between 1974 and 1980, the technical EDF report indicating more realistic values for FP and the existence of activated CP deposits was published in This period was used by EDF to participate in the development by the French Atomic Energy Commission (CEA) of the dedicated codes: PACTOLE and PROFIP respectively for the modelling of CP and FP behaviour (creation, transport, deposition) in a PWR. They were very useful to show which actions (cobalt content of alloys, primary chemistry control, cold shutdown procedure ) would be efficient in keeping sources low. Eventhough a low level of sources was considered essential to obtain low doses, the influence of the layout was recognised and a set of good practices was elaborated and applied. As a result, none of the operating EDF power reactor has been designed with quantified radiation protection targets in terms of individual and collective doses, although noticeable efforts have been made to reduce doses. In the following, EDF feedback of experience and results are summarized in order to justify the methodology to apply throughout the design of a new reactor. This paper is focused on the optimisation of doses to workers eventhough the optimisation process at the design stage of a new reactor has to take into account the protection of the public, waste production and storage, dismantling stage, dose trade-off between categories, or time. 2 - EDF FEEDBACK OF EXPERIENCE This paragraph is limited to the presentation of results that are taken into account in the design of a new PWR reactor. 2.1 Sources Reactor building during operation of the reactor The design of the shielding of EDF reactors is a rather opened one: although the thickness of the shielding is theoretically sufficient to obtain low dose rates (neutrons and associated γ flux, γ emission of 16 N), the actual situation is quite different. In fact, operating requirements (ventilation of the reactor pit, core neutron flux measurement) and safety (pressure relief during accident sequences) requirements lead to the necessity of a number of openings in the shielding. These discontinuities allow the streaming of neutrons and the propagation of γ fluxes a long way from the core. In the reactor building of the first EDF units, although the reactor pit wall was sufficient to respect a design criterion of 0.15 msv.h -1 at contact of the outer wall surface, dose rates up to a few msv.h -1 were measured at some locations. This was not acceptable because accessibility for a limited time during operation was required. This problem has been solved by addition of special shielding. However, it remains a very difficult problem to solve because the shields must respect contradictory requirements. Another problem to solve when access is required is the air contamination: 41 Ar (arising from the activation of the reactor pit ventilation airflow); 3 H when operating with primary coolant leaks (within the limits set by safety requirements); Gaseous and volatile fission products (when operating with primary coolant leaks and fuel cladding defects) Reactor and Nuclear auxiliary buildings during cold shutdown procedure During this period, the large changes in the physical (pressure, temperature) and chemical (ph, [H 2 ], [O 2 ], [B] [Li]) conditions of the primary coolant provoke a spiking of the coolant activity. All categories of isotopes present in the coolant are subject to this spiking phenomenon. The spiking factor depends on many parameters among which the available quantity of the isotopes, properties of the isotopes, and the shutdown procedure itself. The main consequences of the spiking are: The contribution of the activity in the water, which is usually negligible, compared to that of the deposits becomes important. The accumulation of these products in filters and ion exchangers must be taken into account for the shield thickness of the relevant cubicles 1

2 The time necessary to go back to reasonable activity levels in the water should be taken into account for the determination of the purification flow rate and design (type of purification systems) Reactor and Nuclear auxiliary buildings during cold shutdown The collective dose for inspection and maintenance tasks during this period represents ~85% of the annual dose. The dose rates near the circuits are resulting from the deposits of radioactive products. Fission products Fission products generally do not represent a large participation in a given dose rate. The reason is that the occurrence of cladding defects is not frequent. However, even with low coolant activities, some fission products are incorporated in the corrosion product deposits (as well as α emitters with particular types of cladding defects). This has to be taken into account as a possible source of airborne contamination when preparing the task conditions. Corrosion products Corrosion products are the main source of occupational radiation exposure, with two dominant isotopes: 58 Co and 60 Co. This statement should never be disconnected from the design of the units or from the operating conditions. A low corrosion products source term can be obtained using alloys presenting a low release by corrosion or ware of materials and/or alloys with a low content of 58 Ni and 59 Co (and, more generally, of other precursors of other radioactive isotopes). A low content of the alloys in precursors should not be considered as a sufficient condition. It has been observed that the release could vary in large proportions with the manufacturing process (from the alloy to the piece of component). For example, the alloy 600 used for the manufacturing of the steam generator tubing, in a same unit, presents a release rate (and deposition rate) by normal corrosion that can differ by a factor 3 to 5 with the manufacturing process. Hard facing materials with high 59 Co represent at least a potential source in normal conditions and a real high source of 60 Co in case of abnormal ware or corrosion. It is also recognised that the primary coolant chemistry (pht, O 2 ingress ) is very important to control the release and deposition of corrosion products. Activation products 110M Ag at a high activity level is observed when the control rod cladding presents defects. 124 Sb can result from defects of the cladding of neutron secondary sources (Sb-Be) but also from the wear of pump bearings using antimony to reduce friction. 2.2 Lay-out Some improvements in the layout have been made in the recent units. However, based on the experience of the people operating the units, further improvements have been suggested for new units. Of course, they cannot be detailed here but as a summary, the general idea is to improve the ergonomy of task places (mainly maintenance tasks). 2.3 EDF results in terms of dose rates and Operation and maintenance of the units The average dose rate around the primary circuit after shutdown is representative of the source level. The dose rates measurements are performed within the frame of a procedure. This allows one to perform studies such as evolution with time, comparisons between units with the same design There is not a direct correlation between the average dose rate and the collective dose per year of the units. However, one can expect that moderate or low dose rates are favorable conditions. AVG DOSE RATE 1E-2 msv/h A V G 900 M W e O LD A V G 900 M W e A V G M W e Figure 1. Evolution of the average dose rate near the primary circuit versus cycle number and unit type NB: 900 MWe old : 6 units; 900 MWe: 28 units ; 1300 MWe: 20 units 2

3 Figure 1 shows that: The average dose rate are nearly stable or decreasing with time after 3 cycles There is a decrease with improvement of the design (avg 900 MW does not include the 900 MW old) Nevertheless, these conditions were not sufficient to keep the doses low as shown in figure 2 below man*sv*unit -1 *y -1 FRA USA ALL JAP ARROW SHOWS TENDENCY FOR FRANCE Figure 2. Evolution of the dose index in man. Sv/unit/year versus calendar year Figure 2 shows that the moderate dose rates in the units were not sufficient to maintain the fairly good results in terms of collective dose obtained in the 80s. The analysis of the causes lead to 3 major conclusions: The exposed work volume was steadily increasing The organisation and preparation of work could be largely improved The ALARA culture was not sufficient (management commitment, workers risk awareness) This lead to the launching of the ALARA PROJECT at the beginning of the 90s with two objectives for 2000: Annual average collective dose per unit: 1.2 man.sv No individual dose higher than 20 msv (excepted exceptional circumstances). The first target was reached in 1999 (1.16 man. Sv/unit/y) and it is expected to reach the individual dose target this year (in 1999: 8 on a total of ~40000 workers were above 20 msv).these results were obtained without significant modifications in the layout of the units, thus showing the importance of a good ALARA management and thinking. The approach of the problems has been based on experience and pragmatism as opposed to an analytic methodology Steam Generator Replacement Operations (SGR) The first SGR operations outside of France were performed at a high dose cost. Thus, EDF decided an extensive and careful preparation. The study addressed all relevant parameters (dose rate, exposed work volume,) on a task by task basis. Simulations were performed to predict dose rates for which measurement in due time was not possible (for example: after cutting primary pipes and lifting the SG). The corresponding data were stored in the dedicated data base software DOSIANA that allows to point out easily dose consuming tasks, either collective, individual or for a given craft type. This lead to ALARA option studies. 3

4 7,00 6,00 5,00 man.sv/sg 4,00 3,00 2,00 1,00 Figure 3. Doses for SGR throughout the world. EDF results in dark. This figure shows that the analytic approach is very performing: EDF results are reproducible, even decreasing, and among the best. 3 - RADIATION PROTECTION AT NEW REACTORS The project of a new reactor benefits from the experience (EDF and international knowledge). As far as EDF is concerned, lessons learned (some of them quite evident) are as follow: A number of methods to reduce sources (particularly corrosion products deposits) are known as efficient and agreed upon by the international community; however, particularities in the design or operating parameters and practices can have a large influence on the efficiency. For this reason, the efficiency of a given action should always be checked against the characteristics of the project when operation practices are proved efficient to reduce sources, provisions should be made at the design stage to facilitate it s application (for example: the primary coolant chemistry control is not easily performed without appropriate means) Even when all provisions are made to obtain a low source term level, less favourable conditions should be taken into account. This is usually done by providing to set of values: design values and realistic values provisions should be made at the design stage in order to allow the decontamination of piping and components (decontamination operations can be thought as routine or exceptional tasks, and to facilitate dismantling operations) a low level of radioactive sources and corresponding dose rates is not sufficient to ensure that doses to workers will be kept low; however, the dose rate is certainly more structuring than other parameters the layout is not independent of the technology of components with a given source level, the resulting dose rates depend on the lay-out and the technology of components providing large areas for the lay-out of components and piping is not sufficient in itself to reduce doses : the area provided should be sufficient in order to facilitate the works to be performed, but the utilisation of the provided area should be carefully thought The option to use remote handling or robots should be accompanied by the corresponding provisions in the layout. Organisation and preparation of tasks have a large influence on dose results. Provisions should be taken at the design stage in order to facilitate the work to be performed. This can be done by a careful study of the ergonomic aspects 4

5 INTERACTION COOLANT / ALLOYS + cladding defects OPERATING CONDITIONS CARACTERISTICS of DESIGN & LAY- OUT Accessibility PREPARATION & ORGANISATION of WORK W A L L T H I C K N E S S T E C H N O L O G Y Type : FP, CP,AP Activity, spectrum Repartition ON COMPONENTS Localised sources Spatial Repartition OF COMPONENTS Spread sources Actual duration & number of workers Working Conditions [ Contact DR ] [ Ambiant DR ] DOSE RATES EXPOSED WORK VOL. [actual job] D O S E CARACTERISTICS of COMPONENTS Figure 4. Liability Schematic representation of collective doses Minimum task duration Nature of task Necessity Frequency [theoretical job] The individual and collective doses depend on a large number of parameters, generally not independent, as tentatively shown in figure 4. This justifies an analytic approach when predicting doses. 4 - INTRODUCTION OF THE OPTIMISATION PROCESS IN THE DESIGN OF A NEW PROJECT The management of a large project such as a new reactor is a very complex matter. On the one hand, all components should be taken into account, but on the other hand, not any of these components should stay on the critical path of the project studies. The status of radiation protection in a new project should stay within the frame of the optimisation process (ALARA) for at least two reasons: it is imposed by the legislation and it is the best way to cut undue costs. A system of monetary values of the saved man.sv is useful. Using such a system leads to more rationality and coherence in the decision making process. This system should always be considered as only an aid to the decision making process. Collective and individual dose targets should be set for a new project. The collective dose target value should be fixed taking into account the current dose results for similar operating installations and the period at which the installation will be operating. One can expect the dose acceptance to go downward rather than upward. The target value will reflect the ambitions of the project. It should not be considered as a contractual value but rather than a guiding value. Thus, the status of the target value would be a dose constraint. The individual dose targets values should be aimed at respecting the legal limits in the first place and at reducing the differences in average doses between categories of workers. These values should have the status of dose constraints. The notion of dose constraint is here interpreted as a useful guide to limit the number of ALARA studies. 4.1 Practical radiation protection optimisation There are many reasons to think that an overall management process, that is, covering at the same time all the parameters, and is not practicable. The usual method to solve this kind of problem is to split it following the different steps of the project (usually basic design, basic design optimisation, detailed design ). However, this should not imply that radiation protection optimisation is note taken into account in any of the above steps. 5

6 Whatever the number or definition of the steps, the project should include a dedicated ALARA organisation Basic design step During this step, basic options are considered that can have incidence on radiation protection optimisation. For example: Overall dimensions of the nuclear buildings : that will somewhere determine the extent of available space in cubicles and thickness of shield walls between cubicles or components Status of the reactor building as regard to accessibility during operation at power Source term description and quantification (what isotopes? in which form? Where and when? How much?) Flow rate of the primary coolant purification system During this step, passive radiation protection principles are applied. These principles address (not exhaustive lists): Source term Provisions to have an acceptable source term (optimisation is not carried out at this step): Limitation of 59 Co in materials Elimination of hard facing materials with high 59 Co content as much as possible Interdiction of the use of antimony in materials of components linked to the primary coolant system Use of materials presenting a good resistance to normal (or particular) corrosion and ware General layout rules They are aimed at: Limiting dose rates (shield between cubicles and components), preventing accumulation of radioactive products (bottom of tanks, discontinuities in piping,) facilitating access to cubicles and components making provisions for the ergonomic of working locations, the use of robots preventing dissemination of radioactive products throughout the cubicles and buildings The result of this passive protection is somewhat translated by the zoning of the controlled area and the flow chart of the radioactivity (activity in components, transfer of activity in systems). In both cases, design and realistic values are considered, as well as operating conditions (reactor in operation or during shutdown) During this step, the possibility to respect the collective dose target is roughly checked, based on experience and correcting factors as opposed to an analytic approach. These principles and rules should be detailed in the form of checklists Basic design optimisation step At this step, only those options that can have a not negligible impact on the general design are studied. A general list cannot be established because the options are specific to the project Detailed design step. The optimisation studies will be performed during this step. They will address general matters, with reference to general principles as described before For example: The extent of the limitation of 59 Co as impurities in materials will be checked against costs (if any) The limitation of the use of hard facing materials with high 59 Co will be checked on a case by case basis (pump bearings, parts of the core internal structures ) against feasibility, liability and costs The analytic approach described in this paper will be performed during this step. It will result in a task by task evaluation of doses (collective and contribution to individual doses). This will be performed with the help of any dedicated data base provided the software allows an easy screening of the results on a given criterion or a set of criteria. This will show the points where it is necessary to carry out detailed option studies in the first place, necessary to the respect of the dose constraints on the one hand, and to obtain an optimised situation on the other hand (check options to reduce dose rates or doses against feasibility and costs). Of course, the process will be iterative versus the level of detail. The end of the three steps should be materialised by a radiation protection reference document including: The list of regulatory texts taken into account during the design The list of the internal rules for the design 6

7 The data base with the predictive values of all dose relevant parameters The references of optimisation studies This document will be somewhat a user s guide as well as the one of the summary of the first step of the optimisation process (prevision). 5 - CONCLUSION The theoretical knowledge and the feedback of operating experience concerning radiation protection at nuclear power plants are now considerable. It is available to the designer in the form of predictive software and databases. Thus, it is in principle possible to include the radiation protection component throughout all the design process. However, the application is not straightforward because of the complexity of a project on the one hand, and of the transverse characters of radiation protection on the other hand. For this reason, a dedicated ALARA organisation should be set within the general organisation of the project. A practical solution to deal with the complexity of the project is to adopt a stepwise approach taking into account the different steps of the project itself. More precisely, a three steps approach can be: The definition of principles and rules to be applied during the basic design step in order to obtain a favourable situation for radiation protection optimisation in the next steps ( passive radiation protection ) The study in the context of radiation protection optimisation of options that have a not negligible impact on the project during a basic design optimisation step The optimisation is carried out in the detailed design step and comprises general and detailed options, taking into account all relevant parameters (feasibility, liability of alternative solutions and technologies, costs ) Whatever the steps, the first action is to set dose targets taking into account the performance in term of doses of existing units and the time at which the unit will be operating. These values should be interpreted as dose constraints as opposed to contractual values. They are guidance throughout the duration of the project and are useful to limit the number of options to study. Another helpful guidance is to use a system of monetary values of the saved man.sv is useful. Using such a system leads to more rationality and coherence in the decision making process. This system should always be considered as only an aid to the decision making process. The final result should be materialised by a radiation protection reference document that can be considered as the result of the first step of any optimisation (ALARA) process (prediction / measurements / analysis of discrepancies) but also as some kind of user s guide. Provided relevant adaptations, the optimisation approach described in this paper can be applied to nuclear facilities other than nuclear power plants. 7