Decommissioning of light-water reactor nuclear power plants

Nuclear Engineering and Design 45 (1978) 1-51 1 © North-Holland Publishing Company

DECOMMISSIONING OF L I G H T - W A T E R R E A C T O R N U C L E A R P O W E R P L A N T S *

R. B A R D T E N S C H L A G E R , D. B O T T G E R , A. G A S C H and N. M A J O H R

Nuklear-lngenieur-Service GmbH, Frankfurt/M, Fed. Rep. Germany

Received 5 April 1977

This study deals with the technical and economic questions posed by the decommissioning of light-water reactor nuclear power plants of the 900-1300 MW class, account being taken of the distinctions between boiling- and pressurized-water reactors. Possible decommissioning alternatives and the disposal or confinement of activity are discussed, It emerges from the discussion that decommissioning, and even total dismantlement of these nuclear power plants is in principle feasible.

The activity inventory, one year after shutdown, is calculated to be about 3 × 107 Ci for the BWR and 4 × 106 Ci for the PWR; 40 years after shutdown these figures are reduced to 2 × 106 and 4 × 105 Ci, respectively.

~l'he decommissioning costs to be expected are also estimated. This estimate serves as the basis for an economic comparison by the present worth method. The economic comparison shows that total dismantlement after a cooling time of one year is more than four times as expensive as interim eonf'mement followed by total dismantlement waiting period of 40 years. The present worths for immediate total dismantlement are estimated at DM 200 million for the BWR and DM 170 miUion for the PWR; for the other alternative, they are put at DM 45 million for the BWR and DM 42 million for the PWR.

A still open question is posed by the final storage of the large quantities of bulky radioactive waste arising in partial or total dismantlement. Since no decision on the storage method has yet been taken, disposal in casks is stipulated as a boundary condition in the estimation of the costs, although this is an unrealistic assumption. It is to be presumed that the costs of disposal can be reduced given appropriate final storage.

C o n t e n t s

1. Introduction 2. Fundamentals 2.1. Reasons for decommissioning of nuclear power plants 2.2. Decommissioning: definitions 3. Sequence of operations in the decommissioning alternatives 3.1. General boundary conditions 3.2. Interim corff'mement 3.3. Partial dismantlement with secure residual confinement 3.4. Total dismantlement 4. Activity inventory 4.1. Calculations of activity inventory of structural materials 4.2. Estimation of contaminated inventory 4.3. Estimation of dose rates 5. Decommissioning costs

* A study carried out by Nukleax-lngenieur-Service GmbH (NIS) for the Commission of the European Communities, Energy Industry Directorate, Power Plant and Nuclear Fuel Industries Division. This report was originally prepared in German: the German version may be obtained from the Office for Official Publications of the European Communities, B.P. 1003, Luxembourg, refering to document No. EUR 5728.

2 NIS / Decommissioning ofL WR nuclear power plants

5.1. Aspects which cannot be assessed in financial terms 5.2. Boundary conditions 5.3. Cost estimates 5.4. Surveillance requkement for decommissioned plants 6. Economic comparison of the decommissioning alternatives 6.1. Boundary conditions 6.2. Calculation of cash values 6.3. Comparison of costs 7. Proposals for the design of future nuclear power plants with a view to their decommissioning

1. Introduction

Current nuclear programmes in the European Com- munity are primarily based on PWR and BWR power plants with ratings ranging from 900 to 1300 MW (e). With the decommissioning of these nuclear power plants at the end of their normal lifetime, a number of specific problems arise, due to such factors as the activity inventory of the plant and contamination of components and systems.

2. Fundamentals

2.1. Reasons for the decommissioning of nuclear power plants

A nuclear power plant may be closed down because an economic limit has been attained, due, for example, to higher prime costs of electricity generation, labour costs and fuel cycle costs, adaptation to new safety constraints, low availability, etc. Another reason for decommissioning may be the reaching of the technical lifetime limit. Another possibility is that decommissioning may become necessary after a major accident which makes the continued operation of the plant impossible for technical or safety reasons.

The following discussion deals only with decommissioning following a normal shutdown; cases of decommissioning due to major accidents are not considered.

2.2. Decommissioning: definitions

The first distinction to be made in connection with decommissioning is between active (activated and contaminated) and inactive substances. The former can be confined safely within the power plant unit or dis-

posed of externally. The latter are not covered by nuclear legislation and can therefore be dealt with by traditional methods.

Several alternatives are conceivable for the decommissioning of a nuclear power plant. The alternative eventually chosen is the result of a cost optimization in which account is taken of licensing and contractual as well as of technical possibilities.

The principal alternatives are as follows (see also fig. 1):

(1) interim conf'mement; (2) partial dismantelment with secure residual con-

finement; O) total disposal; (4) interim confinement and subsequent partial

dismantlement with secure residual confinement; (5) interim confinement and subsequent total

dismantlement; and (6) conversion of the nuclear power plant into a

final repository for the storage of radioactive waste.

Circumstances associated with the plant itself or with

the site, as well as requirements by the authorities, may lead to intermediate or other solutions, such as

(7) interim confinement and subsequent partial dismantlement with secure residual confinement, followed ultimately by total dismantlement;

(8) interim confinement within the existing containment and total dismantlement of the remaining buildings;

(9) dismantlement of active components; and (lO) dismantlement of all components.

Since the last four items are basically combinations or parts of the principal alternatives ( I ) - (6) , they will not be discussed further.

For purposes of 'final decommissioning', all the alternatives except that of total dismantlement con-

NIS / Decommissioning o f L WR nuclear power plants 3

Decommissioning of nuclear power plan±

Interim / ~ confinement I

| f Partial [ [dismantlement ! |with secure l |residual

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Fig. 1. Possible combinations of the decommissioning alternatives.

stitute interim solutions which must all eventually be followed by total dismantlement. The principal alternatives are defined in the following subsections.

2.2.1. Interim confinement The plant is brought to a condition in which the

safe confinement of all solid and insoluble active substances is ensured. Safety surveillance of and hazard- free access to the plant are made possible for the entire duration of the confinement. No conservation measures are taken; all buildings remain standing. It can be reliably assumed that the authorities will sanc- tion this condition for a limited period only. The reasons for this are given in section 5.1. Interim confinement is thus in fact only one stage in the decommissioning process.

is kept under safety surveillance and to wldch hazard-free access is possible. The materials to be disposed of are, if necessary, broken up into small pieces, packed, and stored in the remaining residual confinement.

The final condition of this alternative may vary with configuration. It is largely determined by the type of plant (BWR or PWR) and also depends on the construction (pressure vessel at high or low level within the reactor building) and on the requirements laid down by the authorities as to disposal (e.g. all buildings above ground level, except the confinement structure, may have to be demolished).

In commercial nuclear power plants in the 9 0 0 - 1300 MW bracket, the foundations of the reactor building as a rule do not lie deeper than 10 m. This means that the spaces below site datum level are seldom sufficient to take all activated and contaminated parts. For this reason, additional structures above site datum must be used for storage. Furthermore, the reactor pressure vessel is itself usually above site datum. In BWR plants, the pressure vessel is situated relatively high up in the reactor building. Even if the vessel can be lowered to the deepest point in the confinement structure for decommissioning, a considerable part still remains above site datum.

This alternative was originally considered in this study as a method of ultimate elimination. Closer examination showed, however, that this is not very realistic, so that this alternative is only an interim solution, which is, in fact, not even very attractive.

2.2.3. Total dismantlement In 'total dismantlement' all active and inactive Com-

ponents, including foundations, are demolished and the active parts broken down into small pieces - if necessary - and transferred to a final repository. Af- ter elimination of the activity inventory the site is no longer subject to the provisions of nuclear legislation and can therefore subsequently be used for other put- poses without restriction.

2.2.2. Partial dismantlement with secure residual confinement

In this alternative partial dismantlement of active and inactive components takes place. All solid and insoluble active residual components remain for a long period in the sealed reactor containment which

2.2.4. Interim confinement and subsequent partial dismantlement with secure residual confinement or total dismantlement

These solutions are combinations of the alternatives already discussed. The time elapsing between the stages of decommissioning serves to permit cooling of activity

4 NIS / Decommissioning elL WR nuclear power plants

so as to facilitate handling during transfer to the final decommissioning configuration. This 'interim period' depends on the individual plant, the permit situation and the possibilities of disposal; it may range between a few years and several decades.

2.2.5. Conversion era nuclear power plant into a final repository for the storage o f radioactive waste

There exists at present no satisfactory method for the final storage of large quantities of bulky, mainly low-activity waste. One possibility, valid at least as a temporary solution, is to use the reactor and auxiliary buildings of decommissioned nuclear power plants.

For instance, if a nuclear power station consists of several units, the first unit to be decommissioned can be converted into a 'nuclear cemetery' to take radioactive decommissioning waste requiring storage and surveillance from the other units. This possibility is also applicable to a number of plants which, whilst not at a single location, are relatively close together. In this way transport over long distances can be avoided. Sources of radioactivity already present in the plant are dismantled and stored en masse in a single area of the plant. At the same time the structure of the buildings is prepared for the accommoda- tion of bulky radioactive waste from other nuclear power plants. The inactive parts are removed.

The confinement structure and monitoring facilities already available in the plant are very well suited to this use.

3. Sequence of operations in the decommissioning alternatives

In this section the sequence of operations entailed in the following decommissioning alternatives is described and illustrated with flow charts: (1) interim conf'mement; (2) partial dismantlement with secure residual con-

finement; and (3) total dismantlement.

The alternatives of (1) interim conf'mement and subsequent partial dis-

mantlement with secure residual confinement and

(2) interim confinement and subsequent total dismantlement,

represent combinations of the alternatives mentioned above. For this reason, the sequence of operations in these cases is not described here.

Each type of nuclear power plant, indeed each individual plant, requires a decommissioning procedural plan of its own, because differences in the sequence of operations may arise owing to variations in design and boundary conditions. However, the present discussion has deliberately been kept general enough to be applicable to all commercial LWR power plants of the 900-1300 MW class.

3.1. General boundary conditions

It is assumed in all cases that when decommissioning work begins there are no longer any fuel elements or active operating waste such as filter cartridges, sludges or concentrates in the plant. Removal of these substances is a normal operating procedure.

3.2. Interim confinement

3.2.1. Boundary conditions , (1) Interim conf'mement relates to the reactor

building and the auxiliary building in which there are radioactive substances.

(2) All other buildings are left in their original condition and sealed after removal of operating media.

(3) Except for part of the ventilation, lighting and drain pump installations, which are used where necessary, and part of the radiological protection instrumentation, no systems continue to be operated after completion of the decommissioning works.

(4) No active or inactive components are removed. This means that maintenance costs will be somewhat higher.

(5) Hazard-free access to the plant is ensured after decommissioning.

3.2.2. Sequence o f operations The sequence of decommissioning operations is il-

lustrated diagrammatically in fig. 2 (the figures in brackets refer to the numbering of the operations in the diagrams). The general executive planning (1), during which the decommissioning documentation is compiled, is followed by the preparation of a decommissioning safety report (2). General preliminary works (3) not affecting the existing permits can already be commenced. The documents compiled under

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items (1) and (2) constitute the basis for the project consent (4). The detailed planning phases (10) which follow the preparation of the executive plan keep pace with all the decommissioning operations right up to their completion. The individual partial consent procedures (9) also take place in parallel with the decommissioning operations. When the project consent has been obtained, all liquids except the reactor coolant and storage pool water are treated sufficiently to allow them to be discharged (5). The resulting concentrates are removed.

Then the pipework and tanks are flushed and dried (7). Readily inflammable substances and gases are removed from the plant (6). After the first partial consent has been obtained (9), a start can be made on the operations subject to consents (8). Next, the remaining operating fluids (reactor coolant, storage pool water, etc.) can be treated and discharged (11). The concentrates thus formed are disposed of. The reactor pressure vessel is sealed with its cover (12). Dose rate measurements (13) are then carried out. At the same time the drain pump installations are converted (14) so that en- try into the reactor building is unnecessary for operation and maintenance of the drain pumps. After the pressure vessel has been sealed with its cover and dose rate measurements have been carried out, the cavity is covered over with the existing concrete slabs (15). All tanks and pipework which have not yet been treated are flushed out and dried with hot air (16). All pipework and electrical cables leading out of the confinement zone, except for necessary ventilation lines and lighting cables, part of the measuring lines of the radiological protection instrumentation, and the drain pump lines, are cut and sealed off (17). On completion of these operations, all areas are decontaminated where necessary. Contamination which cannot readily or economically be removed is fixed with paint to permit subsequent hazard-free access (18). Final dose rate measurements then follow (19).

The crane facilities not only have their electricity supply cut off but are also mechanically locked in position (20). At the same time the entire electrical in- staUation, except for facilities necessary for lighting and inspection, is disconnected (21). All access paths, except that to the controlled areas (reactor and auxiliaries buildings), are closed off and blocked (22). These operations are followed by final acceptance (23) and the issue of the fmal consent (24).

3.2.3. Safety A distinction is made between safety during the

decommissioning operations and safety after completion of decommissioning. While decommissioning is in progress, the general accident prevention requirements and those of the radiation protection regulations must be observed. External safety must be assured. Maintenance of safety calls for detailed planning of sequences of operations and specialized training of the persons responsible for the individual tasks.

The owner of the plant is responsible for the cor- rect execution of the decommissioning according to safety regulations. Nuclear safety is ensured by continuous monitoring of personal doses and local dose rates. Before work is carried out with sources of radioactivity or in the radiation zone, the resulting received doses are estimated and duty rosters are then compiled. The overall radiation load is kept to a minimum by decontamination and movable shields.

3.2.4. Final condition When the interim confinement operations are com-

pleted, safety corresponds at least to that of the plant when it was in service. The outward appearance of the plant is largely unaltered. The condition of the decommissioned plant is monitored by periodic inspections and checks. Any drainage water present is pumped off as necessary and treated by a contractor. A part of the existing remote-reading radiation protection instrumentation is kept in operation. Access to the plant is allowed only after prior monitoring of its condition and clearance by the responsible supervisory staff with radiation monitors and walkie-talkie sets.

3.3. Partial dismantlement with secure residual confinement

This heading includes all ' intermediate alternatives' between interim confinement and total dismantlement. To permit a description of a decommissioning project, a specific alternative has been selected and the appropriate boundary conditions are set out in the following subsections.

3.3.1. Boundary conditions (1) All radioactive parts and materials, except for

resins, slurries and filter elements, are stored in the

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confmement structure of the reactor building. (2) As far as possible, all high-activity parts are

conf'med in the pressure vessel. (3) Decontamination of parts to be removed is

carried out mainly inside the reactor building. (4) The concrete structures are, as far as possible,

to be completely decontaminated. (5) At least the foundations of the building remain. (6) It is assumed that the pressure vessel is not

broken up. (7) The pressure vessel of a BWR power plant is,

as far as possible, lowered. (8) The pressure vessel of a PWR plant remains

in position, since it is usually already at the lowest point in the reactor building.

3.3.2. Sequence o f operations The sequence of operations in partial dismantle-

ment with secure residual confinement is illustrated in fig. 3. The differences in the sequence between PWR and BWR are shown in the flowchart (the figures in brackets refer to the operation numbers in the diagrams).

The general executive planning (1), during which the decommissioning documentation is compiled, is followed by the preparation of a decommissioning safety report (2). General preliminary works (3) which do not affect the existing consents can already be commenced. The documents compiled under items (1) and (2) constitute the basis for the project consent (4). Detailed tenders can be obtained from outside firms; after these have been examined, orders can be placed (5). A start can be made on the development of methods and equipment specifically designed for the decommissioning of the particular plant concerned (6). The next stage is the provision of site infrastructures (7), e.g. fences, temporary site buildings, interim storage facilities for non-active waste, utilities lines, roads, etc.

The detailed planning schedules (10) and partial consent procedures (8) continue in step with the decommissioning operations until the latter are com- pleted. When partial consent has been obtained, a start can be made on the operations subject to consent (9). These include, for example, the provision of additional means of access to the buildings and the removal of non-load-bearing concrete structures which are in the way (masonry walls). Next, parts

which can be easily decontaminated (11) and low- activity liquids which are no longer required (12) are removed. Meanwhile the equipment developed under item (6) is set up and tested (13), and all inactive components no longer required are removed from the future confinement structure (14) to make room for the storage of active components. Now all active fluids are removed after suitable treatment (15). The drained tanks and pipework are flushed out and dried (16).

The following operations are shown separately in the flowchart for the BWR and the PWR. Operations (17)-(20) apply to the BWR, and (21)-(25) to the PWR.

In the BWR, the pressure vessel internals are removed and placed in temporary storage to save weight prior to the lowering of the vessel (17). It depends on individual circumstances whether the pressure vessel has to be lowered (see section 2.2.2), and the relevant decision must be made from case to case.

The active parts of the turbosets are broken up (18), packed and placed in the confinement structure. The free space available in the pressure vessel owing to removal of the fuel elements can be used for the storage of other active parts (e.g. parts of the turbosets). The pressure vessel is then sealed (19), all pipe work being cut and also sealed off. Dose rate measurements are then carried out on the pressure vessel (20).

In the case of a PWR, the pressure vessel head structures are first dismantled (21). They are placed in the pressure vessel together with the vessel internals (22), and the vessel is then sealed. Dose rate measurements are then carried out on the pressure vessel (23). The steam generators are dismantled (24), the active part s being packed and placed in the confinement structure. It is sometimes possible to integrate the steam generators in the confinement structure, so that they need not be dismantled.

The following operations again apply to both the BWR and the PWR.

The remaining active parts (25) are, as far as possible, decontaminated and then removed or placed in storage. Contamination which cannot be economically removed is f'Lxed by spraying over with paint. The pool water is drained and the pool and its pipelines are decontaminated and dried (26). The conf'mement structure is then sealed with a concrete slab (27). The remaining facilities are then dismantled (28) and re-

NIS /LDecommissioning of L WR nuclear power plants 9

moved. Dose rate measurements are carried out in the entire plant (30). If the result is satisfactory, the controlled area zone is officially abolished (31). The remaining concrete structures are removed (32). After radiological acceptance measurements (33), final earthworks are carried out (34). This is followed by final acceptance (35), which constitutes the last Con- dition for the issue of the final consent (36).

3.3.3. Final condition On completion of partial dismantlement with

secure residual confinement, the safety of the plant is ensured by the confinement structure. The integrity of the confinement structure is checked periodic- ally by external inspections and measurements. The confmement structure is so designed as to withstand external agencies such as flood, fire, etc. so that the safe confinement of radioactivity remains assured.

The stored radioactivities are collected together in groups according to their intensity and nature and the condition of the individual parts. Each part is identified from the point of view of its activity and a record is kept of its storage location. This is particularly important with a view to the possibility of subsequent total dismantlement.

Several times a year, surface does rates are measured and soil samples taken for determination of activity.

3.4. Total dismantlement

3.4.1. Boundary conditions (1) A suitable final repository for bulky radioac-

tive waste is available. (2) Decontamination work necessary is carried out

within the controlled area. (3) All equipment and buildings, including founda-

tions, are demolished.

3.4.1.1. Notes on size reduction techniques. This alternative calls for more size reduction work than the alternatives described earlier. For this reason, and also to facilitate comprehension, possible and existing size reduction techniques for the pressure vessel and for concrete are explained below.

Owing to the high level of radiation, remote control is necessary for most of the size reduction operations concerning the pressure vessel and its internals (almost all of the operations concerning the internals

are executed under water). The cutting processes which can be used are both mechanical (sawing, chiselling, etc.) and thermal (flame cutting, electric- arc cutting, and plasma cutting); at present, plasma cutting can only be used for wall thicknesses of up to 170 mm. Particular attention must be devoted to the 'secondary waste' (swarf, smoke, aerosols, etc.) arising from the cutting processes: this must be collected and drawn off by suction. Plasma cutting is capable of being developed to cope with wall thicknesses of up to 500 mm (PWR vessel flange).

There are two main standard techniques for breaking up concrete. In the explosive technique, explosive charges are placed in holes, and these loosen up the entire structure or break it up in layers; demolition in layers can also be carried out hydraulical- ly. Other techniques must be used to cut the steel reinforcement.

The explosive method is relatively expensive and time-consuming. Cutting with oxygen lances is substantially faster, but is accompanied by ifftensive smoke formation. In this process, steel tubes filled with iron wire or iron dust, through which oxygen flows, burn holes in the concrete; the holes should be as close together as possible. The iron filling serves to produce the required temperature. Strong reinforcement is an advantage in this case, as this also burns and generates additional heat. With this technique it is possible to demolish concrete up to 2.5 m thick.

3.4.2. Sequence o f operations The sequence of 'total dismantlement' is illustrated

diagrammatically in fig. 4. The differences between the PWR and the BWR are indicated in the flowchart (the figures in brackets refer to the numbering of the operations in the diagrams).

The general executive planning (1), in which the decommissioning documents are compiled, is followed by the preparation of a decommissioning safety report (2). General preliminary work (3) not affecting the existing consents can already be commenced. The documents prepared in accordance with items (1) and (2) constitute the basis for the project consent (4). Detailed tenders can be obtained from external firms; when these have been examined, orders can be placed (5). A start can be made on the development of methods and equipment specifically designed for the decom-

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ivIS/ Decommissioning o f L WR nuclear power plants 11

missioning of the particular plant concerned (6). The next stage is the provision of the site infrastructures (7), e.g. fences, temporary site buildings, interim storage facilities for nonactive waste, utilities lines, roads, etc.

The detailed planning schedules (10) and partial consent procedures (8) continue in step with the decommissioning operations until the latter are com- pleted. When partial consent has been obtained, a start can be made on the operations subject to consent (9). These include, for example, the provision of additional means of access to the buildings and the removal of non-load-bearing concrete structures which are in the way (masonry walls). Next, parts which can easily be decontaminated (11)and low- level liquids which are no longer required (12) are removed. At the same time the equipment developed in accordance with item (6) is set up and tested (13).

The pressure vessel and its internal structures are dismantled (14) and packed (20). After this, the remaining active parts and liquids are removed (15). In a PWR, the principal parts concerned are the steam generators, which are dismantled and decontaminated (16); in the case of a BWR, the parts concerned are the active components of the turbosets which cannot be decontaminated (17). Tanks and pipework can then be flushed and dried (18). The activated parts of the biological shield are removed (19). All sources of radioactivity are packed in special containers (with shielding if necessary) (20) and can then be transported away (21). Conventional transport facilities are used. The location of the fmal repository deter- mines whether road or rail transport is used.

The pool water is drained off and the pool is dried (22). The reactor building is then decontaminated (23), after which dose rate measurements are carried out (24). If these show that the reactor building is free of activity, the reactor building (RB) controlled area can be abolished (25). Its remaining fixtures are then removed (26) and the building is demolished (27). In parallel with operation 25, the remaining major components are dismantled, decontaminated if necessary, and removed from the site (28). This is followed by total decontamination of the remaining buildings (29); the reactor building has already been decontaminated (23) and its controlled area abolished (25). Dose rate measurements are also carried out on the remaining buildings (30), and the relevant con-

trolled area is abolished (31). After this, the remaining plant installations are removed (32) and all buildings are demolished (33). Demolition of the foundations (34) is followed by final radiological acceptance (35) and concluding earthworks (36). After final acceptance (37), the final consent is issued (38).

3.4.3. Final condition On completion of the decommissioning alternative

of 'total dismantlement', the site can again be used fully without restrictions and is no longer subject to the provisions of nuclear law.

4. Activity inventory

The total activity inventory existing when a nuclear power plant is decommissioned comprises the activity of the structural materials directly exposed to heutron radiation (chiefly the pressure vessel and its internals), the activity of the operating media and corrosion products, and the activity of fission products and fissile materials (see fig. 5).

Active corrosion products, fission products and fissile materials form deposits in the systems of a plant; these constitute contamination. The following contribute to the activity inventory of a nuclear reactor plant:

(1) reactor pressure vessel; (2) biological shield; (3) fuel elements and control rods; (4) contaminated systems and facilities; and (5) radioactive filter materials, resins, slurries, etc.

The movable activities (radioactive filter materials, resins, slurries, etc.) are not taken into account since their handling is part of normal plant operation. The fuel elements are likewise disregarded in the evaluation of the activity inventory. Their removal and treatment are also part of normal operations. They therefore do not form part of the activity inventory to be considered in the decommissioning of a nuclear power plant. Radioactivity due to surface contamination is estimated roughly, because it does not readily lend itself to calculation. The AKAT computer program developed by NIS was used for calculation of the activity inventory of nuclear reactors.

In PWR plants the control rods are integral with


Structural | lAotivated Operating materials activate~ media and corrosion by neutron I r [ products

r a d i a t i o n [ I

1

I [ Fission products ] l~d fissile ~-

t e r i a l s r e l eased

[ I

Ooncentratio 4 ~posits ~posits in operatin~ in in fluids I systems filters

I -

i in~nt cry on decommissioning

Fig. 5. Composition of activity inventory.

the fuel elements; for this reason they too are disregarded in the assessment of the activity inventory.

In BWR plants, on the other hand, the control rods are taken into account. However, their activation can only be determined if the following parameters are known: (a) the thermal neutron flux in the control rod, which

is greatly reduced owing to the absorption in the boron, and

(b) the insertion times and the associated average load factor.

For the boundary conditions for the control rod calculation, see 'BWR activity inventory'.

First of all the activity of the structural materials, due to neutron irradiation, is discussed. This discussion is followed by estimates of contamination and dose rates.

4.1. Calculation of activity inventory of structural materials

The activity A*(Ci) of a component of mass m(g) is calculated by the formula

A * = m ~ A ~ , (1) k~

in which A~ [Ci g-1] stands for the specific activity of the nuclide k. Again,

• ¢ ' g k , (2)

in which

~i = 1.628 × 10 -11 O¢ihfi A1 , (3)

gk = q(1 - e -xkr ) e -xkt (4)

The symbols introduced in eqs. (2)-(4) represent the following parameters (indices: ] = parent element; i = parent isotope; k = radioactive daughter product):

7 i = proportion by mass of the parent element ] in the component material,

Ai = relative atomic mass (atomic weight) (gmol-1), h/i = relative abundance of isotope i in parent element

i , Oei = effective thermal activation cross-section (barn),

= flux density of thermal neutrons (cm -2 s-1), X k = decay constant (y r - l ) , q = load factor (ratio of actual to rated power)

during period of operation, T = duration of period of operation (yr), t = cooling time (yr).

Of the radionuclides occurring in neutron irradiation, those which are excessively short-lived or long- lived can be disregarded owing to their low contribu- tion to activity; the factor e -hkt or (1 - e -x/r/') in eq. (4) is very small with these nuclides. Unless ~i is too large, the following selection criterion can be assumed (Tl/2 = half-life):

4 months ~< Tll2 ~ 104 yr.

,NIS / Decommissioning of L ICR nuclear power plants 13

Table 1. Nuclide data.

Z/ Element .af/ N i hfi Oei,2200 ffi Daughter Typeof (TII2) k Xk[yr -1] Stable product decay (*) daughter

product

26 Fe 55.85 54 0.058 27 Co 58.93 59 1.0 28 Ni 58.71 62 0.037

2.7 2.30(-2) SSFe e 2.96 yr 2.342(-1) SSMn 36.3 5.01 60Co (a) #- 5.28 yr 1.313(-1) 60Ni 15 7.61(-2) 63Ni #- 92 yr 7.534(-3) 63Cu

Values in parentheses are powers of 10. e = capture of an extranuclear electron #- = beta decay (a) = with metastable intermediate state

The data used are summarized in table 1 for the nuclides considered in the following discussion; in this table Z i = atomic number of element; N i = mass number of isotope; and Oci,22o o = capture cross-section for neutrons of velocity 2200 m/see (barn).

It is assumed that Oci = 0.5 X Oci,22o 0 . In determination of the activity inventory; only

n-~ /and n-/3 processes have been taken into account. The nuclides SSFe, 6°Co and 63Ni were examined. The following assumptions were made for calculation of the activity inventory.

Load factor o f 0.75. Owing to inadequate documentation, only the specific activity o f the concrete was determined for the steel and concrete parts of the biological shield.

The breakdown of the pressure vessel components into individual masses is shown in figs. 7 and 8. The distances from the core periphery given relate in most cases to the nearest outside edge of the component to the core. It is thus assumed that the entire length or thickness of the component (unless divided into layers) is exposed to the same neutron flux. The calculation of the activity inventory of the component is thus on the safe side. In addition, in order to deter- mine the dose rate (which is important for handling), it is necessary to know the maximum specific activity of each relevant component, and this can only be determined in this way.

The thermal neutron fluxes considered as boundary conditions in sections4.1.1 and 4.1.2 below are assumed values. Values outside the core have been ex- trapolated from the known fluxes in the core zone, allowance being made for the effect of shielding.

These values may be regarded as representative for nuclear power plants of the 1200 MW class.

The activity inventory was calculated for 17, 25 and 40 years' operation, in each case after cooling times o f 1, 10, 20, 40 and 100 yr. Certain simplifying assumptions were necessary for,calculation of the activity of the individual components. The data necessary for the calculations were in some cases unavail- able or inexact. However, pessimistic values have been used in each case (e.g. upper limit for neutron fluxes). The calculation results for the individual components have a maximum accuracy of within 25%. In the case of components close to the core (higher specific activity), greater accuracy is obtained, owing to the more exact flux data, than in that of components further

Table 2. Materials data on the assumed percentage weights of the elements considered in the activity calculation.

Material Density Elements (g/cm 3) Fe Co Ni

Thermanit 7.8 66.5 0.014 11,1 HS 3 8.38 3 53 3 HS 25 9.15 2 51 10 22 NiMoCr 37 7.86 96.89 0.2 0.8 X 5 CrNi 189 7.8 70.8 0.2 10 X 10 CrNiNb 199 7.8 68.98 0.08 9.5 X 10CrNiNb 189 7.96 68 0.2 * 10 Concrete 2.4 1.1 * * - -

* The following cobalt limRs apply to the internal structures of PWR pressure vessels: 0.08 w/o; for internal structures which are removed during refuelling: 0.022 w/o.

** Assumption for calculation; exact analysis is necessary.

14 CEC / Decommissioning of L WR nuclear power plants

from the core. In the latter, variations by a factor of 2 - 3 are possible, but these have relatively little effect on the total activation inventory.

On commencement of the decommissioning works, it will be necessary to check the calculation results by means of measurements (see table 2).

4.1.1. BWR activity inventory The activity inventory of a substantially represen-

tative 1200 MW BWR (KRB II) was calculated by means of the computer program mentioned earlier, the following boundary conditions being stipulated.

(a) Thermal neutron flux in axial direction, upwards:

core periphery 9 X 1013 cm -2 sec -1 ,

cover 1 X 10 -9 cm -2 sec -1 .

(b) Thermal neutron flux in axial direction, down- wards:

core periphery 4 X 1013 cm -2 s e c - 1 ,

cladding and bottom 1 X 10 - 4 cm -2 sec -1 .

(c) Thermal neutron flux in radial direction:

core periphery 3.5 X 1013 ,

pressure vessel 1.2 X 10 9 ,

biological shield 1.1 X 10 a .

The flux assumed for components situated in each direction between the components mentioned above was interpolated exponentially. Since relatively long distances (approximately 13 m) are involved with the thermal neutron flux in the upward direction, a simple interpolation appeared too inexact. For this reason a shielding calculation in plate geometry was carried out with the MAC-RAD computer code, the existing materials being homogenized in layers. A more detailed description will not be given here. The

10 l&

1012

1010 7 ~ 10 #

0 "-" I0 6

c~ 10/.

~, 10 2

10 0

10 -2

I 0 "¢

10"6

10-0

\ \

0 t O0 200 I ; I i

300 400 500 600 I I I I I I~ ' ~

?oo eoo 9o0 Iooo 12oo If.On

Distance from core periphery (cm)

Fig. 6. Thermal neutron flux of a 1200 MW BWR in the upward axial direction, calculated with the MAC-RAD shielding program.

NIS / Decommissioning o f L WR nuclear power plants

Table 3. Thermal neutron fluxes of the relevant components of the 1200 MW BWR

15

No. Component Volume Density Mass Material Flux (cm 3 ) (gem -3) (kg) (era -2 s -1 )

1 Upper guide grid 0.897 (+6) 7.8 7,000 X5 CrNi 189 9 (+ 13) 2 Emergency cooling spray ring 0.22 (+6) 7.8 1,700 X5 CrNi 189 6 (+ 9) 3 Feedwater distributor 0.41 (+6) 7.8 3,200 X5 CrNi 189 7.5 (+ 1) 4a Cyclone, top 0.642 (+7) 7.8 50,000 X5 CrNi 189 1 ( - 1) 4b Cyclone, bottom 0.321 (+7) 7.8 25,000 X5 CrNi 189 1 (+ 5) 5 Steam drier 0.425 (+7) 7.8 33,200 X5 CrNi 189 1.7 ( - 5) 6 PV top closure 0.145 (+8) 7.86 114,000 22 NLMoCr 37 1 ( - 9) 7a PV cladding, section 1 0.623 (+6) 7.8 4,900 Thermanit 4 ( - 5) 7b PV cladding, section 2 0.36 (+6) 7.8 2,800 Thermanit 4 ( - 1) 7c PV cladding, section 3 0.397 (+6) 7.8 3,100 Thermanit 8 (+ 6) 8a PV, section 1 0.178 (+8) 7.86 140,000 22 NLMoCr 37 4 ( - 5) 8b PV, section 2 0.767 (+7) 7.86 60,300 22 NiMoCr 37 4 ( - 1) 8c PV, section 3 0.86 (+7) 7.86 67,700 22 NiMoCr 37 8 (+ 6) 9 PV cladding, section 4 0.12 (+7) 7.8 9,400 Thermanit 1.4 (+ 9)

10a PV, section 4, inside 0.608 (+7) 7.86 47,800 22 NiMoCr 37 1.2 (+ 9) 10b PV, section 4, centre 0.62 (+7) 7.86 48,500 22 NLMoCr 37 7.3 (+ 8) 10c PV, section 4, outside 0.128 (+8) 7.86 100,600 22 NiMoCr 37 3.4 (+ 8) l l a PV cladding, section 5 0.295 (+6) 7.8 2,300 Thermanit 1.6 (+ 7) l lb PV cladding, bottom 0.34 (+6) 7.8 2,600 Thermanit 2 ( - 1) 12a PV, section 5 0.96 (+7) 7.86 75,500 22 NiMoCr 37 1.6 (+ 7) 12b PV, bottom 0.123 (+8) 7.86 97,000 22 NiMoCr 37 2 ( - 1) 13a Controlrod guide tubes 0.25 (+6) 7.8 1,930 X5 CrNi 189 4 (+13) 13b Control rod guide tubes 0.6 (+6) 7.8 4,655 X5 CrNi 189 8 (+12) 13c Control rod guide tubes 0.138 (+7) 7.8 10,743 X5 CrNi 189 6 (+10) 13d Control rod guide tubes 0.183 (+7) 7.8 14,324 X5 CrNi 189 8 (+ 5) 14a Core barrel, core zone 0.296 (+7) 7.8 23,100 X5 CrNi 189 1 (+13) 14b Core barrel, bottom section 1 0.24 (+7) 7.8 18,750 X5 CrNi 189 1 (+10) 14c Core barrel, bottom section 2 0.24 (+7) 7.8 18,750 X5 CrNi 189 F (+ 2) 15 Lower guide grid 0.1 (+7) 7.8 7,900 X5 CrNi 189 4 (+13) 16 Axialpumps 0.15 (+7) "J.8 12,000 X5 CrNi 189 1 ( - 2) 17 Core fluxmeter housing 0.5 (+6) 7.8 4,000 X5 CrNi189 9 (+13) 18 Biological shield 0.418 2.39 0.001 * Normal concrete 1.1 (+ 8)

Controlrod 0.93 (+4) 7.8 72.5 X5 CrNi 189 1 (+13) Control rod guide rollers, upper 0.835 (+1) 8.38 0.07 HS 3 1 (+13) Control rod guide rollers, lower 0.332 (+2) 8.38 0.278 HS 3 1 (+ 5) Studs, upper 0.984 9.15 0.009 HS 25 1 (+13) Studs, lower 0.546 (+1) 9.15 0.05 HS 25 1 (+ 5)

for calculation of specific activity only

8 (+5) means 8 × 10 s

result of the shielding calculation is shown in fig. 6. The curve coincides well with the data published for KRB-I.

It was assumed that the reactor pressure vessel was made of 22NiMoCr37, with Thermanit cladding (ap- plied by welding), the internals being made of

X5CrNi189 (see alSO table 2).

It was assumed that the control rods had an average life of 4 yr. i.e. all 193 control rods have been re- placed once after 4 yr. The relevant activity inventory on decommissioning was assumed to include one set of control rods irradiated for 4 yr *. The activity was

• . . . .

Pessh'mStlc assumptlon.

16 NIS / Decommissioning of L WR nuclear power plants

1332

,652

415

310

PV Section 4

6) PV top closure

5) Steam drier

4b) Cyclone, top

3) Feedwater distributor

[ 4a) Cyclone, bottom

210

2) Emergency cooling spray rings

80

1) Upper guide grid

0

+.

.,++_i o

+..i++ 230 ] Z

D i s t a n c e f r o m c o r e

periphery (cm) ! For calculation of I specific activity

I

only I

13) Control rod gu%d~ tubes

,

16) Axial D11mD S ....... I_

12b I P¥ bottom lib) Cladding

I 7a Cladding ,

8b I PV Section 2 I 7b Cladding

8ol PV Section3 t 7c Cladding

m

co c o

15) Lower core ~ri~

14) Core barrel- lower section

12a I PV Section 5 lla) Cladding

60g

393

151

I r ' - - ' i I I , - + I

I ~ I o ,~ "~,~ I

i._ ~ I

,, 0

100

-340

184

526

Fig. 7. Diagram of configuration of masses for calculation of the activity inventory of a 1200 MW BWR plant.

first calculated for one control rod on the assumption that it was inserted continuously during the period of operation. The average insertion time was allowed for

by correcting the resulting activity by a factor of 0.73. The neutron fluxes and materials assumed are set out in table 3. It was assumed in the calculation that the

NI8 / Decommissioning of L WR nuclear power plants 17

ent i re c o n t r o l r o d m a s s o f 72 .5 kg c o n s i s t e d o f

X 5 C r N i 1 8 9 . Th i s t akes n o a c c o u n t o f t he activity o f

t he t r i t i u m f o r m e d o n i r r a d i a t i o n o f b o r o n carbide; th is is negligible fo r t h e p u r p o s e s o f t he ac tua l de-

c o m m i s s i o n i n g p roce s s .

The b r e a k d o w n o f the ind iv idua l ma s se s o f t he

r e a c to r p r e s s u r e vessel a n d its i n t e r na l s t r u c t u r e s as

well as the i r d i s tances f r o m a n d o r i e n t a t i o n w i t h re-

spec t to t he p e r i p h e r y o f t he co re , are i l lus t ra ted in

fig. 7. F o r th is p u r p o s e the PV is d iv ided i n t o a n u m -

Table 4.

Specific activities Ak(Ci/g - t ) of a 1200 MW BWR after 17 years' operation and cooling times of 1, 10, 20, 40 and 100 years.

No. Component 1 year 10 years 20 years 40 years 100 years

1 Upper guide grid 1.4+ 0 3 . 2 - 1 1 . 1 - 1 4 . 9 - 2 2 . 9 - 2 2 Emergency cooling spray rings 9 . 6 - 5 2 , 2 - 5 7 . 1 - 6 3 . 3 - 6 1 . 9 - 6 3 Feedwater distributor 1.2 - 12 2.7 - 13 8.9 - 14 4.1 - 14 2.4 - 14

4a Cyclone, top 1.6 - 15 3.6 - 16 1.2 - 16 5.4 - 17 3.2 - 17 4b Cyclone, bo t tom 1.6 - 9 3.6 - 10 1.2 - 10 5.4 - 11 3.2 - 11 5 Steam drier 2.7 - 19 6.1 - 20 2.0 - 20 9.2 - 21 5.5 - 21

6 PV top closure 3.4 - 23 6.2 - 24 1.3 - 24 1.6 - 25 5.8 - 26 7a PV cladding, section 1 4.0 - 19 7.7 - 20 3.2 - 20 2.3 - 20 1.4 - 20 7b PV cladding, section 2 4.0 - 15 7.7 - 16 3.2 - 16 2.3 - 16 1.4 - 16

7c PVcladding, sect ion3 8 . 1 - 8 1 . 5 - 8 6 , 3 - 9 4 . 5 - 9 2 . 9 - 9 8a PV, section 1 7.6 - 19 1.4 - 19 2.8 - 20 3.5 - 21 1.3 - 21 8b PV, section 2 7.6 - 15 1.4 - 15 2.8 - 16 3.5 - 17 1.3 - 17

8c PV, section 3 1 . 5 - 7 2 . 7 - 8 5 . 6 - 9 6 . 9 - 10 2 . 6 - 10 9 PVcladding, section4 1 . 4 - 5 2 . 7 - 6 1 . 1 - 6 7 . 9 - 7 5 . 0 - 7

lOa PV, section4, inside 2 . 3 - 5 4 . 1 - 6 8 . 3 - 7 1 . 0 - 7 3 . 9 - 8

10b PV, sect ion4, centre 1 . 4 - 5 2 . 5 - 6 5 . 1 - 7 6 . 4 - 8 2 . 3 - 8 10c PV, sect ion4, outside 6 . 4 - 6 1 . 2 - 6 2 . 4 - 7 2 . 9 - 8 1 . 1 - 8 11a PVcladding, sect ion5 1 . 6 - 7 3 . 1 - 8 1 . 3 - 8 9 . 1 - 9 5 . 7 - 9

l l b PV cladding, bo t tom 2.0 - 15 3.8 - 16 1.6 - 16 1.1 - 16 7.1 - 17 12a PV, sect ion5 3 . 0 - 7 5 . 5 - 8 1 . 1 - 8 1 . 4 - 9 5 . 1 - 1 0 12b PV, bo t tom 3.8 - 15 6.9 - 16 1.4 - 16 1.7 - 17 6 .4 . - 18

13a Control rod guide tubes 6 . 4 - 1 1 . 4 - 1 4 . 7 - 2 2 . 1 - 2 1 . 3 - 2 ' 13b Control rod guide tubes 1 . 3 - 1 2 . 9 - 2 9 . 5 - 3 4 . 3 - 3 2 . 6 - 3 13c Control rodgu ide tubes 9 . 6 - 4 2 . 1 - 4 7 . 1 - 5 3 . 2 - 5 1 . 9 - 5

13d Control rod guide tubes 1.3 - 8 2.9 - 9 9.5 - 10 4.3 - 10 2.6 - 10 14a Core barrel, eorezone 1 . 6 - 1 3 . 6 - 2 1 . 2 - 2 5 . 4 - 3 3 . 2 - 3 14b Core barrel, bo t tom section 1 1 . 6 - 4 3 . 6 - 5 1 . 2 - 5 5.4 = 6 3 . 2 - 6

14c Core barrel, bo t tom section 2 1.6 - 12 3.6 - 13 1.2 - 13 5.4 - 14 3.2 - 14 15 Lower guidegrid 6 . 4 - 1 1 . 4 - 1 4 . 7 - 2 2 . 2 - 2 1 . 3 - 2 16 Axial pumps 1.6 - 16 3.6 - 17 1.2 - 17 5.4 - 18 3.2 - 18

17 Core fluxmeter housing tubes 1.4+ 0 3.2 - 1 1.1 - 1 4.9 - 2 2.9 - 2 18 Biological shield 1.6 - 8 1.9 - 9 1.9 - 10 1.7 - 12 1.4 - 18

2- 1 - 2 means 2 • 1 • 10 -2


Table 5. Activities (Ci) of a 1200 MW BWR after 17 years' operation and cooling times of 1, 10, 20, 40 and 100 years.


1 Upper guide grid 1.0+ 7 2.3+ 6 7.5+ 5 3.4+ 5 2.0+ 5 2 Emergency cooling spray rings 1.6 + 2 3.7 + 1 1.2 + 1 5.6 + 0 3.2 + 0 3 Feedwaterdistributor 3 . 8 - 6 8 . 6 - 7 2 . 8 - 7 1 . 3 - 7 7 . 7 - 8

4a Cyclone, top 8 . 0 - 8 1 . 8 - 8 5 . 9 - 9 2 . 7 - 9 1 . 6 - 9 4b Cyclone, bottom 4 . 0 - 2 8 . 9 - 3 2 . 9 - 3 1 . 3 - 3 8 . 1 - 4

5 Steam drier 9 . 1 - 1 2 2 - 1 2 6 . 7 - 1 3 3 . 1 - 1 3 1 . 8 - 1 3


7c PVcladding, section3 2 . 5 - 1 4 . 7 - 2 1 . 9 - 2 1 . 4 - 2 8 . 9 - 3 8a PV, section 1 1.1 - 10 1.9 - 11 3.9 - 12 4.9 - 1 3 1.8 - 13 8b PV, section2 4 . 5 - 7 8 . 3 - 8 1 . 7 - 8 2 . 1 - 9 7 . 8 - 10

8c PC, section3 1.0+ 1 1.8+ 0 3 . 7 - 1 4 . 7 - 2 1 . 7 - 2 9 PVcladding, section4 1.3+ 2 2.5+ 1 1.0+ 1 7.4+ 0 4.7+ 0

lOa PV, section4, inside 1.1 + 3 1.9 + 2 3.9 + 1 5.0+ 0 1.8 + 0

10b PV, section4, centre 6.7+ 2 1.2+ 2 2.4+ 1 3.1+ 0 1.1+ 0 10c PV, section4, outside 6.5 + 2 1.1 + 2 2.3 + 1 2.9+ 0 1.1 + 0 l l a PVcladding, section5 3 . 7 - 1 7 . 0 - 2 2 . 9 - 2 2 . 1 - 2 1 . 3 - 2

l i b PV cladding, bottom 5.2 - 9 9.9 - 10 4.1 - 10 2.9 - 10 1.8 - 10 12a PV, section5 2.3+ 1 4.1+ 0 8 . 4 - 1 1 . 1 - i 3 . 9 - 2 12b PV, bottom 3 . 7 - 7 6 . 7 - 8 1 . 3 - 8 1 . 7 - 9 6 . 3 - 10

13a Control rod guide tubes 1.2+ 6 2.7+ 5 9.1+ 4 4.2+ 4 2.5+ 4 13b Control rod guide tubes 5.9+ 6 1.3+ 5 4.4+ 4 2.0+ 4 1.2+ 4 13c Control rod guide tubes 1.0+ 4 2.3 + 3 7.6 + 2 3.5 + 2 2.1+ 2

13d Control rod guide tubes 1 . 8 - 1 4 . 1 - 2 1 . 3 - 2 6 . 2 - 3 3 . 7 - 3 14a Core barrel, core zone 3.7 + 6 8.3 + 5 2.7 + 5 1.2 + 5 7.4+ 4 14b Core barrel, bottom section 1 3.0+ 3 6.7 + 2 2.2 + 2 1.0+ 2 6.0+ 1

14c Core barrel, bottom section 2 3 . 0 - 5 6 . 7 - 6 2 . 2 - 6 1 . 0 - 6 6 . 0 - 7 15 Lower guide grid 5.0+ 6 1.1+ 6 3.7+ 5 1.7+ 5 1.0+ 5 16 Axial pumps 1.8 - 9 4.2 - 10 1.4 - 10 6.3 - 11 3.8 - 11

17 Core fluxmeter housing tubes 5.6+T 6 1.2 + 6 4.2 + 5 1.9+ 5 1.1 + 5

Total 2.6+ 7 5.8+ 6 1.9+ 6 8.8+ 5 5.2+ 5

2.1 - 2 means 2.1 × 10 -2

ber o f sect ions. Sect ions 1 to 3 are above the core,

sect ion 4 is s i tua ted radially wi th respect to the core,

and sec t ion 5 is be low the core. For sect ions 1, 2, 3 and

5, the assumed flux was tha t prevailing at the appropfi-

ate axial dis tance b e t w e e n the neares t edge to the core

and the pe r iphery o f the core , i.e. axial planes o f equal

flux were assumed to exist . The con t ro l rods are no t

shown in fig. 7 (see tables 3 - 1 0 ) .

NIS / Decommissioning of L WR nuclear power plants 19

Table 6 Specific activities A k(Ci/g - 1 ) of a 1200 MW BWR after 25 years' operation and cooling times of 1, 10, 20, 40 and 100 years.


1 Upper guide grid 1.5+ 0 3 . 6 - 1 1 . 3 - 1 6 . 9 - 2 4 . 1 - 2 2 Emergency cooling spray rings 1 . 0 - 4 2 . 4 - 5 8 . 9 - 6 4 . 6 - 6 2 . 7 - 6 3 Feedwater distributor 1.3 - 12 3.0 - 13 1.1 - 13 5.7 - 14 3.4 - 14

4a Cyclone, top 1.7 - 15 4.0 - 16 1.5 - 16 7.6 - 1 7 4.6 - 17 4b Cyclone, bo t tom 1.7 - 9 4.0 - 10 1.5 - 10 7.6 - 11 4.6 - 11 5 Steam drier 2.9 - 19 6.8 - 20 2.5 - 20 1.3 - 20 7.8 - 21


7c PVcladding, sect ion3 8 . 4 - 8 1 . 8 - 8 8 . 6 - 9 6 . 5 - 9 4 . 1 - 9 8a PV, section 1 7.8 - 19 1.4 - 19 3.0 - 20 4.5 - 21 1.8 - 21 8b PV, section 2 7.8 - 15 1.4 - 15 3.0 - 16 4.5 - 17 1.8 - 17

8c PV, seetion 3 1 . 6 - 7 2 . 9 - 8 6 . 1 - 9 8 . 9 - 1 0 3 . 7 - 10 9 PVcladding, seet ion4 1 . 5 - 6 3 . 1 - 6 1 . 5 - 6 1 . 1 - 6 7 . 1 - 7

10a PV, sect ion4, inside 2 . 3 - 5 4 . 3 - 6 9 . 1 - 7 1 . 3 - 7 5 . 5 - 8

10b PV, sect ion4, centre 1 . 4 - 5 2 . 6 - 6 5 . 5 - 7 8 . 1 - 8 3 . 4 - 8 10c PV, sect ion4, outside 6 . 7 - 6 1 . 2 - 6 2 . 6 - 7 3 . 8 - 8 1 . 6 - 8 l l a PVeiadding, section5 1 . 7 - 7 3 . 6 - 8 1 . 7 - 8 1 . 3 - 8 8 . 2 - 9

l l b PV cladding, bo t tom 2.1 - 15 4.8 - 16 2.1 - 16 1.6 - 16 1 . 0 - 16 12a PV, sect ion5 3 . 1 - 7 5 . 8 - 8 1 . 2 - 8 1 . 8 - 9 7 . 3 - 1 0 12b PV, bo t tom 3.9 - 15 7.2 - 16 1.5 - 16 2.2 - 17 9.2 - 18

13a Control rodgu ide tubes 6 . 8 - 1 1 . 6 - 1 5 . 9 - 2 3 . 0 - 2 1 . 8 - 2 13b Control rod guide tubes 1 . 3 - 1 3 . 2 - 2 1 . 2 - 2 6 . 1 - 3 3 . 7 - 3 13c Control rod guide tubes 1 . 0 - 3 2 . 4 - 4 8 . 9 - 5 4 . 6 - 5 2 . 7 - 5

13d Control rod guide tubes 1.3 - 8 3.2 - 9 1.2 - 9 6.1 - 10 3.7 - t0 14a Core barrel, corezone 1 . 7 - 1 4 . 0 - 2 1 . 5 - 2 7 . 6 - 3 4 . 6 - 3 14b Core barrel, bo t tom section l 1 . 7 - 4 4 . 0 - 5 1 . 5 - 5 7 . 6 - 6 4 . 6 - 6

14c Core barrel, bo t tom section 2 1.7 - 12 4.0 - 13 1.5 - 13 7.6 - 14 4.6 - 14 15 Lower guide grid 6 . 8 - 1 1 . 6 - 1 5 . 9 - 2 3 . 0 - 2 1 . 8 - 2 16 Axial pumps 1.7 - 16 4.0 - 17 1.5 - 17 7.6 - 18 4.6 - 18

17 Core fluxmeter housing tubes 1.5 + 0 3.6 - 1 1.3 - 1 6.9 - 2 4.1 - 2 18 Biological shield 1.6 - 8 1.9 - 9 1.9 - 10 1.7 - 12 1.4 - 18

2.1 - 2 means 2.1 x 10 - 2

4.1.2. PWR activity inventory The ac t iv i ty i n v e n t o r y o f a s u b s t a n t i a l l y r ep re sen t a -

tive 1200 MW P W R (Bibl is A) was ca l cu la t ed w i t h t h e

c o m p u t e r p r o g r a m m e n t i o n e d earl ier , t h e f o l l o w i n g

b o u n d a r y c o n d i t i o n s b e i n g s t i p u l a t e d .

(a) T h e r m a l n e u t r o n f l ux in axial d i r ec t ion , u p w a r d s :

core p e r i p h e r y 9 X 1011 c m - 2 s e e - l ,

t o p c losu re 1 X 102 c m - 2 see - 1 .

( b ) T h e r m a l n e u t r o n f l ux in axial d i r ec t ion , d o w n -

20 NIS / Decommissioning of L teR nuclear power plan ts

Table 7.

Activities (Ci) of a 1200 MW BWR after 25 years' operation and cooling times of 1, 10, 20, 40 and 100 years.


1 Upper guide grid 1.1+ 7 2 .5+ 6 9 .3+ 5 4 .8+ 5 2 .9+ 5 2 Emergency cooling spray rings 1.7 + 2 4.1 + 1 1.5 + 1 7.8 + 0 4.7 + 0 3 Feedwaterdistributor 4 . 0 - 6 9 . 6 - 7 3 . 5 - 7 1 . 8 - 7 1 . 1 - 7

4a 4b 5

6 7a 7b

7c 8a 8b

8c 9

10a

10b 10c l l a

l l b 12a 12b

13a 13b 13c

13d 14a 14b

14c 15 16

17

Cyclone, top 8 . 5 - 8 2 . 0 - 8 7 . 4 - 9 3 . 8 - 9 2 . 3 - 9 Cyclone, bo t tom 4 . 2 - 2 1 . 0 - 2 3 . 7 - 3 1 . 9 - 3 1 . 1 - 3 Steam drier 9.5 - 12 2.3 - 12 8.3 - 13 4.3 - 13 2.6 - 13

PV top closure 4.0 - 15 7.4 - 16 1.5 - 16 2.3 - 17 9.4 - 18 PV cladding, section 1 2.1 - 12 4.4 - 13 2.1 - 13 1.6 - 13 1.0 - 13 PV cladding, section 2 1.2 - 8 2.5 - 9 1.2 - 9 9.1 - 10 5.7 - 10

PVcladding, section3 2 . 6 - 1 5 . 5 - 2 2 . 6 - 2 2 . 0 - 2 1 . 3 - 2 PV, section 1 1.1 - 10 2.0 - 11 4.2 - 12 6.3 - 13 2.6 - 13 PV, section2 4 . 7 - 7 8 . 7 - 8 1 . 8 - 8 2 . 7 - 9 1 . 1 - 9

PV, section3 1.0+ 1 1.9+ 0 4 . 1 - 1 6 . 0 - 2 2 . 5 - 2 PVcladding, section4 1.4+ 2 2 .9+ 1 1.4+ 1 1.0+ 1 6 .7+ 0 PV, section4, inside 1.1+ 3 2 .1+ 2 4 .4+ 1 6 .4+ 0 2 .6+ 0

PV, section4, centre 6 .9+ 2 1.3+ 2 2 .7+ 1 3.9+ 0 1.6+ 0 PC, section 4, outside 6 .7+ 2 1.2+ 2 2 .6+ 1 3.8+ 0 1.6+ 0 PVcladding, section5 3 . 9 - 1 8 . 2 - 2 3 . 9 - 2 2 . 9 - 2 1 . 9 - 2

PV cladding, bo t tom 5.5 - 9 1.1 - 9 5.6 - 10 4.2 - 10 2.6 - 10 PV, section5 2 .3+ 1 4 .4+ 0 9 . 2 - 1 1 . 3 - 1 3 . 5 - 2 PV, bot tom 1 . 0 - 8 1.9 - 9 3.9 - 10 5.8 - 11 2.4 - 11

Control rod guide tubes 1.3 + 6 3.1 + 5 1.1 + 5 5.9 + 4 3.5 + 4 Control rod guide tubes 6 .3+ 5 1.5+ 5 5 .5+ 4 2 .8+ 4 1.7+ 4 Control rod guide tubes 1.1+ 4 2 .6+ 3 9 .5+ 2 4 .9+ 2 2 .9+ 2

Control rodgu ide tubes 1 . 9 - 1 4 . 6 - 2 1 . 7 - 2 8 . 7 - 3 5 . 3 - 3 Core barrel, core zone 3.9+ 6 9 .3+ 5 3 .4+ 5 1.7+ 5 1.1+ 5 Core barrel, bot tom sec t ion l 3.2+ 3 7 .5+ 2 2 .7+ 2 1.4+ 2 8 .6+ 1

Core barrel, bo t tom section 2 3 . 2 - 5 7 . 5 - 6 2 . 7 - 6 1 . 4 - 6 8 . 6 - 7 Lower guide grid 5 .3+ 6 1.2+ 6 4 .6+ 5 2.4+ 5 1.4+ 5 Axial pumps 1.9 - 9 4.7 - 10 1.7 - 10 8.9 - 11 5.4 - 11

Core fluxmeter housing tubes 5.9 + 6 1.4+ 6 5.2 + 5 2.7 + 5 1.6+ 5

Total 2 .8+ 7 6 .5+ 6 2 .4+ 6 1.2+ 6 7 .5+ 5

2.1 - 2 means 2.1 X 10 - 2

wards :

co re p e r i p h e r y

b o t t o m or zone a l

9 × 1011cm - 2 sec - t ,

5 X 106 c m - 2 sec - 1 .

(c) T h e r m a l n e u t r o n f lux in radial d i r ec t ion :

core p e r i p h e r y 1.2 X 10 la,

p r e s su re vessel o r z o n e b 5 X 109 .


Table 8. Specific activities Ak(Ci/g - 1 ) of a 1200 MW BWR after 40 years' operation and cooling times of 1, 10, 20, 40 and 100 years.


1 Upper guide grid 1.6+ 0 4 . 1 - 1 1 . 7 - 1 1 . 0 - 1 6 . 3 - 2 2 Emergency cool ingsprayrings 1 . 0 - 4 2 . 7 - 5 1 . 1 - 5 6 . 8 - 6 4 . 2 - 6 3 Feedwater distributor 1.3 - 12 3.4 - 13 1.4 - 13 8.5 - 14 5.2 - 14

4a Cyclone, top 1.8 - 15 4.5 - 16 1.9 - 16 1.1 - 16 6.9 - 17 4b Cyclone, bo t tom 1.8 - 9 4.5 - 10 1.9 - 10 1.1 - 10 6.9 - 11 5 Steam drier 3.0 - 19 7.7 - 20 3.2 - 20 1.9 - 20 1.2 - 20


7c PVcladding, sect ion3 8 . 9 - 8 2 . 2 - 8 1 . 2 - 8 9 . 7 - 9 6 . 2 - 9 8a PV, section 1 7.9 - 19 1.5 - 19 3.3 - 20 6.0 - 21 2.8 - 21 8b PV, section 2 7.9 - 15 1.5 - 15 3.3 - 16 6.0 - 17 2.8 - 17

8c PV, sect ion3 1 . 6 - 7 3 . 0 - 8 6 . 6 - 9 1 . 2 - 9 5 . 6 - 10 9 PVcladding, sect ion4 1 . 5 - 5 3 . 9 - 6 2 . 2 - 6 1 . 7 - 6 1 . 1 - 6

10a PV, section4, inside 2 . 4 - 5 4 . 5 - 6 9 . 9 - 7 1 . 8 - 7 8 . 4 - 8

10b PV, section4, centre 1 . 4 - 5 2 . 7 - 6 6 . 0 - 7 1 . 1 - 7 5 . 1 - 8 10c PV, section4, outside 6 . 8 - 6 1 . 2 - 6 2 . 8 - 7 5 . 1 - 8 2 . 4 - 8 l l a PVcladding, section5 1 . 8 - ~ 4 . 4 - 8 2 . 5 - 8 1 . 9 - 8 1 . 2 - 8

l l b PV cladding, bo t tom 2.2 - 15 5.5 - 16 3.1 - 16 2.4 - 16 1.5 - 16 12a PV, sect ion5 3 . 2 - 7 6 . 0 - 8 1 . 3 - 8 2 . 4 - 9 1 . 1 - 9 12b PV, bo t tom 3.9 - 15 7.5 - 16 1.6 - 16 3.0 - 17 1.4 - 17

13a Control rod guide tubes 7.1 - 1 1.8 - 1 7.7 - 2 4.5 - 2 2.8 - 2 13b Control rod guide tubes 1 . 4 - 1 3 . 6 - 2 1 . 5 - 2 9 . 1 - 3 5 . 6 - 3 13c Control rod guide tubes 1 . 0 - 3 2 . 7 - 4 1 . 1 - 4 6 . 8 - 5 4 . 2 - 5

13d Control rod guide tubes 1.4 - 8 3.6 - 9 1.5 - 9 9.1 - 10 5.6 - 10 14a Core barrel, corezone 1 . 8 - 1 4 . 5 - 2 1 . 9 - 2 1 . 1 - 2 6 . 9 - 3 14b Core barrel, bo t tom section 1 1 . 8 - 4 4 . 5 - 5 1 . 9 - 5 1 . 1 - 5 6 . 9 - 6

14c Core barrel, bo t tom section 2 1.8 - 12 4.5 - 13 1.9 - 13 1.1 - 13 6.9 - 14 15 Lower guidegrid 7 . 1 - 1 1 . 8 - 1 7 . 7 - 2 4 . 5 - 2 2 . 8 - 2 16 Axial pumps 1.8 - 16 4.5 - 17 1.9 - 17 1.1 - 17 6.9 - 18

17 Coref luxmeter housing tubes 1.6 + 0 4.1 - 1 1.7 - 1 1.0 - 1 6.3 - 2 18 Biological shield 1.6 - 8 2.0 - 9 1 . 9 - 1 0 1 . 8 - 12 1.4 - 18

2.1 - 2 means 2.1 × 10 - 2

T h e f l ux a s s u m e d for c o m p o n e n t s s i t u a t e d in t h e

a p p r o p r i a t e d i r ec t i ons b e t w e e n t h e c o m p o n e n t s m e n -

t i o n e d a b o v e was i n t e r p o l a t e d e x p o n e n t i a l l y . T h e ma-

terials a s s u m e d w e r e 2 2 N i M o C r 3 7 fo r the r e a c t o r

p r e s su re vessel , X 1 0 C r N i N b 1 9 9 fo r t he c ladd ing , a n d

X 1 0 C r N i N b 189 fo r t he i n t e r na l s t r u c t u r e s (see a lso

tab le 2) .

The b r e a k d o w n o f t he ind iv idua l mas se s o f t h e

r e a c t o r p r e s s u r e vessel and its in te rna l s , as well as the i r

d i s tances f r o m and o r i e n t a t i o n s w i t h r e spec t t o t he

22 NIS / Decommissioning o f L ICR nuclear power plants

Table 9. Activities (Ci) of a 1200 MW BWR after 40 years' operation and cooling times of 1, 10, 20, 40 and 100 years.


1 Upper guide grid 1.1 + 7 2.9 + 6 2 Emergency cooling spray rings 1.8 + 2 4.7 + 1 3 Feedwaterdistributor 4 . 2 - 6 1 . 1 - 6

4a Cyclone, top 8 . 8 - 8 2 . 3 - 8 4b Cyclone, bottom 4 . 4 - 2 1 . 1 - 2 5 Steam drier 9.9 - 12 2.6 - 12

6 PV top closure 4.1 - 15 7.7 - 16 7a PV cladding, section 1 2.2 - 12 5.4 - 13 7b PV cladding, section 2 1.2 - 8 3.1 - 9

7c PV cladding, section 3 2.7 - 1 6.8 - 2 8a PV, section 1 1.1 - 10 2.1 - 11 8b PV, section 2 4.8 - 7 9.0 - 8

8c PV, section 3 1.1 + 1 2.0+ 0 9 PV cladding, section 4 1.4 + 2 3.6 + 1

10a PV, section 4, inside 1.1 + 3 2.1 + 2

10b PV, section 4, centre 7.0 + 2 1.3 + 2 10c PV, section 4, outside 6.8 + 2 1.2 + 2 l l a PV cladding, section 5 6.8 + 2 1.2 + 2

l l b PV cladding, bottom 5.8 - 9 1.4 - 9 12a PV, section5 2.4+ 1 4.5 + 0 12b PV, bottom 3 . 8 - 7 7 . 3 - 8

13a Control rod guide tubes 1.3 + 6 3.5 + 5 13b Control rod guide tubes 6.6 + 5 1.7 + 5 13c Control rod guide tubes 1.1 + 4 2.9 + 3

13d Control rod guide tubes 2.0 - 1 5.2 - 2 14a Core barrel, core zone 4.1 + 6 1.0 + 6 14b Core barrel, bottom section 1 3.3 + 3 8.5 + 2

14c Core barrel, bottom section 2 3.3 - 5 8.5 - 6 15 Lower guide grid 5.5 + 6 1.4 + 6 16 Axialpumps 2.1 - 9 5.3 - 10

17 Core fluxmeter housing tubes 6.2 + 6 1.6 + 6

Total 2.9+ 7 7.4+ 6

1.2+ 6 7.2+ 5 4.4+ 5 1.9+ 1 1.2+ 1 7.2+ 0 4 . 6 - 7 2 . 7 - 7 1 . 7 - 7

9 . 6 - 9 5 . 7 - 9 3 . 5 - 9 4 . 8 - 3 2 . 8 - 3 1 . 7 - 3 1 . 1 - 1 2 6 . 4 - 1 3 3 . 9 - 1 3

1 . 7 - 1 6 3 . 1 - 1 7 1 . 4 - 1 7 3 . 1 - 1 3 2 . 4 - 1 3 1 . 5 - 1 3 1 . 7 - 9 1 . 4 - 9 8 . 7 - 1 0

3 . 8 - 2 3 . 0 - 2 1 . 9 - 2 4 . 6 - 1 2 8 . 4 - 1 3 3 . 9 - 1 3 1 . 9 - 8 3 . 6 - 9 1 . 7 - 9

4 . 4 - 1 8 . 1 - 2 3 . 8 - 2 2.0+ 1 1.6+ 1 1.0+ 1 4.7+ 1 8.6+ 0 4.0+ 0

2.9+ 1 5.3+ 0 2.5+ 0 2.8+ 1 5.1+ 0 2.4+ 0 2.8+ 1 5.1+ 0 2.4+ 0

8 . 1 - 1 0 6 . 3 - 1 0 4 . 0 - 1 0 9 . 9 - 1 1 . 8 - 1 8 . 4 - 2 1 . 6 - 8 2 . 9 - 9 1 . 3 - 9

1.5+ 5 8.8+ 4 5.4+ 4 7.2+ 4 4.2+ 4 2.6+ 5 1.2+ 3 7.3+ 2 4.5+ 2

2 . 2 - 2 1 . 3 - 2 8 . 0 - 3 4.4+ 5 2.6+ 5 1.6+ 5 3.6+ 2 2.3+ 2 1.3+ 2

3 . 6 - 6 2 . 3 - 6 1 . 3 - 6 6.0+ 5 3.5+ 5 2.1+ 5 2 . 2 - 1 0 1 . 3 - 1 0 8 . 2 - 1 1

6 .8+ 5 3.9+ 5 2 .4+ 5

3.1+ 6 1.8+ 6 1.3+ 6

2.1 - 2 means 2.1 x 10 -2

core per iphery , is i l lustrated in fig. 8. The pressure

vessel was divided for this purpose in to zones a l and

a2 be low the core , zone b radial to the core , and zones

c l , c2 and c3 above the core. In fig. 8 these masses

are s h o w n to the lef t o f the cen t re line, whilst the inter.

nals are to be found on the right. Internal s t ruc tures

o f X10CrNiNb 189 no t o therwise d e f i n e d were desig-

na ted ' uppe r and lower residual mass ' and al lowed for


Table 10. Activity of the control rods of a 1200 MW BWR. Specific activity [Ci/g] of a control rod after 4 years' irradiation in a 1200 MW BWR after the following cooling times:

Components 1 year 10 years 20 years 40 years 100 years

Control rod 6.3 - 2 1.2 - 2 3.2 - 3 1.0 - 3 5.8 - 4 Control rod guide rollers: upper 5.2+ 0 1.6+ 0 4 . 3 - 1 3 . 1 - 2 1 . 8 - 4 lower 5.2 - 8 1.6 - 8 4.3 - 9 3.1 - 10 1.8 - 12 Studs for control rod guide rollers: upper 5.0+ 0 1.5+ 0 4 . 1 - 1 3 . 1 - 2 5 . 9 - 4 lower 5.0 - 8 1.5 - 8 4.1 - 9 3.1 - 10 5.9 - 12

Activity [Ci] of a control rod after 4 years' irradiation in a 1200 MW BWR after the following cooling times:

Component 1 year 10 years 20 years 40 years

Control rod 4.6 + 3 8.9 + 2 2.3 + 2 7.5 + 1 Control rod guide rollers: upper 3.6 + 2 1.1 + 2 3.0+ 1 2.2 + 0 lower 1.4 - 5 4.4 - 6 1.2 - 6 8.7 - 8 Studs for control rod guide rollers: upper 4.5 + 1 1.4+ 1 3.7 + 0 2.8 - 1 lower 2.5 - 6 7.7 - 7 2.1 - 7 1.5 - 8

Tot~ 5 . 0 + 3 1 . 0 + 3 2 . 6 + 2 7.7+ 1

Activity [Ci] of control rods of a 1200 MW BWR after 4 years' irradiation after the following cooling times:

1 year 10 years 20 years 40 years

9 . 7 + 5 1.9+5 5 . 1 + 4 1.5+ 4

100 years

4.2 + 1

1 . 3 - 2 5.2 - 10

5 . 3 - 3 2.9 - 10

4.2 + 1

100 years

8 . 1 + 3

2.1 - 2 means 2.1 X 1 0 - 2

in the calcula t ion o f the act ivi ty inventory ; the calcu-

lated activities are, however , negligible compared wi th

the total inventory. In te rna ls wi th high cobal t con ten t s

(studs o f high wear resistance wi th Co con ten t s ex-

ceeding 50%), on the o ther hand, were considered sepa-

rately. For these structures, especially in the case o f

componen t s close to the core and exposed to a high

neu t ron flux, the specific activities proved to be very

high. However , 6°Co decays relatively qu ick ly in comparison with the to ta l act ivi ty, for which 63Ni is part icu-

larly impor t an t (see tables 1 1 - 1 7 ) .

4.1.3. SUmmary The results o f the tables set ou t in the preceding

pages are presented graphically in fig. 9. The reactor pressure vessel componen t s will n o w

be divided into act ivi ty groups (low-level, medium. level

and high-level), for the example o f 40 years ' opera t ion

and 1 year 's and 40 years ' cool ing t ime, separate data

being given for the BWR and the PWR, using the fol-

lowing classification [1]:

low-level: < 10 - l Ci m -a ,

medium-level : 1 0 - 1 - 1 0 3 Ci m - 3 ,

high-level: > 103 Ci m - a .

This classification applies only for the purpose o f

handl ing the active componen t s (see fig. 10).

4.2. Estimation o f contaminated inventory

A pract ical calcula t ion o f the con tamina ted in-

ventory is no t possible because this quan t i ty involves

many parameters whose magni tudes and effects can-


PV zone b

296 [

196,8 I

98A !

0

I

I~ P"

0

133.8

7)PV top closure

6c)PV zone C 3

3c)Cladding

6blPV zone C 2 3b)Cladding

6a)PV zone C I

3a)CladdLug

I 4b)PV zone a2

Ib)Cladding

4a)PV zone al la)Cladding

91Cover plate ! 19)Guide pieces ~O)Guide blocks

8)Top grid " I

lO)Supports [

13)Upper residual mas4

ll)Grid plate 12~Centring pieces 21~Guide pieces 222Guide pins 2~)Centring studs

29.2.5

270

180

149

0

16)Lower grid I

24)Brackets ,I

17)Lower residual mas~

18)Stool I

0

0

~D

A

uD t~

45

9.2

105

Distance f rom core periphery cm

Fig. 8. Diagram of configuration of masses for calculation of the aeti~ty inventory of a 1220 MW PRW plant.

not be determined sufficiently precisely. For this reason it is necessary to estimate the contaminated inventory.

Various values are available as the starting point for an estimate of the contaminated inventory of a

1200 MW LWR. Measurement of dose rates at the KWO station after 5 years' operation showed a specific activity of 200 -800 taCi/cm 2 for the internal zones of the primary water line, only 6°Co being considered. Applying the lower limit of these measure-

Table 11. 25 Thermal neutron fluxes of the relevant components of the 1200 MW PWR.

No. Component Volume Density Mass Material Flux cm 3 gcm - 3 Mg cm-2s -1

la PV cladding, zone al 128,000 7.8 1 lb PV cladding, zone a2 372,000 7.8 2.9 2 PV cladding, zone b 462,000 7.8 3.6 3a PV cladding, zone cl 94,900 7.8 0.74 3b PV cladding, zone c2 94,900 7.8 0.74 3c PV cladding, zone c3 94,900 7.8 0.74 4a PV zone al 3,242,000 7.96 25.5 4b PV zone a2 9,313,000 7.96 73.2 5a PV zone bi (inside) 4,622,000 7.96 36.3 5b PV zone bm (centre) 5,531,000 7.96 43.5 5c PV zone ba (outside) 9,828,000 7.96 77.3 6a PV zone cl 6,955,000 7.96 55 6b PV zone c2 6,955,000 7.96 55 6c PV zone c3 6,955,000 7.96 55 7 PV top closure 19,083,000 7.96 150 8 Top grid 2,613,000 7.96 20.8 9 Cover plate 332,000 7.96 2.65

10 Supports 75,400 7.96 0.6 11 Grid plate 352,000 7.96 2.8 12 Centring pieces 25,100 7.96 0.2 13 Upper residual mass 2,883,000 7.96 22.95 14 Core structure 8,555,000 7.96 68.1 15 Core shroud 2,261,000 7.96 18 16 Lower grid 2,198,000 7.96 17.5 17 Lower residual mass 2,688,000 7.96 21.4 18 Stool 415,000 7.96 3.3 19 Guide parts 54.32 8.4 0.456 20 Guide blocks 52.76 8.4 0.443 21 Guide pieces 128.12 8.42 1.08 22 Guide pins 64.3 8.4 0.54 23 Centring studs 602.7 8.3 5 24 "Maximum credible 420 8.4 3.53

accident' brackets

10 - 3 10 - 3 10-3 10 - 3 10-3 10-3

X10CrNiNb199 1 (+ 7) X10CrNiNb199 1 (+ 8) X10CrNiNb199 1 (+10) X10CrNiNb199 1 (+ 8) X10CrNiNb199 1 (+ 6) X10CrNiNb199 1 (+ 4) 22 NiMoCr 37 5 (+ 6) 22 NiMoCr 37 5 (+ 7) 22 NiMoCr 37 5 (+ 9) 22 NiMoCr 37 7 (+ 7) 22 NiMoCr 37 5 (+ 8 ) 22NiMoCr 37 5 (+ 7) 22 NiMoCr 37 5 (+ 5) 22 NiMoCr 37 5 (+ 3) 22 NiMoCr 37 1 (+ 2) X10CrNiNb189 4 (+ 2) X10CrNiNb189 1 ( - 1) X10CrNiNb189 1 (+ 11) X10CrNiNb189 9 (+ 11) X10CrNiNb189 9 (+ 11) X10 CrNiNb189 2 (+ 6) X10CrNiNb189 1 (+ 12) X10 CrNiNb189 1.2 (+ 13) X10CrNiNb189 9 (+ 11) X10CrNiNb189 1 (+ 6) X10 CrNiNb189 1 (+ 4) 0 .28kgCo=61% 4 (+ 2) 0 .27kgCo=61% 4 (+ 2) 0.66 kgCo = 61% 9 (+11) 0.33 kg Co = 61% 9 (+11) 2.5 kgCo--50% 9 (+II)

2.15 kg Co = 61% 9 (+11)

8 (+5) means 8 x 10 s

. ~ lO 8

o

~o ~

10 5

I0 t

BWR 40 years' operation

BWR 25 " " BWR 17 " "

PWR 40 " " PWR 25 " " PWR 17 " "

,o 2b io ,'o ~ r~ '~ ~ ~:o loo

Years after shutdown

Fig. 9. Activity inventory of 1200 MW LWR for different operating and cooling periods.


Table 12. Specific activities Ak(Ci/g) o f a 1200 MW PWR after 17 years ' operat ion and cooling t imes o f 1, 10, 20, 40 and 100 years.

No. Componen t 1 year 10 years 20 years 40 years I00 years

l a PVcladding, z o n e a l 1 . 2 - 7 2 . 4 - 8 8 . 6 - 9 4 . 9 - 9 3 . 1 - 9 l b PVcladding, z o n e a 2 1 . 2 - 6 2 . 4 - 7 8 . 6 - 8 4 . 9 - 8 3 . 1 - 8 2 PVeladding, z o n e b 1 . 2 - 4 2 . 4 - 5 8 . 6 - 6 4 . 9 - 6 3 . 1 - 6

3a PVcladding, z o n e c l 1 . 2 - 6 2 . 4 - 7 8 . 6 - 8 4 . 9 - 8 3 . 1 - 8 3b PV cladding, zone c2 1.2 - 8 2.4 - 9 8 . 6 - 1 0 4 . 9 - 1 0 3 . 1 - 1 0

3c PV cladding, zone c3 1.2 - 10 2.4 - 11 8.6 - 12 4.9 - 12 3.1 - 12

4a P V z o n e a l 9 . 7 - 8 1 . 8 - 8 3 . 6 - 9 4 . 0 - 1 0 1 . 3 - 10 4b P V z o n e a 2 9 . 7 - 7 1 . 8 - 7 3 . 6 - 8 4 . 0 - 9 1 . 3 - 9 5a P V z o n e b i ( i n s i d e ) 9 . 7 - 5 1 . 8 - 5 3 . 6 - 6 4 . 0 - 7 1 . 3 - 7

5b P V z o n e b m ( c e n t r e ) 1 . 4 - 6 2 . 5 - 7 5 . 1 - 8 5 . 6 - 9 1 . 8 - 9 5c P V z o n e b a ( o u t s i d e ) 9 . 7 - 6 1 . 8 - 6 3 . 6 - 7 4 . 0 - 8 1 . 3 - 8 6a P V z o n e e l 9 . 7 - 7 1 . 8 - 7 3 . 6 - 8 4 . 0 - 9 1 . 3 - 9

6b P V z o n e c2 9.7 - 9 1.8 - 9 3.6 - 10 4.0 - 11 1.3 - 11 6c PV zone c3 9.7 - 11 1.8 - 11 3.6 - 12 4 . 0 - 13 1.3 - 13 7 PV top closure 1.9 - 12 3.6 - 13 7.3 - 14 8.1 - 15 2".6 - 15

8 Top grid 4.2 - 12 7.7 - 13 3.0 - 13 2.0 - 13 1.3 - 13 9 Cover plate 1 . 0 - 15 1.9 - 16 7.5 - 17 5.1 - 17 3.2 - i7

10 Supports 1 . 0 - 3 1 . 9 - 4 7 . 5 - 5 5 . 1 - 5 3 . 2 - 5

11 Grid plate 9 . 4 - 3 1 . 7 - 3 6 . 7 - 4 4 . 6 - 4 2 . 9 - 4 12 Centring pieces 9 . 4 - 3 1 . 7 - 3 6 . 7 - 4 4 . 6 - 4 2 . 9 - 4 13 Upper res idua lmass 2 . 1 - 8 3 . 8 - 9 1 . 5 - 9 1 . 0 - 9 6 . 4 - 1 0

14 Core structure 1 . 2 - 2 2 . 4 - 3 8 . 9 - 4 5 . 2 - 4 3 . 2 - 4 15 Core sh toud 1 . 4 - 1 2 . 9 - 2 1 . 1 - 2 6 . 3 - 3 3 . 9 - 3 16 Lower grid 1 . 1 - 2 2 . 2 - 3 8 . 0 - 4 4 . 7 - 4 2 . 9 - 4

17 Lower residual mass 1.2 - 8 2.4 - 9 8.9 - 10 5.2 - 10 3.2 - 10 18 Stool 1.2 - 10 2.4 - 11 8.9 - 12 5.2 - 12 3.2 - 12 19 Guide parts 7.2 - 10 2.2 - 10 5.9 - 11 4.4 - 12 6.6 - 14

20 Guide blocks 7.2 - 10 2.2 - 10 5.9 - 11 4.4 - 12 6.6 - 14 21 Guide pieces 1 . 6 + 0 4 . 9 - 1 1 . 3 - 1 9 . 8 - 3 1 . 5 - 4 22 Guide pins 1 .6+ 0 4 . 9 - 1 1 . 3 - 1 9 . 8 - 3 1 . 5 - 4

23 Centring s tuds 1 .3+ 0 4 . 1 - 1 1 . 1 - 1 8 . 4 - 3 2 . 9 - 4 24 MCAbracke t s 1 .6+ 0 4 . 9 - 1 1 . 3 - 1 9 . 8 - 3 1 . 5 - 4

2.1 - 2 means 2.1 x 10 - 2

m e n t s t o t h e a r e a o f a 1 2 0 0 M W P W R w i t h w h i c h t h e

p r i m a r y w a t e r c o m e s i n t o c o n t a c t g ives a 6 0 C o act iv i -

t y o f 6 X 1 0 4 0 .

H o w e v e r , t h e f o l l o w i n g c o n s i d e r a t i o n s h o w s t h a t

t h i s v a l u e is u n r e a l i s t i c a l l y h i g h . I f t h e m o n t h l y r a t e

o f r e l ease o f Co i n t o t h e p r i m a r y w a t e r s y s t e m o f a

P W R o f t h e 1 2 0 0 M W class ( a b o u t 2 0 g ) is m u l t i p l i e d

b y t h e spec i f i c Co a c t i v i t y o f a c o m p o n e n t c l o se t o t h e


Table 13. Activities (Ci) of a 1200 MW PWR after 17 years' operation and cooling times of 1, 10, ~ 20, 40 and 100 years

27


la PVcladding, z o n e a l 1 . 2 - 1 2 . 4 - 2 8 . 6 - 3 4 . 9 - 3 3 . 1 - 3 lb PVcladding, zonea2 1.2+ 0 2 . 4 - 1 8 . 6 - 2 4 . 9 - 2 3 . 1 - 2 2 PVcladding, zoneb 3,5+ 2 7.1+ 1 2.5+ 1 1.4+ 1 8.9+ 0

3a PVcladding, z o n e c l 9 . 0 - 1 1 . 8 - 1 6 . 4 - 2 3 . 7 - 2 2 . 3 + 2 3b PVcladding, zonec2 9 . 0 - 3 1 . 8 - 3 6 . 4 - 4 3 . 7 - 4 2 . 3 - 4 3c PVcladding, zonec3 9 . 0 - 5 1 . 8 - 5 6 . 4 - 6 3 . 7 - 6 2 . 3 - 6

4a P V z o n e a l 2 .5+ 0 4 . 6 - 1 9 . 3 - 2 1 . 0 - 2 3 . 3 - 3 4b PVzonea2 7.1+ 1 1.3+ 1 2.7+ 0 2 . 9 - 1 9 . 4 - 2 5a PVzonebi ( ins ide) 3.5+ 3 6 .5+ 2 1.3+ 2 1.4+ 1 4 .7+ 0

5b PVzonebm(cen t r e ) 5 .9+ 1 1.1+ 1 2.2+ 0 2 . 4 - 1 7 . 9 - 2 5c PVzoneba(outs ide) 7 .5+ 2 1.4+ 2 2.8+ 1 3.1+ 0 9 . 9 - 1 6a P V z o n e c l 5 .3+ 1 9 .9+ 0 2.0+ 0 2 . 2 - I 7 . 1 - 2

6b PVzonec2 5 . 3 - 1 9 . 9 - 2 2 . 0 - 2 2 . 2 - 3 7 . 1 - 4 6c PVzonec3 5 . 3 - 3 9 . 9 - 4 2 . 0 - 4 2 . 2 - 5 7 . 1 - 6 7 PVto-pclosure 2 . 9 - 4 5 . 4 - 5 1 . 1 - 5 1 . 2 - 6 3 . 8 - 7

8 Top grid 8 . 6 - 5 1 . 6 - 5 6 . 2 - 6 4 . 2 - 6 2 . 7 - 6 9 Cover plate 2.7 - 9 5.1 - 10 1.9 - I0 1.3 - 10 8.5 - 11

10 Supports 6 .2+ 2 1.1+ 2 4 .5+ 1 3.1+ 1 1.9+ 1

11 Gridplate 2 .6+ 4 4 .9+ 3 1.9+ 3 1.3+ 3 8.1+ 2 12 Centring pieces 1.8+ 3 3.5+ 2 1.3+ 2 9.2+ 1 5.8+ 1 13 Upper residualmass 4 . 8 - 1 8 . 9 - 2 3 . 4 - 2 2 . 3 - 2 1 . 5 - 2

14 Core structure 8 ,2+ 5 1.7+ 5 6 .1+ 4 3.5 + 4 2.2 + 4 15 Core shroud 2,6+ 6 5.3+ 5 1.9+ 5 1.1+ 5 6 .9+ 4 16 Lower grid 1.9+ 5 3.9+ 4 1.4+ 4 8.2+ 3 5 .1+ 3

17 Lower residualmass 2 . 6 - 1 5 . 3 - 2 1 . 9 - 2 1 . 1 - 2 6 . 9 - 3 18 Stool 4 , 0 - 4 8 . 1 - 5 2 . 9 - 5 1 . 7 - 5 1 . 1 - 5 19 Guideparts 3 . 2 - 7 1 . 0 - 7 2 . 7 - 8 2 . 0 - 9 3 . 0 - 1 1

20 Guide blocks 3 . 1 - 7 9 . 7 - 8 2 . 6 - 8 1 . 9 - 9 2 . 9 - 1 1 21 Guide pieces 1.7+ 3 5.3+ 2 1.4+ 2 1.0+ 1 1 . 6 - 1 22 Guide pins 8 .7+ 2 2 . 7 + 2 7.2+ 1 5.3+ 0 8 . 0 - 2

23 Centring studs 6.6 + 3 2 .0+ 3 5.5 + 2 4 .2+ 1 1,5 + 0 24 MCAbrackets 5 .7+ 3 1.7+ 3 4 .7+ 2 3.5+ 1 5 . 2 - 1

Total 3.6+ 6 7.5+ 5 2.7+ 5 1.5+ 5 9 .7+ 4

2.1 - 2 means 2.1 X 10 - 2

core (A = 8 . 4 Ci/g), t he resul t is a 6°Co ac t iva t ion rate

o f 1.7 X 10 2 O / m o n t h . The ca lcu la ted ac t iv i ty over 5

years is thus a b o u t 1 X 10 4 Ci.

Clearly, t h e n , even o n t he a s s u m p t i o n o f e x t r e m e l y

conserva t ive values, a t o t a l ac t iv i ty o f the o rde r o f

t h a t o b t a i n e d b y e x t r a p o l a t i o n o f the KWO measure-

m e n t s is n o t a t t a ined .

I t fo l lows t h a t th is value m u s t be regarded as a

28 NIS / Decommissioning o f L WR nuclear power plants

Table 14. Specific activities Ak(Ci/g ) of a 1200 MW PWR after 25 years' operat ion and cooling t imes of 1, 10, 20, 40 and 100 years

No. Componen t 1 year 10 years 20 years 40 years 100 years

l a PVcladding, z o n e a l 1 . 3 - 7 2 . 8 - 8 1 . 1 - 8 7 . 0 - 9 4 . 4 - 9 lb PVcladding, z o n e a 2 1 . 3 - 6 2 . 8 - 7 1 . 1 - 7 7 . 0 - 8 ~ 4 . 4 - 8 2 PVcladding, z o n e b 1 . 3 - 4 2 . 8 - 5 1 . 1 - 5 7 . 0 - 6 4 . 4 - 6

3a PVcladding, z o n e c l 1 . 3 - 6 2 . 8 - 7 1 . 1 - 7 7 . 0 - 8 4 . 4 - 8 3b PV cladding, zone c2 1.3 - 8 2.8 - 9 1.1 - 9 7.0 - 10 4.4 - 10 3c PV cladding, zone c3 1.3 - 10 2.8 - 11 1.1 - 11 7.0 - 12 4.4 - 12

4a P V z o n e a l 1 . 0 - 7 1 . 9 - 8 3 . 9 - 9 5 . 0 - 1 0 1 . 8 - 1 0 4b P V z o n e a 2 1 . 0 - 6 1 . 9 - 7 3 . 9 - 8 5 . 0 - 9 1 . 8 - 9 5a P V z o n e b i ( i n s i d e ) 1 . 0 - 4 1 . 9 - 5 3 . 9 - 6 5 . 0 - 7 1 . 8 - 7

5b P V z o n e b m ( c e n t r e ) 1 . 4 - 6 2 . 7 - 7 5 . 6 - 8 7 . 1 - 9 2 . 6 - 9 5c P V z o n e b a ( o u t s i d e ) 1 . 0 - 5 1 . 9 - 6 3 . 9 - 7 5 . 0 - 8 1 . 8 - 8 6a P V z o n e c l 1 . 0 - 6 1 . 9 - 7 3 . 9 - 8 5 . 0 - 9 1 . 8 - 9

6b P V z o n e c2 1 . 0 - 8 1.9 - 9 3.9 - 10 5.0 - 13 1.8 - 11 6c PV zone c3 1.0 - 10 1.9 - 11 3.9 - 12 5.0 - 13 1.8 - 13 7 PV top closure 2.0 - 12 3.8 - 13 7.9 - 14 1 . 0 - 14 3.7 - 15

8 Top grid 4.3 - 12 8.9 - 13 4 . 0 - 13 2.9 - 13, 1.8 - 13 9 Cover plate 1.1 - 15 2.2 - 16 1.0 - 16 7.3 - 17 4.6 - 17

10 Supports 1 . 1 - 3 2 . 2 - 4 1 . 0 - 4 7 . 3 - 5 4 . 6 - 5

11 Grid plate 9 . 8 - 3 2 . 0 - 3 9 . 1 - 4 6 . 6 - 4 4 . 1 - 4 12 Centring pieces 9 . 8 - 3 2 . 0 - 3 9 . 1 - 4 6 . 6 - 4 4 . 1 - 4 13 Upper res idua lmass 2 . 2 - 8 4 . 5 - 9 2 . 0 - 9 1 . 5 - 9 9 . 2 - 10

14 Core structure 1 . 3 - 2 2 . 8 - 3 1 . 1 - 3 7 . 4 - 4 4 . 6 - 4 15 Core shroud 1 . 5 - 1 3 . 4 - 2 1 . 4 - 2 8 . 9 - 3 5 . 5 - 3 16 Lower grid 1 . 1 - 2 2 . 5 - 3 1 . 0 - 3 6 . 7 - 4 4 . 1 - 4

17 Lower residual mass 1 . 2 - 8 2 . 8 - 9 1 . 1 - 9 7 . 4 - 1 0 4 . 6 - 10 18 Stool 1.3 - 10 2.8 - 11 1.1 - 11 7.4 - 12 4.6 - 12 19 Guide parts 7.7 - 10 2.4 - 10 6.4 - 11 4.8 - 12 9.4 - 14

20 Guide blocks 7.7 - 10 2.4 - 10 6 . 4 - 11 4.8 - 12 9.4 - 14 21 Guide pieces 1 .7+ 0 5 . 3 - 1 1 . 4 - 1 1 . 1 - 2 2 . 1 - 4 22 Guide pins 1 .7+ 0 5 . 3 - 1 1 . 4 - 1 1 . 1 - 2 2 . 1 - 4

23 Centring s tuds 1 .4+ 0 4 . 4 - 1 1 . 2 - 1 9 . 2 - 3 4 . 2 - 4 24 MCAbracke t s 1 .7+ 0 5 . 3 - 1 1 . 4 - 1 1 . 1 - 2 2 . 1 - 4

2 . 1 - 2 m e a n s 2 .1 X 1 0 - 2

m a x i m u m , w h i c h c a n n o t b e u s e d as t h e ave r age fo r

t h e s y s t e m as ~a w h o l e . A n e s t i m a t e o f t h e spec i f i c ac-

t i v i t y o f t h e c o n t a m i n a t i o n s o f 1 - 1 0 # C i / c m z a p p e a r s

rea l i s t ic . T h e h i g h e r v a l u e l e a d s t o a n a c t i v i t y d u e t o

c o n t a m i n a t i o n o f 2 . 9 X 103 Ci f o r a P W R o f t h e 1 2 0 0

M W class . F o r l a 4 C s , l a T C s a n d 60Co, measurements o n t h e s t e a m s ide o f K R B - I s h o w e d a n ave rage spec i f i c

a c t i v i t y o f 1 .4 X 10 - 2 / ~ C i / c m 2 ( a f t e r 5 y e a r s ' o p e r a -

t i o n ) . T h i s i n d i c a t e s a n a c t i v i t y o f 3 Ci f o r a B W R o f

t h e 1 2 0 0 M W class . I f a s t r i p p i n g f a c t o r o f 103 is al-


Table 15. Activities (Ci) of a 1200 MW PWR after 25 years' operation and cooling times of 1, 10, 20, 40 and 100 years.

29


la PVcladding, z o n e a l 1 . 3 - 1 2 . 8 - 2 1 . 1 - 2 ~ 7 . 0 - 3 4 . 4 - 3 lb PVcladding, zonea2 1.3+ 0 2 . 8 - 1 1 . 1 - 1 7 . 0 - 2 4 . 4 - 2 2 PVcladding, zoneb 4 .6+ 2 1.0+ 2 4.0+ 1 2.5+ 1 1.6+ 1

3a PVcladding, z o n e c l 9 . 5 - 1 2 . 0 - 1 8 . 3 - 2 5 . 2 - 2 3 . 2 - 2 3b PVcladding, zone c2 9 . 5 - 3 2 . 0 - 3 8 . 3 - 4 5 . 2 - 4 3 . 2 - 4 3c PVcladding, zonec3 9 . 5 - 5 2 . 0 - 5 8 . 3 - 6 5 . 2 - 6 3 . 2 - 6

4a P V z o n e a l 2 .8+ 0 4 . 8 - 1 1 . 0 - 1 1 . 3 - 2 4 . 7 - 3 4b PVzonea2 7.4+ 1 1.4+ 1 2.9+ 0 3 . 7 - 1 1 . 3 - 1 5a PVzonebi ( ins ide) 3.6+ 3 6.9+ 2 1.4+ 2 1.8+ 1 6 .7+ 0

5b PVzonebm(cen t r e ) 6 .2+ 1 1.1+ 1 2.4+ 0 3 . 1 - 1 1 . 1 - 1 5c PVzoneba(outs ide) 7 .8+ 2 1.5+ 2 3.1+ 1 3.9+ 0 1.4+ 0 6a P V z o n e c l 5 .5+ 1 1.0+ 1 2.2+ 0 2 . 8 - 1 1 . 0 - 1

6b PVzonec2 5 . 5 - 1 1 . 0 - 1 2 . 2 - 2 2 . 8 - 3 1 . 0 - 3 6c PVzonec3 5 . 5 - 3 1 . 0 - 3 2 . 2 - 4 2 . 8 - 5 1 . 0 - 5 7 PV top closure 3 . 0 - 4 5 . 7 - 5 1 . 2 - 5 1 . 5 - 6 5 . 5 - 7

8 Top grid 9 . 1 - 5 1 . 8 - 5 8 . 4 - 6 6 . 1 - 6 3 . 8 - 6 9 Cover plate 2.9 - 9 5.9 - 10 2.6 - 10 1.9 - 10 1.2 - 10

10 Supports 6 .5+ 2 1.3+ 2 6.0+ 1 4 .2+ 1 2.8+ 1

11 Grid plate 2 .7+ 4 5.6+ 3 2.5+ 3 1.8+ 3 1.1+ 3 12 Centring pieces 1.9 + 3 4 .0+ 2 1.8+ 2 1.3+ 2 8.3+ 1 13 Upper residualmass 5 . 0 - 1 1 . 0 - 1 4 . 6 - 2 3 . 3 - 2 2 . 1 - 2

14 Core structure 8.7 + 5 1.9 + 5 7 .9+ 4 5.0+ 4 3.1+ 4 15 Core shroud 2.7 + 6 6 .1+ 5 2.5 + 5 1.6 + 5 9.9 + 4 16 Lower grid 2 .0+ 5 4 .4+ 4 1.8+ 4 1.2+ 4 7.3+ 3

17 Lower residualmass 2 . 7 - 1 6 . 0 - 2 2 . 5 - 2 1 . 6 - 2 9 . 9 - 3 18 Stool 4 . 2 - 4 9 . 3 - 5 3 . 8 - 5 2 . 4 - 5 1 . 5 - 5 19 Guideparts 3 . 5 - 7 1 . 1 - 7 2 . 9 - 8 2 . 2 - 8 4 . 3 - 1 1

20 Guide blocks 3 . 4 - 7 1 . 1 - 7 2 . 8 - 8 2 . 1 - 9 4 . 1 - 1 1 21 Guide pieces 1.8+ 3 5.7+ 2 1.5+ 2 1.2+ 1 2 . 3 - 1 22 Guide pins 9 .4+ 2 2 .9+ 2 7 .8+ 1 5 .8+ 0 1 . 1 - 1

23 Centring studs 7 .1+ 3 2.2+ 3 5.9+ 2 4 .6+ 1 2 .1+ 0 24 MCAbrackets 6 .1+ 3 1.9+ 3 5.1+ 2 3.8+ 1 7 . 4 - 1

Total 3 .8+ 6 8.5+ 5 3.5+ 5 2.1+ 5 1.4+ 5

2.1 - 2 means 2.1 X 10 - 2

lowed as the d i f fe rence b e t w e e n the BWR and t he

PWR, these values c o n f i r m the a s s um pt i ons for the PWR.

The da ta avai lable as t o the evo lu t i on o f t he con-

t a m i n a t i o n s in t ime are pa r t l y c o n t r a d i c t o r y . Some

au tho r s cons ider t h a t t he act ivi t ies due to c o n t a m i n a -

t i on show a l inear increase w i t h t ime. One reason for

this wou ld be t h a t the en t i re surface o f t he p r ima ry


Table 16 Specific activities Ak(Ci]g) of a 1200 MW PWR after 40 years' operation and cooling times of 1, 10, 20, 40 and 100 years


la PVcladding, zoneal 1 . 3 - 7 3 . 2 - 8 1 . 5 - 8 1 . 1 - 8 6 . 6 - 9 Ib PVcladding, zonea2 1 . 3 - 6 3 . 2 - 7 1 . 5 - 7 1 . 1 - 7 6 . 6 - 8 2 PVcladding, zoneb 1 . 3 - 4 3 . 2 - 5 1 . 5 - 5 1 . 1 - 5 6 . 6 - 6

3a PVcladding, zonec l 1 . 3 - 6 3 . 2 - 7 1 . 5 - 7 1 . 1 - 7 6.6-~ 8 3b PVcladding, zonec2 1 . 3 - 8 3 . 2 - 9 1 . 5 - 9 1 . 1 - 9 6 . 6 - 10 3c PV cladding, zone c3 1.3 - 10 3 . 2 - 11 1.5 - 11 1.1 - 11 6.6 - 12

4a P Vzonea l 1 . 0 - 7 1 . 9 - 8 4 . 2 - 9 6 . 6 - 1 0 2 . 8 - 1 0 4b PVzonea2 1 . 0 - 6 1 . 9 - 7 4 . 2 - 8 6 . 6 - 9 2 . 8 - 9 5a PVzonebi(inside) 1 . 0 - 4 1 . 9 - 5 4 . 2 - 6 6 . 6 - 7 2 . 8 - 7

5b PVzonebm(centre) 1 . 4 - 6 2 . 7 - 7 5 . 9 - 8 9 . 3 - 9 3 . 9 - 9 5e PVzoneba(outside) 1 . 0 - 5 1 . 9 - 6 4 . 2 - 7 6 . 6 - 8 2 . 8 - 8 6a PVzonee l 1 . 0 - 6 1 . 9 - 7 4 . 2 - 8 6 . 6 - 9 2 . 8 - 9

6b PVzone c2 1.0 - 8 1.9 - 9 4.2 - 10 6.6 - 11 2.8 - 11 6c PV zone c3 1 . 0 - 10 1.9 - 11 4.2 - 12 6.6 - 13 2.8 - 13 7 PV top closure 2.0 - 12 3.9 - 13 8.5 - 14 1.3 - 14 5.6 - 15

8 Top grid 4.6 - 12 1.1 - 12 5.8 - 13 4.4 - 13 2.8 - 13 9 Cover plate 1.1 - 15 2.7 - 16 1.4 - 16 1.1 - 16 6.9 17

10 Supports 1 . 1 - 3 2 . 7 - 4 1 . 4 - 4 1 . 1 - 4 6 . 9 - 5

11 Grid plate 1 . 0 - 2 2 . 4 - 3 1 . 3 - 3 9 . 9 - 4 6 . 3 - 4 12 Centring pieces 1 . 0 - 2 2 . 4 - 3 1 . 3 - 3 9 . 9 - 4 6 . 3 - 4 13 Upper residualmass 2 . 3 - 8 5 . 4 - 9 2 . 9 - 9 2 . 2 - 9 1 . 4 - 9

14 Core strucmre 1 . 3 - 2 3 . 3 - 3 1 . 6 - 3 1 . 1 - 3 6 . 9 - 4 15 Co~eshroud 1 . 6 - 1 3 . 9 - 2 1 . 9 - 2 1 . 3 - 2 8 . 4 - 3 16 Lower grid 1 . 2 - 2 2 . 9 - 3 1 . 4 - 3 1 . 0 - 3 6 . 3 - 4

17 Lower residualmass 1 . 3 - 8 3 . 3 - 9 1 . 6 - 9 1 . 1 - 9 6 . 9 - 10 18 Stool 1.3 - 10 3.3 - 11 1.6 - 11 1.1 - 11 6.9 - 12 19 Guide parts 8.0 - 10 2.4 - 10 6.6 - 11 4.9 - 12 1.4 - 13

20 Guide blocks 8.0 - 10 2.4 - 10 6.6 - 11 4.9 - 12 1.4 - 13 21 Guide pieces 1.8 + 0 5 . 5 - 1 1 . 5 - 1 1 . 1 - 2 3 . 2 - 4 22 Guide pins 1.8+ 0 5 . 5 - 1 1 . 5 - 1 1 . 1 - 2 3 . 2 - 4

23 Centring studs 1.5+ 0 4 . 5 - 1 1 . 2 - 1 9 . 8 - 3 6 . 3 - 4 24 MCAbrackets 1.8+ 0 5 . 5 - 1 1 . 5 - 1 1 . 1 - 2 3 . 2 - 4

2.1 - 2 means 2.1 X 1.0 - 2

loop is con t inuous ly changing owing to corros ion,

the two mos t i m p o r t a n t processes being the release

o f material to the coo lan t circuit and the accumula-

t ion o f act ivated material .

The o the r school o f t h o u g h t considers tha t the ac-

t ivated material is depos i ted only at cer tain po in t s

in the pr imary sys tem which are part icularly suscept.

ible for geometr ical reasons (bl ind holes, angles,

backed-o f f points , etc.) . In this case the act ivi ty due

to co n t ami n a t i o n would t end towards a sa tura t ion


Table 17 Activities (Ci) of a 1200 MW PWR after 40 years' operation and cooling times of 1, 10, 20, 40 and 100 years

31


la PVcladding, z o n e a l 1 . 3 - 1 3 . 2 - 2 1 . 5 - 2 1 . 1 - 2 6 . 6 - 3 lb PVcladding, zonea2 3.9+ 0 9 . 4 - 1 4 . 4 - 1 3 . 1 - 1 1 . 9 - 1 2 PVcladding, zoneb 4 .8+ 2 1.2+ 2 5 .5+ 1 3.8+ 1 2.4+ 1

3a ] cladding, zone cl 9 . 9 - 1 2 . 4 - 1 1 . 1 - 1 7 . 8 - 2 4 . 9 - 2 3b Pv cladding, zonec2 9 . 9 - 3 2 . 4 - 3 1 . 1 - 3 7 . 8 - 4 4 . 9 - 4 3c PVcladding, zonec3 9 . 9 - 5 2 . 4 - 5 1 . 1 - 5 7 . 8 - 6 4 . 9 - 6

4a P V z o n e a l 2 .6+ 0 5 . 0 - 1 1 . 1 - 1 1 . 7 - 2 7 . 1 - 3 4b PV zone a2 7.5 + i 1.4 + 1 3.1 + 0 4.8 - 1 2.0 - 1 5a PVzonebi ( ins ide) 3.7+ 3 7.1+ 2 1.5+ 2 2.4+ 1 1.0+ 1

5b PVzonebm(cen t r e ) 6 .2+ 1 1.2+ 1 2.6+ 0 4 . 0 - 1 1 . 7 - 1 5c PVzoneba(outs ide) 7 .9+ 2 1.5+ 2 3.3+ 1 5.1+ 0 2 .1+ 0 6a P V z o n e c l 5 .6+ 1 1.1+ 1 2.3+ 0 3 . 6 - 1 1 . 5 - 1

6b PVzonec2 5 . 6 - 1 1 . 1 - 1 2 . 3 - 2 3 . 6 - 3 1 . 5 - 3 6c PVzonec3 5 . 6 - 3 1 . 1 - 3 2 . 3 . 4 3 . 6 - 5 1 . 5 - 5 7 PV top closure 3 . 1 - 4 5 . 9 - 5 1 . 3 - 5 1 . 9 - 6 8 . 4 - 7

8 Top grid 4.6 - 12 1.t - 12 5.8 - 13 4.4 - 13 2.8 - 13 9 Cover plate 3.0 - 9 7.2 - 10 3.8 - 10 2.9 - 10 1.8 - 10

10 Supports 6 .9+ 2 1.6+ 2 8.7+ 1 6.6+ 1 4 .2+ 1

11 Grid plate 2 .9+ 4 6 .9+ 3 3.6+ 3 2.8+ 3 1.7+ 3 12 Centring pieces 2 .1+ 3 4 .9+ 2 2 .6+ 2 1.9+ 2 1.2+ 2 13 Upper residualmass 5 . 2 - 1 1 . 2 - 1 6 . 6 - 2 5 . 1 - 2 3 . 2 - 2

14 Core structure 9 .1+ 5 2 .2+ 5 1.1 + 5 7 .6+ 4 4.7 + 4 15 Core shroud 2.9+ 6 7.1+ 5 3.4+ 5 2.4+ 5 1.5+ 5 16 Lower grid 2 .1+ 5 5 .2+ 4 2.5+ 4 1.7+ 4 1.1+ 4

17 Lower residualmass 2 . 8 - 1 7 . 1 - 2 3 . 4 - 2 2 . 4 - 2 1 . 5 - 2 18 Stool 4 . 4 - 4 1 . 1 - 4 5 . 3 - 5 3 . 7 - 5 2 . 3 - 5 19 Guide parts 3 . 6 - 7 1 . 1 - 7 3 . 0 - 8 2 . 3 - 9 6 . 4 - 1 1

20 Guide blocks 3 . 5 - 7 1 . 1 - 7 2 . 9 - 8 2 . 2 - 9 6 . 3 - 1 1 21 Guide pieces 1.9+ 3 5.9+ 2 1 . 6 - 2 1 . 2 - 2 3 . 4 - 1 22 Guide pins 9 .7+ 2 2 .9+ 2 8 .5+ 1 6 .1+ 0 1 . 7 - 1

23 Centring studs 7 .4+ 3 2.3 + 3 6 .1+ 2 4 .9+ 1 3.1+ 0 24 MCAbrackets 6 .3+ 3 1.9+ 3 5 .2+ 2 3.9+ 1 1.1+ 0

Total 4 .1+ 6 9 .9+ 5 4 .8+ 5 3.4+ 5 2.1+ 5

2.1 - 2 means 2.1 X 10 - 2

value, w h i c h is r eached w h e n all possible pos i t ions

have fil led c o m p l e t e l y w i t h mater ia l . Dose ra te mea-

s u r e m e n t s o n c o m p o n e n t s o f KRB-1 for whose acti-

v i ty on ly c o n t a m i n a t i o n can be respons ib le ind ica te

s a tu ra t i on a f te r a b o u t 5 years . However , th is does

n o t necessar i ly m e a n t h a t t he same applies to all

sizes a n d types o f reactors .

Rad ioac t ive masses o f a 1200 MW nuc lear power

3 2 NIS / Decommissioning of L WR nuclear power plan ts

i ~rear's cooling time

low-level

medium-level

high-level

40 29ars' cooling time

low-level

me dium-leve 1

high-leve i

BWR

564 54.9

371 36 • 1

92 9.0

724.5

239.8

62.6

70.5

23.3

6.2

287.5

360.5

! Ii0.0

530.8

121.4

105.8

PWR

1%

37.9

47.6

14.5

7O

16

14

724.5

I

i

i

I 564 Ng I

i year

"" - 40 years

371 M

-2-• 62.6

Low-am diwm-high-leve 1

B~R

530.8 ~4g r ----I I

I

I

I

I I t~87.5~4~

360.5 );~

L21.4 ~,I~ no ~!~ 1o5.8

Low-me dium-hi~h-leve i

PWR

r~

Fig. 10. 'Graphical presentation of pressure vessel component masses divided into low-level, medium-level and high-level waste,

plant are set out in table 18. In addition to the systems already mentioned, the table includes particu- lars of the contaminated masses of the auxiliaries. The contaminations here are contained primarily in the reactor water purification system (or volume compensation system and coolant treatment system), the emergency cooling and aftercooling system, and

the effluent treatment plant (collection system and concentrate treatment system).

In addition, buildings (reactor and auxiliaries buildings) may be contaminated, though such contamination is slight compared with the 'secondary' contaminations which can occur as a result of the decommissioning works. The extent of the contamination

NIS / Decommissioning of L WR nuclear power plants

Table 18. Radioactive masses of a 1200 MW nuclear power plant

33

System BWR [Mg] PWR [Mg] activated contaminated activated contaminated

1. Reactor pressure vessel a) vessel 777 580 b) internals, excluding 238 178

fuel elements

2. Primary loop a) main coolant pumps 12 b) pipework 1000 c) fittings 100 d) other components and systems 400

3. Secondary loop ./.

4. Turbine

5. Auxiliaries

Total scrap 1027

6. Concrete - biological shield 430

1100

1850

4450 (excluding items 1 and 2a)

758

200 450

50 1800

430

free

free

850

3350 (excluding item 1)

is ascertained from checks, and decontamination or fixing is effected (see also sections 3 and 5).

4.3. Estimation o f dose rates

Estimation o f the dose rates in the intensely irradiated areas of a nuclear power plant with an electrical rating of 1200 MW at the time of decommissioning is necessary in order to evaluate the extent to which it is possible to work in specific parts o f the buildings and what protective facilities will be necessary.

Since no more activities are formed after final shutdown of the reactor and unloading of the reactor core, it can be stated as a general rule that, at the time when the decommissioning works are commenced, a considerably lower radiation level is to be expected in the controlled area than during power generation.

For the purpose of estimating dose rates, the origin of the activities is irrelevant. The dose rate at a field

point depends on: (a) nature of radiation (a, #, -/), (b) energy of radiation, (c) source strength (see- l ) , (d) emission probability, (e) source geometry, (f) distance o f field point from source, and (g) layer thickness and material of shielding. Since determination o f the above parameters is often a matter of considerable difficulty, and since errors already attach to the input values, even an exact com- putation can only indicate the order of magnitude of the dose rate at a field point. For determination o f dose rates in closed systems (where the activities occur in a tank, pipeline, etc.), only the "/-radiation emitted is important. The c~ and/~ radiation is negligible because o f its short range where the wall thicknesses of the tanks or pipelines act as shields. However, if the

and/3 radiation can act direct (i.e. without any


shielding) on a person, this radiation must also be taken into account. This applies in the case of work in or on an 'open' source.

In what follows, details are given of the dose rates on nuclear components to be expected at the time of decommissioning. The component which is responsible for the highest dose rate at the time of decommissioning is the reactor pressure vessel with inter- hal core structures. The dose rates are estimated by the standard methods described in the literature [2] and in accordance with the activities determined in section 4.1, contamination being disregarded. Dose rates of the order of 104-105 R/hr are to be expected in the centre of the pressure vessel at the level of the cover flange at the time of decommissioning, if the vessel is not filled with water. If it is filled with water up to the cover flange, the dose rate due to direct radiation from the internal structures is negligibly small at this field point. The dose rate due to the contaminated water is predominant here. As an example we may cite the replacement of the core stool in the KKS, where a reactor water activity of 10-3Ci/m a caused a dose rate of 10 mrem/hr. Dose rates of the order of 5 R/hr are expected on the outside of the pressure vessel radially half-way up the core if the internal structures of the vessel have been removed. In this connection it is unimportant whether the vessel is empty or filled with water, because the dose rate is substantially deter-

mined by the radiation of the vessel cladding and the vessel itself.

Table 19 sets out the dose rate at individual field points on the reactor pressure vessel of a BWR and a PWR plant with a gross electrical rating of 1200 MW.

The dose rates on pipelines and components of the primary loop are due mainly to the isotopes 6°Co and S4Mn. Since the nonvolatile active fission and corrosion products are present in the primary water, pipelines and components with which it comes into contact are responsible for high dose rates.

In a PWR plant the main large components and pipelines which come into contact with the primary water are as follows: steam generators, pressurizer, main coolant pumps, and primary water lines. In a PWR plant currently in service, dose rates of 30 R/hr have already been measffred on steam generators. Dose rates of up to 40 R/hr were measured on parts of the forced circulation pumps at the Stade nuclear power plant after only 2 years' operation [3]. These dose rates appear somewhat too high for the time of decommissioning, since the dose rate contributions of some isotopes (sl Cr, Saco) decline in the standstill period between final shutdown and the decommissioning works. However, dose rates in the region of a few R/hr are likely to be encountered.

In a BWR plant, by far the largest part of the solid active fission and corrosion products is also to be

Table 19. Estimated dose rate on commencement of decommissioning of a 1200 MW nuclear power plant after 40 years' operation (R/hr)

A With water Without water

with without with without internal internal internal internal structures structures structures structures

BWR

A 10 -20 10 -20 6 • 104 5 • 10 -2

B - 10 -6 - 11

C - 7 - >7

D - 5 - 5

PWR

A I0 -15 10 -9 2 • 104 4

B - 6 • 10 -4 - 70

C - 50 - >50

D - 5 - 5

NIS / Decommissioning o fLieR nuclear power plants 35

found in the primary water. However, owing to the low humidity of the live steam and the corresponding- ly high stripping factor, only very small quantities of solid active fission and activation products are transported into the steam power plant, so that the dose rates on these components and pipelines at the time of decommissioning will be significantly lower than in the case of components and pipelines with which the primary water comes into contact. On the basis of published analytical values for KRB-I [4], dose rates of less than 100 mR/hr are estimated on the turbine. The values apply to a time 5 days after reactor shutdown, so that given a longer cooling period the level of the dose rate falls further.

In practice, since these estimates can give no more than a general impression, only a programme of measurement and extensive calculations can afford exact values for the dose rates.

5. Decommissioning costs

Compared with the decommissioning of conventional power plants, that of a nuclear power plant poses a number of specific questions due to activation of the structural materials and contamination of the systems and structures. On account of these specific problems, the decommissioning of nuclear power plants involves high costs.

One difficulty in the way of estimating decommissioning costs is that hitherto no plant on the scale in question has been decommissioned. Hence, cost information cannot be derived direct from experience. An exact cost calculation based on detailed work schedules, requests for tenders, etc. was not possible within the scope of this study. For this reason it is only possible to give estimated costs, which are based on the experience of NIS and on a few specific values. The estimated costs are also determined by the boundary conditions, which are still to be defined. Some of the boundary conditions to be taken into account in the cost estimate, such as ultimate storage of radioactive decommissioning waste, have not yet been sufficiently clarified, so that it is necessary to make reasonable assumptions concerning these.

In what follows estimates are given of the decommissioning costs for the alternatives listed below: (1) interim confinement;

(2) partial dismantlement with secure residual confinement;

(3) total dismantlement; and (4) total dismantlement after 40 years' interim conFine-

ment. The cost estimate serves as the basis for the subsequent economic comparison.

The final status in respect of the individual alternatives has already been adequately defined in section 2; it will not be discussed further here.

The intermediate position of the not very realistic alternative of partial dismantlement with secure residual confinement (see section 2.2.2) becomes clearly evident in the cost data. Depending on the extent of the partial dismantlement, which may vary from plant to plant, the form (and hence also the costs) of decommissioning departs from interim confinement and approximates to total dismantlement. The cost estimate for this 'intermediate alternative' can therefore only indicate an order of magnitude.

5.1. Aspects which cannot be assessed in financial terms

For an evaluation of the decommissioning alternatives, it is necessary to consider not only the decommissioning costs proper but also other aspects which cannot readily be assessed in cash terms.

5.1.1. Financial risk It is not yet possible to predict the likely develop-

ment costs of the special equipment required. How- ever, a large number of special facilities of this kind are required for dealing with the high activities concerned, having regard to the compact design of the steel components and concrete structures, especially for decontamination and size reduction. It is assumed for the purpose of cost estimation that these facilities are available and can be hired, but the development costs are not taken into account.

It is also impossible to calculate the costs arising out of technical difficulties which emerge during the process of decommissioning and were unforesee- able.

Other risks which cannot be assessed in cash terms are: (a) changes in the official authorization practices; (b) accidents in the handling and transport of sources

of activity;


(c) inadvertent release of gaseous or liquid activity (e.g. discharge of decontamination effluent into a river); and

(d) decontamination not feasible to the extent planned, so that the decontamination or disposal costs are increased.

At the present time it is not possible to say to what extent such risks can be covered by insurance. However, these risks also exist with operating nuclear power plants and consequently are not a peculiar feature of decommissioning.

5.1.2. Site As the number of nuclear power plants increases,

and in consequence of other circumstances too, a shortage of sites for new plants may arise. It is an ob- vious course of action to utilize the site of a decommissioned nuclear power plant for a new construction, and this has the additional advantage that a licence already exists. The resulting financial benefit cannot, however, be expressed in cash terms. The economic advantage arising from the sale of the site or from tur- ning it over to a different use after completion of total dismantlement is negligible.

5.1.3. Changes in decommissioning philosophy For the present it must be assumed that decom-

missioning waste must be packed in 200 or 400 1 casks and transported to a final repository (see also section 5.2.3). This means that the active components must be cut up into pieces of manageable size. If there is a t'mal repository capable of accepting larger containers or larger pieces of scrap, the costs of size reduction will be reduced. It is at present impossible to quantify these savings, since the storage conditions are unknown.

5.1.4. New techniques It is assumed that familiar, relatively well-tried

techniques are used in the decommissioning operation. New or insufficiently proven techniques - e.g. the melting down of active components - and any resulting economic advantages are likewise not susceptible of estimation.

5.1.5. Psychological factors The 'wrecks' remaining in the case of interim con-

finement (and possibly also in that of partial disposal

with secure residual confinement) constitute eyesores The pressure of public opinion may mean that in the long term total disposal becomes the only possibility.

5.2. Boundary conditions

The boundary conditions set out in this section apply to all decommissioning alternatives.

5.2.1. Organizational boundary conditions (1) Decommissioning is the consequence of a sche-

duled closure. No major releases of activity within the plant - e.g. due to a serious incident - have taken place.

(2) Between the final shutdown and commencement of the actual decommissioning procedures a period of one year elapses within which time the fuel elements are removed and short-lived radionuclides can decay.

(3) The consent procedures take place in the same way as with the construction of a nuclear power plant. It is assumed that current authorization practice applies.

(4) Only systems and facilities on the immediate site of the power plant are considered (i.e. no trans- mission lines, access roads, etc.).

(5) During the process of decommissioning, ex- perienced operating staff familiar with the plant are available for radiation protection, operation of systems, supervision of dismantling, and surveillance.

(6) Apart from this, decommissioning of the plant is effected wholly by external commercial firms.

(7) Decommissioning is effected primarily with single-shift working, except for the handling of the reactor pressure vessel.

5.2.2. Technical boundary conditions (1) The dismantling operations and, in particular,

the handling of radioactive facilities, are planned in detail in advance and determined in consultation with all parties concerned. Difficult dismantling operations are, if necessary, rehearsed on mockups.

(2) At the beginning of decommissioning, the plant facilities and systems are, as far as necessary, in an operational condition.

(3) The necessary large-scale special equipment is available at the time of decommissioning and can be hired. This applies, in particular, to cutting appara- tus and decontamination and handling facilities.

NIS / Decommissioning of L WR nuclear power plants 3 7

5.2.3. Radiological boundary conditions (1) Before decommissioning commences, the nu-

clear fuel and movable operational activity sources (filters, resins, slurries) are removed from the plant where this is possible and reasonable.

(2) For calculation of the activity inventory from the activation, the basis assumed is an operating time of 40 years with a load factor of 0.75.

(3) Part of the turbine of the BWR cannot be fully decontaminated ° ; the turbine of the PWR is either free of activity or capable of being decontaminated.

The following additional boundary conditions apply to the alternative of total dismantlement:

(1) The buildings are not demolished until the control level area within the buildings concerned is abolished in agreement with the licensing authorities.

(2) A final repository for nuclear power plant decommissioning waste which cannot be disposed of freely does not yet exist. For this reason it is iml~oss- ibly to give realistic costs for the disposal of such waste. To preserve comparability with the other alternatives, it is assumed that the costs of the subsequent ultimate disposal of this waste correspond to those which would be incurred in the current method of final disposal in drums. The present storage conditions of the experimental final repository at Asse, West Germany, constitute the basis of the cost estimate.

The activated masses are made up almost exclusive- ly of the reactor pressure vessel and its internal structures. Under Section 42 of the First Radiation Protec- tion Order (SSVO) now in force (a new draft has been prepared) in the Federal Republic of Germany, the licensing authority may allow solid waste for disposal containing radioactive substances with half-lives of more than 100 days to be treated as ordinary waste if its average specific activity before release does not exceed 10 - s Ci/m 3.

This limit is not exceeded by, for example, the pressure vessel head of a BWR. However, since Section 42 is permissive rather than prescriptive, the pessimistic assumption is made that the entire pressure vessel (the activity of the internals will never be below the limit) must be regarded as waste which cannot be disposed of freely.

(3) Although it may sometimes be possible to de- contaminate contaminated systems, it is usually very difficult to demonstrate that adequate decontamination has been effected. Severe contaminations often cannot be economically eliminated, so that in the final analysis ultimate disposal (in drums) is more cost-effective. For this reason it is assumed in the cost evaluation that all the masses set out in table 18 are to be stored. The decontamination works allowed for in the cost evaluations are confined to those car- fled out to improve accessibility or reduce the radiation load.

5.2.4. Economic boundary conditions (1) The cost basis is the year 1975. (2) The proceeds from the sale of reclaimable fa-

cilities and usable waste (scrap, etc.) are disregarded, since the effect on costs is smaller than the accuracy of the cost estimates.

(3) Development costs for special equipment are substantially disregarded.

(4) The costs of large-scale special equipment and facilities are considered on a rental basis. They are, however, uncertain, since the development costs cannot be estimated.

5.3. Cost estimates

5.3.1. Interim confinement

5.3.1.1. Decommissioning planning. The only costs involved here are those of engineering office work. The estimated workload for this item is 40 man- months (MM). Assuming a rate of 10 000 DM/MM (I0 kDM/MM), the cost of this item works out as 40 MM × 10 kDM/MM = 400 kDM.

5.3.1.2. Ascertainment o f initial situation. Here again only engineers' time is involved, since the measurements and compilation of the inventory lists can ex- pediently be carried out by the station staff. The costs of these operations are included in the operating costs (section 5.3.1.6). The engineering office workload for this item can be estimated at 40 MM:

40 MM X 10 kDM/MM = 400 kDM.

* This assumption is utilized for the cost estimate; possibly also the turbine of the BWR can be decontaminated.

5.3.1.3. Preliminary works and infrastructure. The estimate for engineering office work is 5 MM. Additional

38 NIS / Decommissioning of L lCR nuclear power plants

costs arise for the provision of infrastructures (dose rate measuring facilities, decontamination facilities, additional sanitary facilities for outside staff, etc.):

engineering office work: 5 MM X 10 kDM = 50 kDM

infrastructure = 50 kDM

100 kDM

5.3.1.4. Decontamination. The basis of the 'interim confinement' form of decommissioning is minimiza- tion of the workload and expenditure. This applies also to decontamination. Decontamination is effected only if it will substantially facilitate access for subsequent inspections and if additional contamination has occurred as a result of the decommissioning works.

Hardly any specific costs for the decontamination of complex facilities and buildings have hitherto been published; again, these depend very largely on the individual circumstances. In the US, specific costs of 88 and 540 DM/m 2, respectively, have been reported for the decontamination of a uranium and beryllium plant and of a large hot cell.

The basis assumed for decommissioning investiga- tions for a nuclear power plant in West Germany (the HDR) was 100 DM/m 2. If the area to be decontaminated is taken to be 500 m 2 , the cost requirement under this heading is 50 kDM. A further 50 kDM is estimated for other decontamination work and for disposal of the decontamination media:

5.3.1.6. Operating costs. A part of the station staff is required for decommissioning. On average, 30 people are required during decommissioning (not counting project management). The resulting staff costs amount to some 1350 kDM/year.

The specific plant and materials costs for the operation of the systems required during decommissioning and for the provision of energy and auxiliary materials are assumed to be of the same order as those involved in a refuelling (about 4000 DM/day 2 1460 kDM/yr). These costs are assumed to be incurred for two-thirds of the entire decommissioning period, including 1 year's lead time.

The plant and materials costs for radiation protection also come under the heading of operating costs. The specific plant and materials costs per man-year (MY) in this case amount (according to Swiss experience with Diorit I) to about 1.5 kDM. These costs are also allowed for two-thirds of the entire decommissioning period, including 1 year's lead time. In addition to the station staff, 10 men employed by external contractors are allowed for:

staff costs: 3 yr X 1350 kDM/yr = 4050 kDM

plant and material cost: 2 X 3 yr X 1460 kDM/yr = 2920 kDM

radiation protection costs: ~ X3 yr X40 M × 1.5 kDM/MY = 120 kDM

7090 kDM

decontamination of buildings: 500 m 2 X 100 DM/m 2 remainder

= 50 kDM

= 50 kDM

100 kDM

5.3.1. Z Consent procedures. The costs arising under this heading for engineering office work are already included in section 5.3.1.5. The consent costs are estimated at 800 kDM.

5.3.1.5. Pro/ect management and supervision. It is assumed that the work of interim confmement involves a lead time of 1 yr and a period of execution of 2 yr. The workload for project management and supervision is put at 50 MM:

engineering office work: 50 MM X 10 kDM/MM =

= 500 kDM.

5.3.1.8. Expert's reports and studies. These costs are estimated at 300 kDM.

5.3.2. Partial dismantlement with secure residual confinement

5.3.2.1. Decommissioning planning. As section 5.3.1.1., except that the workload is estimated to be 50 MM:

50 MM X 10 kDM/MM = 500 kDM.

NIS / Decommissioning e l l ICR nuclear power plants 39

5.3.2.2. Ascertainment o f initial situation. As section 5.3.1.2, i.e. 40 MM X 10 kDM/MM = 400 kDM.

5.3.2.3. Preliminary works and infrastructure. The engineering office workload is assumed to be only slight- ly greater than in the case of section 5.3.1.3, amounting to 6 MM (1 MM more). However, the workload and expenditure involved in providing the infrastructure and in improving accessibility to the plant and to the activities within it is substantially greater:

engineering office work: 6 MM X 10 kDM/MM = 60 kDM


accessibility = 150 kDM

400 kDM

5.3.2.4. Handling o f reactor pressure vessel. The mini. mal costs of removing and inserting the internal structures are included in the operating costs (5.3.2.12). The only costs involved are those of cutting off all pipelines and ancillary structures (see section 2.2.2) and sealing off the openings. The costs are estimated at 1000 kDM.

In sections 2.2.2 and 3.3 it was shown that it is sometimes necessary to lower the reactor pressure vessel in the case ofa BWR. This involves substantial costs, which can be estimated as follows:

engineering Office work: 20 MM X 10 kDM/MM = 200 kDM

execution = 2000 kDM

2200 kDM

5.3.2.5. Breaking up o f controlled areas. This item con- cerns the activated and/or contaminated parts of the controlled areas (apart from the pressure vessel). A relatively wide margin of uncertainty attaches to estimates of this operation, since the costs vary considerably from instance to instance. The size of the residual confinement depends largely on individual plant conditions. However, the portion (i.e. the nuclear part) which accounts for the difference between the controlled area and the residual confinement must be broken up (and stored). The following costs are

assumed:

engineering office work:

10 MM × 10 kDM/MM = 100 kDM

execution ~.

1500 Mg X 6 kDM/Mg = 9000 kDM

9100 kDM

3.3.2.6. Decontamination. The volume of work in this case is considerably greater than that described in section 5.3.1.4. The estimate is as follows, in accordance with the specific costs assumed in that section:

decontamination of building: 40 000 m 2 X 100 DM/m 2 = 4000 kDM

decontamination of systems

decontamination of small

parts removed (e.g. = 2000 kDM fittings)

decontamination of tools diposal of decontamination

media

6000 kDM

5.3.2. Z Conventional demolition. According to the quotations obtained for the HDR (West Germany), the specific costs for the demolition of concrete structures are in the region of 450 DM/m 3. The actual costs in the case of the Elk River Reactor were about 200 DM/m 3.

We assume here a figure of 300 DM/m a. The cost of breaking up the nonradioactive steel scrap is in the region of 400 DM/Mg. As to the quantity of waste to

be disposed of, only rough estimates are possible. For this reason, no distinction is made between the BWR and the PWR.

Assumptions:

breaking up of nonactive scrap: 10 000 Mg X 0.4 kDM/Mg = 4 000 kDM

demolition of concrete structures: 50 000 m a X 0.3 kDM/m a = 15 01313 kDM

19 000 kDM


5.3.2.8. Constructural measures. It is estimated that the constructural measures for secure residual confinement (see section 3.3, operation 27) will cost 600 kDM.

5.3.2.13. Experts' reports and studies. These costs are put at 500 kDM.

5.3.3. Total dismantlement

5.3.2. 9. Project management and supervision. It is assumed that the work of partial dismantlement with secure residual confinement requires a lead time of one year and an actual execution time of three years. The project management and supervision workload is estimated to be 200 MM:

5.3.3.1. Decommissioning planning. As 5.3.1.1, but a workload of 70 MM is estimated:

70 MM × 10 kDM/MM = 700 kDM .

5.3.3.2. Ascertainment o f initial situation. As 5.3.1.2, i.e. 40 MM × 10 kDM = 400 kDM.

engineering office work: 200 MM × 10 kDM/MM = 2000 kDM.

5.3.2.10. Special equipment. In view of the high levels of activity to be handled and the fact that many of the steel components and concrete structures are of compact design, a great deal of special equipment is necessary, particularly for decontamination and size reduction. The value of this equipment will probably lie in the region of DM 10 million. However, it is assumed that some of this equipment will already be available for an impending decommissioning project, or that other equipment, to be newly developed, can subsequently be used for work on other reactors; for this reason only half the total value of the special equipment required is taken into account:

special equipment: 5000 kDM.

5.3.2.11. Operating costs. As in 5.3.1.6, but:

station staff: 50 men = 2250 kDM/year

outside staff: 50 men

staff costs: 4 yr X 2250 kDM/yr =

plant and materials costs: 2 X4 yr X 1460 kDM/yr =

radiation protection equipment: X4 yr X 100 M × 1.5 kDM/Myr =

9 000 kDM

3 900 kDM

400 kDM

13 300 kDM

5.3.2.12. Consent procedures. The cost of obtaining the necessary consents is estimated at 4000 kDM.

5.3.3.3. Preliminary works and infrastructure. As 5.3.2.3 but with the following values:

engineering office work: 7MMX10kDM/MM = 70kDM


access = 250 kDM

570 kDM

5.3.3.4. Handling and breaking up o f a reactor pressure vessel. The costs are broken down under the two headings of engineering office work and execution. In the case of the quotations for the HDR, the engineering office costs ranged between DM 0 • 4 and 0 • 6 million for this item. The equivalent costs for the Elk River Reactor (ERR)were put at about $350 000. The costs for 1000 MW light-water reactors were estimated in 1973 by Gulf United'Nuclear Fuels Corpora- tion (GNF) at $270 000. Since the LWRs in question have higher ratings and will probably have operated for longer periods so that the activity inventory is higher, it is not unreasonable to assume engineering office costs of 1200 kDM. The sum corresponds to an engineering office workload of 120 man-months.

The execution costs can best be related to the mass of the reactor pressure vessel to be dealt with and its internal structures. According to US information, the specific costs here are in the region of 2 • 6 kDM/Mg. The quotations for the HDR showed values ranging between 8 and 18 kDM/Mg. The figures from the US and for the HDR are based on considerably lower specific activities for the pressure vessel than in the case of the LWRs with which we are concerned after 40 years' operation.


Moreover, in the case of the considerably lower US values, the parts of the pressure vessel did not have to be broken down to drum size. Hence the specific costs for execution of dismantling and size reduction will amount to at least 30 kDM/Mg.

We thus obtain the following costs for dealing with the reactor pressure vessel:

BWR engineering office work: 120 MM X 10 kDM/MM = 1 200 kDM

execution: 1,030 Mg X 30 kDM/Mg = 30 900 kDM

PWR engineering office work: 120 MM × 10 kDM/MM

execution: 760 Mg X 30 kDM/Mg

32 100 kDM

= 1 200 kDM

= 22 800 kDM

24 000 kDM

5.3.3.5. Breaking up o f controlled areas. This con- cerns the activated and/or contaminated parts of the controlled areas (except the pressure vessel). The engineering office workload for this will be about the same as for the handling of the reactor pressure vessel.

According to US experience (with the ERR), the specific costs for execution of the breaking-up operations are in the region of 1 • 6 kDM/Mg. This figure covers both contaminated and uncontaminated components, including engineering. Note that the parts concerned did not have to be broken to drum size. The estimated specific costs for the HDR ranged between 5 and 11 kDM/Mg. In the case of the LWRs to be considered, allowance must be made for the higher contamination, in particular of the major components. The costs are estimated at 20 kDM/Mg for highly contaminated parts and 6 kDM/Mg for other contaminated parts. According to US experience (ERR), the specific costs of breaking up the concrete ring of the shield cooler (approximately 180 m 3) into blocks are about 750 DM/m 3. The figure assumed for reduction to drum size is 1000 DM/m 3 .

We assume the following quantities: contaminated steel: 3000 Mg; highly contaminated steel: 300 Mg; concrete: 180 m 3.

We thus obtain the following costs:

engineering office work: 120 MM X 10 kDM/MM

execution - steel: 300 Mg X 20 kDM/Mg

execution - steel: 3,000 Mg × 6 kDM/Mg

execution - concrete: 180 m 3 X 1000 DM/m 3

= 1 200 kDM

= 6 000 kDM

= 18 000 kDM

= 180 kDM

25 380 kDM

5.3.3.6. Decontamination. Assuming the same specific costs as in section 5.3.1.4, the estimate is as follows:

decontamination of buildings: 60,000 m 2 X 100 DM/m ~

decontamination of systems decontamination of small parts

removed (e.g. fittings) decontamination of tools disposal of decontamination

media

= 6 000 kDM

= 4 000 kDM

10 000 kDM

5.3.3. 7. Packing and storage. The approximate quantities (Mg) that go for ultimate disposal are given in table 20.

The distinction between low-, medium- and high- level items is a handling criterion. It is assumed that the breakdown given in table 20 corresponds to the final storage criteria. A different classification may sometimes be possible through cost optimization (dilution effects), as shown in table 21. In the case of low-level final disposal, the maximum permissible dose rate is not attained owing to the limitation of volume or weight. With medium-level final disposal, however, the limit is given by the total activity per drum. The maximum permissible limit for the dose rate is observed at 5000 Ci for the materials X5CrNi189 and X10CrNiNb 189, which are virtually the only materials concerned. Table 22 gives the overall total costs.

Total BWR: 39 204 kDM,

PWR: 12 257 kDM.


Table 20 Approximate quantities for disposal (Mg)

Low-level Medium-level High-level

Contaminated * Activated

BWR 4450 565 + 371 92 430

PWR 3350 288 + 361 110 430

* Items which cannot be decontaminated.

It is again emphasized that these costs are estimated on the high side. They can probably be substantially reduced by the construction of a final repository which does not require storage in drums.

5.3.3.8. Conventional demolition. As 5.3.2.7, but the following assumptions are made:

breaking up of residual nonactive scrap: 11 000 Mg × 0 • 4 kDM/Mg = 4 400 kDM

demolit ion of concrete structures, including foundations: 120 000 m a X 0 . 3 kDM/m 3 = 36 000 kDM

5.3.3. 9. Earthworks. Filling in of building excavations:

130 000 m 3 X20 DM/m 3 2600 kDM

Removal of buried conduits and lines: 200 kDM

Levelling of site: 5 0 0 0 0 m 2 X4 DM/m 2 200 kDM

3000 kDM

5.3.3.10. Project management and supervision. It is assumed that total dismantlement would require a lead time of i yr and a period of execution of 5 - 6 yr. The workload for project management and super° vision is estimated at 350 MM:

engineering office work: 350 MM X 10 kDM/MM = 3500 kDM.

5.3.3.11. Special equipment. As 5.3.2.10, but because the pressure vessel has to be broken up and on account o f other additional work, the value of the special equipment is estimated at DM 20 million. Special equipment: 10 000 kDM

5.3.3.12. Operating costs. As 5.3.1.6, but:

station staff: 60 men = 2700 kDM/yr

outside staff: 60 men

40 400 kDM Results:

Table 21 Cost optimization

Storage criteria Low-level Low-level Medium-level without shielding with shielding

Type of cask 400 1 drum without 400 1 drum with 200 1 drum shielding cast-in 200 1

liner (concrete)

Max. quantity to be stored 1.250 kg 700 kg 5.000 Ci

Cost (DM) * 900 3.000 5.000

* These figures comprise the costs of packing, transport, and storage.


Table 22 Total costs (kDM)

Low-level Medium-level

Without With 700 kg shielding shielding

5000 Ci criterion

drums 3560 + 100 * 1420 530 5800 BWR

costs 3294 4260 2650 29000

drums 2680 + 100 * 1026 516 820 PWR

costs 2502 3077 2580 4100

* Secondary waste is estimated to amount to 100 drums for each instance of total dismantlement. This is made up of waste arising during decommissioning which cannot be adequately decontaminated, e.g. contaminated tools, air filters, working clothing, tar- paulins, etc.

staff costs:

7 yr X 2700 kDM/yr = 18 900 kDM

plant and materials costs:

X7 yr X 1460 kDM/yr = 6 814 kDM 3

radiation protect ion equipment: 2_ X 7 yr X 120 X 1 • 5 kDM/M yr = 840 kDM 3

26 554 kDM

3.3.3.13. Consent procedures. The cost of obtaining the necessary consents is est imated at 6000 kDM.

3.3.3.14. Experts" reports and studies. These costs are put at 800 kDM.

3.3.4. Total dismantlement after 40 years' prior interim confinement

5.3.4.1. Decommissioning planning. As 5.3.3.1, but the workload is est imated to be 40 MM:

40 MM X 10 kDM/MM = 400 kDM.

5.3.4.2. Ascertainment o f initial situation. As 5.3.3.2, i.e. 10 MM X 10 kDM/MM = 100 kDM.

5.3.4.3. Preliminary works and infrastructure. As 5.3.3.3, but with the following figures:

engineering office work: 6 MM X 10 kDM/MM

infrastructure

access

= 60 kDM

= 200 kDM

= 250 kDM

510 kDM

5.3.4.4. Handling of reactor pressure vessel As 5.3.3.4, but the activity is lower so that the specific costs are also reduced:

BWR/engineering office work: 100 MM X 10 kDM/MM = 1 000 kDM


PWR/engineering office work: 100 MM X 10 kDM/MM


= 20 600 kDM

21 600 kDM

= 1 000 kDM

= 15 200 kDM

16 200 kDM

5.3. 4. 5. Breaking up of controlled areas. As 5.3.3.5, but based on the following quantities: contaminated


steel: 3200 Mg; highly contaminated steel: 100 Mg; Concrete 150 m 3. The following costs were assumed: highly contaminated parts: 20 kDM/Mg; other contaminated parts: 5 kDM/Mg; concrete: 1000 DM/m 3.

The following costs result:

engineering office work: BWR 4000 100 MM × 10 kDM/MM = 1 000 kDM

excecution - steel PWR 3000

100 Mg ×20 kDM/Mg = 2 000 kDM 3,200 Mg × 5 kDM/Mg = 16 000 kDM

-concrete: 150m3 X 1000 DM/m 3 '= 100 kDM

19 100 kDM

3.3.4. 6. Decontamination. It is assumed that the workload is less than that mentioned in section 5.3.3.6, because of the 40 year cooling period. The estimate is as follows:

decontamination of buildings 30 000 m 2 X 100 DM/m 2 = 3000 kDM

decontamination of systems decontamination of parts = 2000 kDM decontamination of tools disposal of decontamination media

5000 kDM

5.3.4. Z Packing and storage. As 5.3.3.7, except that because of the 40-year cooling period the relative proportions of high, medium- and low-level material have changed. The quantities that are to be disposed of (Mg) are given in table 23, and the resulting costs (kDM) are given in table 24.

Total: BWR 11 435 kDM,

PWR 7 994 kDM.

5.3.4.8. Conventional demolition. As 5.3.3.8, but the following figures are assumed:

breaking up of residual scrap: 11 500 Mg X 0 • 4 kDM/Mg = 4 600 kDM

demolition of concrete structures, including foundations: 120 000 m 3 X 0 . 3 kDM/Mg = 36 000 kDM

40 600 kDM

Table 23 Quantities for disposal (Mg)

Low-level Mediumqevel High-level

Contaminated * Activated

725 + 240 63 430 531 + 121 106 430

* Quantities which cannot be decontaminated (estimated values).

5.3.4.9. Earthworks. As 5.3.3.9, i.e. 3000 kDM.

5.3.4.10. Project management and supervision. As 5.3.3.10, but assuming:

engineering office work 300 MM X 10 kDM/MM = 3000 kDM.

5.3.4.11. Special equipment. Since it is impossible to make a reasonable estimate of the reduced worldoad due to the decline in activity, the same figure is as. sumed as in 5.3.3.11, i.e. special equipment: 10 000 kDM.

5.3.4.12. Operating costs. It is assumed that after 40 years no qualified station staff is available. In the estimates of these costs for the decommissioning alternatives considered so far, the costs of outside staff required are included under the relevant headings. It is assumed that an outside staff of 20 persons is available to assist with the project management.

Results:

staff costs: 6 yr X 900 kDM/yr =

plant and materials costs: 2 X 6 yr X 1460 kDM/yr =

radiation protection equipment × 6 y r X 7 0 M X 1 . S k D M / M y r =

5 400 kDM

5 900 kDM

200 kDM

11 500 kDM

5.3.4.13. Consent procedures. The cost of the consent procedures is estimated to amount to 5000 kDM.


Table 24 Total costs (kDM)

Low-level Medium-level

Without With 700 kg 5000 Ci criterion shielding shielding

drums 3200 + 100 1650 343 360 BWR

costs 2970 4950 1715 1800

drums 2400 + 100 1373 173 + 152 * PWR

costs 2250 4119 1625

* The 5000 Ci/drum criterion is not applicable here, because the 700 kg limit would be exceeded.

5. 3. 4.14. Experts' reports and studies. These costs are

e s t ima ted at 6 0 0 kDM.

5.3.5. Summary o f costs A s u m m a r y o f costs is p r e s e n t e d in t ab le 25 .

5.4. Surveillance requirement for decommissioned plants

The survei l lance r e q u i r e m e n t for a d e c o m m i s s i o n e d

p l an t d e p e n d s o n t he c o n d i t i o n o f the p l a n t or o n the

decommis s ion ing c o n f i g u r a t i o n chosen . The surveil lan-

ce r e q u i r e m e n t for the t w o basic a l te rna t ives (a) in-

Table 25. Costs in thousands of DM(kDM), rounded up

I II II III Ill IV IV BWR/PWR BWR PWR BWR PWR BWR PWR

1 Decommissioning planning 400 500 500 700 700 400 400 2 Ascertainment of initial situation 400 400 400 400 400 100 100 3 Preliminary works and infrastructure 100 400 400 600 600 500 500 4 Handling of reactor pressure vessel - 3 200 1 000 32 100 24 000 21 600 16 200 5 Breaking up of controlled area - 9 100 9 100 25 400 25 400 19 100 19 100 6 Decontamination 100 6 000 6 000 10 000 10 000 5 000 5 000 7 Packing and storage - - - 39 200 12 300 11 500 8 000 8 Conventional demolition - 19 000 19 000 40 400 40 400 40 600 40 600 9 Constructional measures - 600 600 . . . .

10 Earthworks - - - 3 000 3 000 3 000 3 000 11 Project management and supervision 500 2 000 2 000 3 500 3 500 3 000 3 000 12 Special equipment - 5 000 5 000 10 000 10 000 10 000 10 000 13 Operating costs 7 100 13 300 13 300 26 600 26 600 11 500 11 500 14 Consent procedures 800 4 000 4 000 6 000 6 000 5 000 5 000 15 Experts' reports and studies 300 500 500 800 800 600 600

Sub total 9 700 64 000 61 800 198 700 163 700 131 900 123 000 Contingencies 1 900 12 800 12 300 39 700 32 700 26 400 24 600

(approx. 20%) Total 11 600 76 800 74 100 238 400 196 400 158 300 147 600

Decommissioning alternatives: I: Interim confinement II: Partial dismantlement with escure residual continement

III: Total dismantlement IV: After 40 years III (following I)


terim confinement, and (b) partial dismantlement with secure residual confinement, is discussed in what follows. The third basic alternative - total dismantlement - is not considered because no surveillance is necessary in this case. Combined decommissioning alterna- fives are not treated separately; they are made up of the basic alternatives already described. However, the alternative of conversion of the nuclear power plant into a final repository for the storage of radioactive waste is considered.

It is assumed that the licensing authority does not demand continuous surveillance: Continuous surveillance involves costs of the order of 0 • 5 million DM/yr.

5.4.1. Interim confinement The final condition with this decommissioning

configuration is defined in section 2.2.1. With this type of decommissioning, the integrity of the containment must be preserved, i.e. the main surveillance task is to ensure that this requirement is met.

The staff necessary for this purpose must therefore carry out regular checks, inspections and measurements (dose rate, aerosol activity, etc.), and must monitor the condition of the plant and carry out maintenance work. In the period immediately after decommissioning, these checks take place at relatively short intervals (possibly monthly); later, the interval can be extended (quarterly or half-yearly). To prevent un- authorized access to the plant, an electronic alarm system connected to the nearest police station is required.

Any more extensive surveillance (e.g. by a security firm) can have no more than symbolic significance; it depends on the requirements stipulated by the authorities in the individual country concerned. If there is another nuclear power plant on the same site, security can be provided by its staff. The following is an estimate of surveillance costs:

surveillance, 3 persons 4 times a year for 10 days = 120 6Y6-6

maintenance, 3 persons once a year for 5 days _ I s - ~ g ' 6

las X60 000 DM/yr = 22 500 DM/yr

electricity, water and drainage = 5 000 DM/yr

material and other items = 20 000 DM/yr

47 500 DM/yr

5. 4. 2. Partial dismantlement with secure residual confinement

The final condition with this form of decommissioning is defined in section 2.2.2. The surveillance requirement with this alternative is similar to that for interim confinement. However, no maintenance work is carried out, so that less frequent inspections are called for. Surveillance is limited to checking of the confinement structure, measurement of dose rate and aerosol activity, and documentation of these inspections. If necessary, the water (condensation) collecting at the bottom of the confinement can be pumped off by a contractor. The following is an estimate of surveillance costs:

surveillance, 3 persons twice yearly for 10 days =

6o X 60 000 DM/yr = 10 000 DM/yr

electricity, drainage, etc. = 6 000 DM/yr

16 000 DM/yr

5.4.3. Conversion o f nuclear power plant into a final repository for the storage o f active waste

In the final condition with this form of decommissioning, the activated and contaminated waste from the decommissioning of a number of nuclear power plants is stored in a decommissioned plant specially converted for this purpose.

The surveillance requirement is greater than in the case of the alternatives already discussed. Additional maintenance work is necessary for the heat evacuation facilities to be installed. Since waste from a number of nuclear power plants is stored here, it is practical to group the components which give off the most heat (components close to the core such as the lower and upper grid plates and core barrel or shroud) together in a single room, in which cooling facilities can be installed. These cooling facilities require maintenance, thus again necessitating more frequent inspections.

For the purpose of the cost estimate it is assumed that the cooling facilities incorporate sufficient re- dundancy or that there is a nuclear power plant in the vicinity which can, if necessary, provide sufficient cooling on a short-term basis. In addition, as with the alternatives already discussed, dose rate measurements, checking of the confinement structure, and documen-


tation of the measurements taken are required. The costs given below are not comparable with

those quoted for the two basic alternatives. Whilst they also refer to a form of decommissioning, they do not relate to a single decommissioned nuclear power plant, since waste from a number of plants is stored in these facilities. The following is an estimate of surveillance costs:

surveillance, 3 persons 4 times per year for 10 days

_ 120

maintenance, 3 persons twice a year for 5 days = ao Yg'6

150 JC6 X 60 000 DM/yr = 25 000 DM/yr

electricity, water, drainage = 20 000 DM/yr

material and other items = 20 000 DM/yr

65 000 DM/yr

~ so

<

40

$0

20

10

o Io 2o 30 4o so eo m go go 10o

Cooling time (years)

Fig. 11. Decrease in the activity inventory o f a 1200 MW nuclear power plant with t ime.

6. Economic comparison of the decommissioning alternatives

To compare the costs of the decommissioning alternatives to be considered, an economic calculation using the cash value method is carried out. The results are intended to permit an initial appraisal of the alternatives. Economically relevant criteria which depend on individual circumstances, such as site utilization, licensing constraints, and possibilities of ultimate storage, are not covered by the calculation.

6.1. Boundary conditions

6.1.1. General boundary conditions

For the purposes of the economic calculation, the following assumptions are made based on current values applicable to power plant operators: (a) rate of cost increase (wages and materials costs) 8%, and (b) interest rate 12%. It is assumed that after 40 years the alternative of interim confinement is converted into that of total disposal, because, as fig. 11 shows, there is no further appreciable reduction in the activity inventory after this time has elapsed. For this reason, further postponement of ultimate disposal by several decades will afford no further advantage. For the

purpose of economic comparison of the alternatives, the same 40-year period is taken as the basis, even if conversion into another decommissioning configuration does not take place or is not planned. The figures given in section 5 are taken as the basis for the decommissioning, surveillance and inspection costs. The remaining boundary conditions are: (a) cost advantages accruing from the reclamation of

components or fittings are not taken into account; (b) the licensing constraints for the period after de-

commissioning have no greater effect on costs than is at present foreseeable; and

(c) the costs of permanent surveillance of radioactive waste in a final repository are included as a lump sum in the decommissioning costs.

6.1.2. Boundary conditions specific to individual alternatives

6.1.2.1. Interim confinement followed after 40 years by dismantlement or by partial dismantlement with secure residual confinement. Interim con-

finement is a temporary expedient; the decommissioning process is concluded when, after 40 years have elapsed, either the plant is totally dismantled or partial dismantlement with secure residual confinement

4 8 NIS / Decommissioning of L WR nuclear power plants

is carried out. It is assumed that large-scale conver- sions or repairs are unnecessary during the 40-year surveillance period. Surveillance costs arising subsequent to the concluding residual confinement are not taken into account.

6.1.2.2. Total dismantlement. After complet ion of decommissioning, no further costs arise.

6.1.3. Boundary conditions for calculation of cash values

On the basis of these boundary conditions and of the information given in section 5, the following decommissioning alternatives are subject to an economic comparison by the cash value method. First, interim confinement followed by total dismantlement after 40 years, compared with immediate total dismantlement; in both cases, the BWR and PWR are considered separately. A comparison with the alternative of partial dismantlement with secure residual confinement is dispensed with (see section 5). The cost information in table 25 has comparative significance only, since it relates to a single year. However, a decom-

missioning alternative extends over a number of years, so that these costs must be apport ioned among the relevant number of decommissioning years. The rate of cost increase and the interest rate apply from one year to the next. The number of years required for decommissioning is as follows (some of these figures have already been given in section 5): interim confinement, 3 years; partial dismantlement with secure residual confinement, 4 years; total dismantlement after prior interim confinement (40-year interval), 6 years; total dismantlement, 7 years.

Table 26 apportions the costs from table 25 among

the decommissioning years, the rates of cost increase and interest being disregarded.

6.2. Calculation of cash values

Cash values are calculated by the following formu- lae:

rate of cost increase p = 8%:

q~= 1 ~(1--~) n " qn=l.O8n" ' k '

Table 26 Assignment of the costs given in table 25 to the decommissioning years

Interim confinement followed by Partial dismantlement with Total dismantlement total disposal after 40 years secure residual confinement

Year Costs [kDM] Year Costs kDM Year Costs BWR PWR BWR

[kDM] PWR

1 3 600 3 600 1 4 400 4 400 1 7 700 2 3 900 3 900 2 24 500 23 000 2 28 900 3 4 100 4 100 3 28 300 27 700 3 32 000

4 19 600 19 000 4 36 500 11 600 11 600 5 37 100

X 76 800 74 100 6 33 600 1(40) 6 100 6 100 7 62 600 2(41) 21 100 18 100 3(42) 22 700 19 700 ~ 238 400 196 400 4(43) 28 700 25 700 5(44) 25 100 24 200 6(45) 54 600 53 800

158 300 147 600

7 700 21 100 24 100 28 700 29 500 28 000 57 300

These values are based on an estimated distribution of the costs of the individual groups of operations over the years of decommissioning. Together with the surveillance costs (Section 5.4), they form the basis of the costs comparison by the cash value method set out in the following pages.


Table 27 Cash value of decommissioning costs (interim cofifinement)

Year Costs n(yr) Factor. (I) K × qn K(kDM) qn

1 3600 0 1.00 3600

2 3900 I 0.96 3761

3 4100 2 0.93 3812

Total I 11 173

rate of interest p = 12%:

1 1 1 1

q~b [1 +(p/ lO0)] n ' q~tb 1"12n"

Quotient allowing for both cost increase and interest. rates:

. 1 . qn /1"08~ n

q" = qT, q b' = iSi 1

6.2.1. Interim confinement o f a BWR with total dismantlement after 40 years

A summary of cash values of decommissioning costs for interim confinement and total dismantlement are given in tables 27 and 28, respectively. The cash value of surveillance costs is as follows:

annual costs

costs in fourth year after commencement of decommissioning

Kl = 48 kDM,

K4 = KI Xq~

= 4 8 × 1 .083 = 6 0 k D M .

Cash value in fourth year with surveillance up to the fortieth year after commencement of decommissioning:

surveillance period n = 37 years

cash value B4 = K4 - -

Cash value at commencement of decommissioning

1 - qn

1- -q

1 - (1.08/1.12) 3,7 B4 = 60

1 - ( 1 . 0 8 / 1 . 1 2 ) '

B4 = 1243 kDM,

1 B1 = B4 q---~b,

1 B 1 = 1243 ~ .

Total III = BI = 884 kDM.

Total cash value = totals I + II + III = 44 862 kDM.

6.2.2. Partial dismantlement with secure residual confinement for a BWR

A summary of the cash value of decommissioning costs is given in table 29.

6.2.3. Total dismantlement o f a BWR A summary of the cash value of decommissioning

costs is given in table 30.

6.2.4. Interim confinement o f a PWR with total dismantlement after 40 years

The cash value of interim confinement is as given in section 6.2.1, i.e. total I, 11 173 kDM; the cash

value of surveillance costs is as given in section 6.2.1, i.e total III, 884 kDM. A summary of the cash value of

Table 28 Cash value of decommissioning costs (total dismantlement)

Year Costs n(yr) Factor (II) K X q~" K(kDM) q n

1 6 100 40 0.23 1 424 2 21 100 41 0.23 4 750 3 22 700 42 0.22 4 928 4 28 700 43 0.21 6 008 5 25 100 44 0.20 5 067 6 54 600 45 0.19 10 628

Table 29 Cash value of decommissioning costs (partial dismantlemenO

Year Costs n(yr) Factor K × qn K(kDM) qn

1 4 400 0 1.00 4 400 2 24 500 1 0.96 23 6 2 5 3 28 300 2 0.93 26 315 4 19 600 3 0.90 17 574

Total II 32 805 Total 71 914


Table 30 Cash value of decommissioning costs (total dismantlement)

Year Costs n(yr) Factor K X qn K(kDM) qn

1 7 700 0 1.00 7 700 2 28 900 1 0.96 27 868 3 32 000 2 0.93 29 755 4 36 500 3 0.90 32 727 5 37 100 4 0.86 32 077 6 33 600 5 0.83 28 014 7 62 600 6 0.80 50 328

Table 32 Cash value of decommissioning costs

Year Costs n Factor K X qn K(kDM) (yr) qn

1 4 400 0 1.00 4 400 2 23 000 1 0.96 22 178 3 27 700 2 0.93 25 757 4 19 000 3 0.90 17 036

Total 69 371

Total 208 469

total dismantlement is given in table 31. Total cash value = totals I + II + III = 42 570 kDM.

6.2.5. Partial dismantlement with secure residual confinement for a PWR

A summary of the cash value of decommissioning costs is given in table 32.

6.2.6. Total dismantlement for a PWR A summary of the cash value of decommissioning

costs is given in table 33.

6.3. Comparison of costs

expensive. Even if, at the time of an impending total dismantlement, technologies are available which permit cheaper disposal, Alternative 1 will still be the cheaper solution. Unless there are technical or other reasons (see also section 5.2) for an immediate total dismantlement, Alternative 1 is to be preferred.

The cash values depend largely on the difference between the rates of cost increase and interest. The less this difference the smaller the difference between the cash values of the three alternatives. However, the trend remains the same unless the interest rate falls below the rate of cost increase. But this appears extremely unrealistic and is unlikely to be the case even in the remote future.

A summary of cash values from section 6.2 (kDM) is given in table 34. A comparison o f the cash values with the rates o f cost increase and interest assumed in section 6.2 shows that alternative 1 is the cheaper solution. Alternative 3 is more than four times as

7. Proposals for the design of future nuclear power plants with a view to their decommissioning

The studies made so far have shown that no fun- damental difficulties with the decommissioning are

Table 31 Cash value of total dismantlement

Year Costs n(yr) Factor K X qn K(kDM) qn

I 6 lO0 40 0.23 1 424 2 18 I00 41 0.23 4 075 3 19 700 42 0.22 4 277 4 25 700 43 0.21 5 380 5 24 200 44 0.20 4 885 6 53 800 45 0.19 I0 472

Table 33 Cash value of decommissioning costs

Ye~ Costs n(yr) Factor K X qn K(kDM) qn

I 7 700 0 1.00 7 700 2 21 I00 I 0.96 20 346 3 24 100 2 0.93 22 409 4 28 700 3 0.90 25 733 5 29 500 4 0.86 25 506 6 28 000 5 0.83 23 345 7 57 300 6 0.80 46 067

Total II 30 513 Total 171 106


Table 34 Summary of cash values from section 6.2

BWR PWR

Alternative I: Interim conf'mement, followed by total dismantlement after 40 years

Alternative 2: Partial dismantlement with secure residual confinement

Alternative 3: Total dismantlement

44 862 42 570

71 914 69 371

208 469 171 106

perceptible which would give a reason for changing the basic design of present nuclear power plants. However, with regard to a number of identified difficulties in detail, requiring costly measures, improve- ments are conceivable, as is pointed out in the following.

One way of reducing the decommissioning workload, and hence the costs involved, is to make concrete structures from prefabricated parts where the statics so allow. This would make it possible to re- move some prefabricated parts before or during disposal of the major components (pressure vessel, steam generators, etc.), in order to provide sufficient room for the handling of these components. This would also substantially reduce the amount of work involved in the breaking up of concrete structures.

In the disposal of the biological shield, there is the additional difficulty that the concrete is radioactive. Here again construction from prefabricated segments, where possible, could simplify disposal.

However, the extent to which the safe design of nuclear power plants will allow the use of prefa-

bricated parts, e.g. with regard to shaking of the structure due to external impact (explosions, air- plane crash) and also stability in the case of earth- quakes, will have to be examined carefully.

Problems also arise in the work of decontamination. This work is made possible if there is sufficient room for installation of the decontamination facilities. To limit the amount of decontamination work necessary, 'dead angles' should be avoided both in buildings and in systems and components, i.e. the opportunities for the depositing of contamination should be minimized. The decontamination workload can also be kept down by spatial separation of systems which may become contaminated during operation from 'clean' systems. These aspects are already increasingly observed at modern nuclear power plants, particularly with regard to the facilitation of repair works, which ultimately serves also to aid the decommissioning.

In order substantially to avoid long cooling times, the use of materials leading to nuclides with long half. lives should at any rate be minimized. For the purposes of the present study, it has necessarily been as. sumed that radioactive material is packed in drums and taken away. When a large nuclear power plant is about to be totally dismantled, a final repository should already be available in which large, e.g. container-sized, items can be stored. There should be enough room inside the reactor building to load these containers.

References

[1] H. Ramdohr, Kerntechnik 11. Jahrgang (1969) no. 5. [2] Th. Rockwell IIl, Reactor Shielding Design Manual

(McGraw-Hill Book'Company, 1956). [3] W. Ahlfinger, G. Herbsleb and G. Resch, Dekontamination

nuklearer Anlagenteile (VGB-Speisewassertagung 1971). [4] N. Eickelpasch, Aktennotiz 540/73, Kernkraftwerk

RWE-Bayernwerk (1973).

Documents

Decommissioning of light-water reactor nuclear power plants