Nuclear-safety problems are examined and the results of investigations of nuclear safety of storage sites are

presented for spent nuclear fuel from nuclear power plants. The initial events of anticipated and

unanticipated accidents, methods and errors in the calculation of keff taking account of burnup to ensure

nuclear safety, the possibility of measuring keff of storage sites experimentally, and new forms of fuel with

a consummable absorber are calculated.

At the present time, no more than 5% of accumulated and no more 14% of the annually produced spent nuclear fuel

from nuclear power plants are reprocessed. Consequently, there is a need for large-capacity storage that meets the safety

requirements.

Storage for RBMK-1000, VVÉR-440 and -1000, BN-600, and ÉGP-6 spent nuclear fuel was designed in the

1960s–1970s. In the last few years, when safety requirements changed together with an actual change in the concept of a

closed fuel cycle, compacted storage of fuel has become necessary. This is due, first and foremost, to the need for increasing

the storage capacity, since RBMK-1000, ÉGP-6, and VVÉR-1000 fuel is not reprocessed. Another reason for increasing

capacity is the need to have reserve locations for off-loading at any time a damaged core or a part of such a core in accor-

dance with PNAÉ G-14-029-91 [1]. New safety requirements have made it necessary to analyze an extensive list of emergen-

cy situations including unanticipated accidents. The rules are now being reworked. In this connection it is important to analyze

the concepts on which the rules and the problems appearing during handling of spent fuel and ways to solve them are based.

RBMK Storage. At present RBMK fuel is first stored underwater in holding pools near the reactors and then in

long-term storage sites. In the initial design the spent fuel assemblies in storage sites near the reactors must be arranged with

a spacing of 250 × 160 mm, and the spacing for long-term storage must be 230 × 110 mm. Since the 1980s fuel assemblies

have been stored in a more closely packed arrangement, first in storage sites near the reactors because long-term storage sites

were not available, and then after several years of operation they were also stored in long-term storage sites. At present spent

fuel assemblies at the Smolensk nuclear power plant are packed uncased in sites near the reactors with a triangular-lattice

spacing 130 × 130 × 130 mm, and in long-term storage sites with a design spacing of 230 × 110 mm they are stored in

110 × 2 mm in diameter cases. At the Kursk nuclear power plant a design has been developed for packing fuel assemblies

with a spacing 125 × 139 × 139 mm in storage sites near the reactors, and in long-term storage sites closer packing is used

in which 40 fuel assemblies are placed in each gap between the beams of the compartment cover with two assemblies per

suspension. The spacing between the cases on the suspension is 110 mm and the spacing between the suspensions along the

gap is 130 mm. At the Leningrad nuclear power plant most spent fuel assemblies in the holding pools near the reactor are

packed uncased with spacing 90 × 250 mm, and in long-term storage sites encased fuel assemblies are stored with the max-

imum density with spacing 110 × 115 mm. Considering the state of the cases, which determine the storage time of spent fuel

assemblies in the water as 30 yr, in 2003–2012 it will be necessary to switch to dry storage in steel or reinforced concrete

Atomic Energy, Vol. 91, No. 4, 2001

PROBLEMS AND EXPERIENCE IN GUARANTEEING

THE NUCLEAR SAFETY OF THE STORAGE OF

SPENT FUEL FROM NUCLEAR POWER PLANTS

V. S. Vnukov and B. G. Ryazanov UDC 621.039.546.3

State Science Center of the Russian Federation – A. I. Leipunskii Physics and Power Engineering Institute. Translated

from Atomnaya Énergiya, Vol. 91, No. 4, pp. 263–272, October, 2001. Original article submitted April 9, 2001.

1063-4258/01/9104-0801$25.00 ©2001 Plenum Publishing Corporation 801

containers. The general problem for old storage sites is to prove the nuclear safety on switching to uranium–erbium fuel with

up to 3% 235U enrichment. With long-term storage sites operating for 40 yr, by 2020–2035 the fuel will have to be removed

from them and they will have to be decommissioned. By this time the fuel will have to be reloaded into containers or trans-

ported to an RT-2 plant for specially designed dry storage or new storage, currently under construction,at the RT-1 plant.

VVÉR-1000 Storage. Spent VVÉR-1000 fuel is stored on racks with a 400 mm spacing. Racks with a closer spac-

ing 300 mm for spent fuel assemblies in boron steel casings have been developed and installed for individual nuclear power

plants. This arrangement is used because unoccupied space must be available for situations where it is impossible to transport

assemblies in strictly determined time intervals or for emergency unloading of a reactor and nuclear safety in the case of acci-

dents associated with a decrease in water density in storage sites,though for such accidents nuclear safety is obtained with ordi-

narily racks with a 400 mm spacing. An additional safety measure is the presence of 16 g/liter of boric acid in the water. Dry

storage of spent fuel assemblies in TK-13 containers is also available in order to decrease the dependence on the fuel shipment

times from a nuclear power plant to a storage site at the RT-2 plant. The possibility that fuel assemblies with uranium–gadolin-

ium fuel and zirconium,instead of steel,spacing lattices appear in the storage sites must be taken into account.

VVÉR-440 Storage. The holding pool for spent VVÉR-440 fuel assemblies consists of two compartments.

Two-level racks, in which fuel assemblies with up to 3.6% enrichment are packed with a spacing of 225 mm,are present in

one compartment and the other compartment (a universal cradle) is intended for holding a 30-seat casing with fuel assem-

blies or a TK-6 shipment container. Additional safety measures have been adopted for storing 4.4% enrichment fuel:this fuel

alternates with lower-enrichment fuel. Methods have been proposed for increasing the storage density in the racks with spac-

ing 160–170 mm between fuel assemblies with 4.4% enrichment in boron steel cases,but these methods have not been imple-

mented in practice. The appearance of new types of fuel assemblies with uranium–gadolinium fuel and zirconium spacing

lattices requires additional analysis of nuclear safety. An additional safety measure is a homogeneous neutron absorber con-

sisting of a 12–16 g/liter boric acid solution.

BN-600 Storage. It consists of three compartments,where fuel is stored in 35- and 28-seat casings without or with

containers (unsealed fuel assemblies). At the present time, since such fuel is reprocessed, there is no need for closer packing.

The presence of a large number of experimental fuel assemblies with mixed uranium–plutonium fuel must be taken into

account when analyzing nuclear safety.

Bilibino Nuclear Power Plant Storage. Here there are three holding pools each designed to hold 836 fuel assem-

blies. At the present time the capacity of each holding pool has been increased to 2030 fuel assemblies by increasing the pack-

ing density. The fuel assemblies are stored in 108 × 4 mm steel containers with minimum spacing 115 × 177 mm in a rect-

angular array. Hermetically sealed containers with fuel assemblies are filled with nitrogen. The storage site is filled with

water. After being completely filled, it is dried out. Since long-term storage sites are not available, even higher density pack-

ing of fuel assemblies is needed.

AMB-100, -200 Storage (Nos. 1 and 2 Units of the Beloyarsk Nuclear Power Plant). This fuel is not reprocessed,

though by design it should be. It is stored in 35- and 17-seat stainless-steel casings and also in 17-seat casings made of No.3

steel in two holding pools. Both pools contain 4994 fuel assemblies, of which 15% are not hermetically sealed. By now 35

No. 3 steel 17-seat casings with 935 fuel assemblies have exhausted their corrosion limits,water has penetrated into some of

the casings and interacts with the fuel matrix of unsealed fuel elements; some fuel assemblies are jammed in the casing pipes.

The facing of the No. 3 steel holding pools shows corrosion defects at the locations of the casings. The radioactivity of the

water in the pools is ~1 Ci/m3. From the standpoint of nuclear safety it is important that the wall thicknesses of the pipes in

the casings not decrease as a result of corrosion to less than 1.5 mm,since the steel used for the casing pipes is a neutron

absorber. For safety, it was decided to move the 17-seat casings into universal containers for dry storage on the site of the

nuclear power plant; the containers can then be shipped to storage sites.

Basic Problems of Nuclear Safety. Investigations and analysis of the nuclear safety of storage sites for spent fuel

performed over many years have identified the most important problems that need to be solved:

– selection and determination of the errors keff of computational methods;

– substantiation of nuclear safety for an anticipated accident – decrease of water density or optimal redistribution of

water as a result of an accident;

802

– taking account of fuel burnup in the substantiation of nuclear safety;

– use of neutron absorbers, taking account of consummable absorbers;

– the need and possibility of experimental determination of subcriticality of storage sites and degree of fuel burnup;

– substantiation of nuclear safety in the case fuel assemblies are dropped and damaged;

– analysis and classification of anticipated and unanticipated accidents; and,

– analysis of new types of fuel (uranium–gadolinium,uranium–erbium,uranium–plutonium),substantiation of safe-

ty for storage of new types of fuel assemblies.

Methods for Calculating keff for Spent-Fuel Storage Sites.At the present time various computer codes are being

widely used:Monte-Carlo codes – MCU-RFFI [2],MCNP [3], SCALE [4], MMKFK-2 [5] and also engineering computer

codes WIMS-D4,POLE,KSSETA, PERMAK,and others. A large number of critical experiments simulating the storage con-

ditions for VVÉR fuel assemblies for fuel elements have been performed to determine the computational errors keff. Special

critical experiments have not been performed for fuel assemblies from other types of reactors; in this case the errors can be

determined only by comparing the results from different tested computer codes using simplified physical models. In most

cases the calculation of keff for a storage site under normal conditions reduces to calculating keff for a particular cell; this

makes it possible to simplify the calculations and introduces a certain margin ≤0.3–0.5% keff. For accident conditions with a

decrease of water density this margin can be substantial,reaching 10–20%. Comparative calculations of keff show good agree-

ment among the codes listed above, 0.5–1% in keff, with the exception of water density up to 0.05 g/cm3, where the discrep-

ancy is larger; the reasons for this must be studied. A comparison of the results of calculations of keff with the data from spe-

cial critical experiments with low-enrichment fuel elements also showed that the methodological error is ~0.5% keff.

The use of certif ied computer codes for calculating keff for storage sites,as required by the State Atomic Inspection

Agency, is now becoming urgent. Even though this requirement is formally valid, it is not basic or fundamental from the

standpoint of safety. One of the main requirements is the principle of conservatism, implemented at all stages of analysis of

nuclear safety. Specifically, it is suggested that in a calculation the fuel should be fresher. This assumption itself gives a mar-

gin 20–30% keff. The rules incorporate the need for taking account of the tolerance in all quantities used in the calculations,

so that they would not increase keff. The actual value of keff does not exceed 0.7,as confirmed by measurement of the sub-

criticality of storage sites at nuclear power plants with RBMK-1000 reactors. The margin of subcriticality is adequate to com-

pensate for the methodological error in the calculations and as a rule is less than 1% keff.

Errors due to inadequacies in the accident scenarios, errors in the initial data, errors due to carelessness for lack of

training of the personnel performing the computational analysis, and other errors present a more serious problem. All this

shows that it is necessary to have a program to guarantee the quality of calculations and, most important,qualified analysts.

A program to guarantee the computational quality has been proposed in [6].

Change in Water Density. Analysis of accidents has revealed general problems. In the first place, there is the pos-

sibility that keff increases as a result of a decrease in water density in the entire volume of the storage site or in an individu-

al zone. This situation can occur when the power supply is cut off for a prolonged period of time, and it has been studied from

the standpoint of reaching criticality, the mechanism of the effects [7],and the actual range of variation of water density tak-

ing account of the finite dimensions and other structural features of specific storage sites. It has been established that keff can

exceed 1 when the water density decreases in storage sites where neutron absorbers are not used. The maximum value of keffis observed with water density ~0.05–0.3 g/cm3 and depends on the spacing of the fuel assemblies. If absorbers in the form

of boron steel casings are present in the storage site, keff decreases when the water density decreases. If 3–5 mm thick stain-

less steel casings are used, the increase in keff, as a rule, is negligible. For ~8 mm thickness stainless steel possesses the same

absorbing power as 5 mm thick boron steel with boron content 1.1%.

The Special Design Office Gidropress has performed special thermohydraulic experiments that determine the

coolant density as a function of the height of a fuel assembly in the cell of a storage site for spent fuel with fuel assembly

spacing 400 × 400 mm for nuclear power plants with VVÉR-1000 reactors. It was found that on boiling the coolant density

in the gap between fuel assemblies is 0.9–0.96 g/cm3, and inside fuel assemblies the density is at least 0.45 g/cm3 at the exit.

In addition, situations were studied with additional air introduced in the boiling regime; this simulated the breakdown of stan-

dard operation, when the cooling system is actuated with an incompletely drained section,which will cause air bubbles to

803

float upward and steam-air cavities to accumulate in individual regions of the holding pool for a limited period of time.

Thermohydraulic calculations show that keff does not exceed 0.95 in the accident situations indicated. Similar situations were

calculated for near-reactor storage sites for RBMK reactors. It was found that for closer container-free packing the water den-

sity on boiling does not exceed 0.8 g/cm3, provided that the fuel assemblies are allowed to stand freely for at least 8–15 days

before being placed in a more closely packed arrangement. For this density keff will not exceed 0.95.

Possibility of Increasing the Packing Density of Storage Sites.As a rule, the possibility of increasing and the

degree of increase in the packing density depend primarily on the spacing of the fuel assemblies, i.e., the nuclear-safety

requirements. There are several possibilities for increasing the packing density – decreasing the packing of the fuel assem-

blies by taking account of the geometry of the storage sites accurately, two-tier arrangement of fuel assemblies on racks,use

of special absorbers in the form of boron or stainless steel casings,and taking account of the degree of fuel burnup as a con-

trollable nuclear-safety parameter.

For closer packing of fuel assemblies in storage sites,anticipated and unanticipated accidents must be taken into

account. Measures must be developed for controlling unanticipated accidents. For certain types of storage sites,where the fuel

is confined in containers, it is possible to have a situation where on drying water remains inside the containers or boils out,as

a result of which keff can increase. Such situations are unanticipated. The simplest way to control them,from the standpoint of

nuclear safety, is to take steps to eliminate the possibility of reaching criticality without the use of special absorbers,especial-

ly since sometimes it is impossible to insert absorbers. For example, for RBMK storage sites the strength of the cantilever

beams on which additional absorbers are suspended and rules preventing the use of extractable absorbers limit the arrangement

of additional absorbers. Consequently, it has been suggested [7,8] that the degree of burnup of spent fuel be taken into account

in any analysis of unanticipated accidents,since in reality substantially burned up fuel is stored. If additional absorption of fis-

sion products and decrease of fuel enrichment in the process of burnup in a reactor are taken into account,then this problem

can be solved without introducing additional technical measures. The degree of burnup can be taken into account in long-term

storage sites,where fuel from near-reactor storage sites is stored and where off-loading of a damaged reactor core is not per-

mitted. On this basis a concept has been formulated for nuclear safety for RBMK storage sites [7]. Nuclear safety for normal

conditions and initial events of anticipated accidents must be proved assuming fresher fuel,and for unanticipated accidents the

degree of burnup must be taken into account. An important problem is to determine storage and shipment norms taking fuel

burnup into account. There are five basic factors that must be taken into account when solving this problem:

– the computational error for the isotopic composition of the fuel;

– the computational error for keff, determined by the complicated composition of fuel containing a mixture of ura-

nium,plutonium,and actinides;

– influence of fission products on keff;

– nonuniform distribution of fissioning materials over the height of a fuel assembly; and

– the irradiation history of the fuel assemblies.

The contribution of nuclides present in spent VVÉR-440 fuel with 30 MW·days/kg burnup and 1–5 yr holding peri-

od to keff was determined in [9]. The nuclides were divided into the following groups:main actinides – 235U, 236U, 238U,239Pu,240Pu,241Pu; secondary actinides – 238Pu,242Pu,241Am, 237Np; main fission products – 95Mo, 99Tc, 101Ru, 103Rh,109Ag, 133Cs,143Nd, 145Nd, 147Sm,149Sm,150Sm,151Sm,152Sm,153Eu,155Gd; and secondary fission products – all others.

The value of keff for a holding pool decreases by 0.17 as a result of a change in the nuclide composition of the fuel during

burnup of the main actinides,by 0.03 for secondary actinides,by 0.13 for the main fission products,and by 0.02 for the sec-

ondary fission products. The total change 0.35 is close for VVÉR-1000 holding pools with similar burnup and holding time

and is essentially independent of the holding time. Unfortunately, the estimates of the absorption cross sections of some basic

fission products differ by tens of percent. At the same time comparative tests show that the decrease in keff as a result of the

presence of 101Ru,103Rh,109Ag, 153Eu,155Gd in the fuel,calculated using different computer codes with different cross sec-

tions,is in good agreement. In practice, if the relative contribution of the nuclide groups enumerated above to keff is known,

the absorbing power of, for example, the main fission products,can be neglected, and in the process conservative values of

keff are obtained. The nuclide composition of spent fuel is also nonuniform because of the spatial dependence along the radius

and height of fuel elements and the fuel assemblies. Consequently, using the average composition in calculations can produce

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errors in keff. Investigations have shown that the radial nonuniformity over a fuel element changes keff in a 0.03% range; for

fuel assemblies the range is 0.8%. This nonuniformity can be neglected. The influence of the nonuniform burnup over the

height on keff is more complicated. For example, for RBMK-1000 fuel assemblies with average burnup 20 MW·days/kg the

burnup at the bottom and top ends of a 50 cm long interval are ~13 and 10 MW·days/kg, respectively, and the burnup at the

center is 26 MW·days/kg. Similarly, for VVÉR-440 fuel assemblies with average burnup 40 MW·days/kg the burnup at the

bottom and top end of a 25cm long interval are 30 and 25 MW·days/kg, respectively, and the burnup at the center is

45 MW·days. In practice, to take account of the nonuniformity of burnup over height a fuel assembly is divided into several

zones where burnup is assumed to be constant. In so doing, the zone length must be shorter at the bottom and top sections of

a fuel assembly, where burnup varies substantially, then at the center of a fuel assembly, where the burnup profile is flat. As

a rule, it is sufficient to partition the fuel assembly into 10–15 zones over height,since partitioning into smaller zones does

not affect keff. The influence of axial nonuniformity of the nuclide distribution on keff depends on the spacing of the fuel

assemblies in the storage site, the degree of burnup,and the type of fuel assembly. For example, for RBMK-1000 fuel assem-

blies the burnup nonuniformity increases keff by 2–4% compared with average burnup with water density ~1 g/cm3 and

decreases keff with water density ~0.2g/cm3 and less. In practical calculations,in a conservative approach burnup at the top

end of a 50 cm interval and not at the center of the fuel assembly can be used for VVÉR-440 fuel assemblies.

The influence of the irradiation history of a fuel assembly in a reactor in the course of a reactor run on keff of a stor-

age site is complicated because of the many different variants of fuel assembly irradiation in a reactor. Investigations have

established that the daughter nuclei, belonging to the main fission products and produced as a result of the decay of other fis-

sion products [10],for example, 147Sm,which forms as a result of the decay of 147Pm with 2.6 yr half-life, have the largest

effect on keff. At low reactor power the 155Eu concentration also becomes less,and a lower 155Eu concentration results in a

lower 155Gd concentration and a higher value of keff. However, by and large, the irradiation history changes keff by not more

than 0.5%,so that in practical calculations of storage sites it can be neglected. Analysis shows that nuclear safety requirements

for long-term storage sites for spent RBMK fuel are satisfied for normal operating conditions and for anticipated accidents,

provided that the fuel is fresher. At the same time, for unanticipated accidents,for example, boiling for 7–8 days and escape of

water from the containers in the storage sites with more closely packed fuel assemblies with spacing 110 × 115 mm in con-

tainers or drying of the holding pools with water remaining inside the containers,keff can exceed 1 if fresher fuel is present in

the holding pool. At the same time if the design burnup of fuel, ~20 MW·days/kg, is taken into account,subcriticality

(keff ≤ 0.95) obtains with a larger margin in such unanticipated accidents also,i.e., the storage site becomes self-protective as

a result of the internal properties of the system. The limit of safe burnup,for which keff ≤ 0.95,obtains if the fuel burnup is less

than 10 MW·days/kg for initial 235U enrichment 2.4%. Since defective fuel assemblies with burnup less than 10 MW·days/kg

can be present in a storage site together with fuel assemblies with the design burnup,local clusters of fuel assemblies with small

burnup must not be permitted and they must be distributed uniformly among fuel assemblies with burnup 15–20 MW·days/kg.

PNAÉ G-14-029-91 permits using burnup as the nuclear-safety parameter if this parameter can be monitored using special

devices in order to avoid errors by workers or incorrect calculations of burnup. However, at the present time a substantial num-

ber of fuel assemblies has now been accumulated in RBMK storage sites,so that measurement of burnup under the assump-

tion that the fuel assemblies are moved to the system and back to the storage location will increase unacceptably not only the

loading time in the holding pool but also the number of nuclear and radiation dangerous operations. In this situation it is best

to use methods and means making it possible to confirm experimentally the safety of the storage site as a whole. A pulse

method has been proposed for measuring keff of a storage site. Various forms of the pulse method for measuring keff exist.

However, experience has shown, especially for deep subcriticality conditions,that these methods exhibit a strong dependence

on spatial effects. Consequently, a more general relation between keff and the decay decrement α is used:

α = {[1 – keff(1 – βeff)]} / l,

where the decay decrement is an integral parameter characterizing the subcriticality of the system as a whole and is inde-

pendent of the location of the source and the detector; βeff is the effective fraction of delayed neutrons; and, l is the average

lif etime of the prompt neutrons.

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Comparing the computational results and the experimental data shows that the difference is 5–10% with

keff ~ 0.5–0.6,which indicates that the results of subcritical experiments can be used in practice, considering the large sub-

criticality of holding pools.

New Types of Fuel Assemblies,Mix ed Uranium–Plutonium Fuel.The advent of new fuel – uranium–erbium for

RBMK, uranium–gadolinium for VVÉR, and mixed uranium–plutonium for BN-600 and in the future for VVÉR-1000 reac-

tors – makes it necessary to substantiate the safety of existing storage sites for spent fuel. The special feature of

uranium–erbium fuel assemblies is that the enrichment is increased to 2.8–3% 235U with 0.4–0.8% consummable absorber

erbium is added. At the beginning of a reactor run the fuel assemblies (fresh) possess a lower value of keff than the standard

fuel without a consummable absorber and 2.4% enrichment. As burnup increases,keff of fuel assemblies with

uranium–erbium fuel decreases,in contrast to fuel assemblies with gadolinium as the consummable absorber. According to

PNAÉ G-14-029-91,consummable absorbers should be neglected when analyzing nuclear safety, assuming at the same time

that the fuel is fresh. This requirement is too conservative, which could make it necessary to change the standards for storing

RBMK fuel assemblies,which is economically disadvantageous. For VVÉR uranium–gadolinium fuel,taking account of the

large quantity of the consummable absorber (gadolinium) which is introduced, as burnup increases a large maximum of keffcan appear at 10 MW·days/kg. However, even in this case the PNAÉ G-14-029-91 requirement is too conservative, so that

paragraph 6.8.3 of PNAÉ G-14-029-91 needs to be changed. Nuclear safety should be analyzed for burnup for which keff is

maximum over the entire period of irradiation of a fuel assembly.

Comparative assessments of uranium fuel with 26% 235U enrichment and mixed uranium–plutonium fuel have

shown that under the most conservative assumptions – 30% plutonium content,2% 240Pu in the plutonium – keff for

mixed-fuel storage sites is lower. However, when assessments are made of accident situations associated with fuel leaking

into water it could happen that in a situation with a homogenized mixed fuel in water keff is somewhat larger than for urani-

um fuel in a similar situation.

Dropping of a Fuel Assembly and Casings with Fuel Assemblies. An accident which can occur during ship-

ments–technological operations is dropping of individual fuel assemblies or casings containing fuel assemblies with the

fuel assembly being destroyed and fissioning material spilled, fuel assemblies falling out of the casings and a compact

group of fuel assemblies being formed, damage to the facing of the holding pool,and uncompensated leakage of the hold-

ing pool. Cases where fresh and spent fuel assemblies have been dropped have been noted at home and abroad [11]. Tens

of fuel assemblies fell into an RBMK holding pool because of errors made by the workers or equipment failure. In all cases

the fuel assemblies were deformed, the deformations ranging from negligible to bending in half as a result of deformation

of the central bearing rod. No fuel was spilled in any of these cases. Experiments performed with fresh VVÉR-440 fuel

assemblies to determine damage occurring to them during shipment as a result of being dropped from a height of 9 m onto

a hard foundation showed presence of deformations that do not influence keff of an individual fuel assembly with absence

of fuel spillage.

To prevent damage to the facing of holding pools and to prevent leakage in the case where RBMK fuel assemblies

are dropped during shipment–technological operations it was decided to place additional ~10 mm thick steel sheets on the

bottom of the holding pools. The casings are not damaged when dropped, so that the only danger is from fuel assemblies

falling out of jackets and a compact group of assemblies being formed. One possible solution preventing a self-maintained

chain reaction in this case is placing on the casings devices that would prevent the fuel assemblies from falling out of the cas-

ings even when the casings are turned upside down. Another solution is to limit the number of VVÉR fuel assemblies in a

casing to a number for which keff is less than 0.95 even with compact packing in water with boron concentration ~16 g/kg.

The latter solution dereased the number of VVÉR-1000 fuel assemblies in an 18-seat casing when transporting fresh fuel

assemblies to a holding pond. Analysis of a similar situation with a 30-seat casing filled with VVÉR-440 fuel assemblies [12]

showed that the probability of a compact group of fuel assemblies being formed is 10–12–10–14 1/yr, so that such accidents

are not expected. We note that such a decrease in fuel assemblies in casings does not increase the safety, since in the process

the number of shipping operations and therefore the probability that a casing is dropped both increase.

Analysis of Unanticipated Accidents.According to PNAÉ G-14-029-91,it is necessary to examine the initial

events of unanticipated accidents,associated with the appearance of a self-maintained chain reaction for storage and

806

enrichment systems,complete loss of water from storage sites for spent fuel,and dropping of heavy technological equip-

ment at a storage site. The latter initial event,as a rule, reduces to the first two events. The appearance of a self-maintained

chain reaction is the result of initial events or a superposition of failures or worker errors. The most likely scenarios of a

self-maintained chain reaction are a decrease in water density as a result of prolonged shutdown of the electric power sup-

ply, loss of water with water remaining inside containers, decrease or convergence of a substantial number of fuel assem-

blies at the bottom of the holding pools as a result of dropping of the casings or heavy objects or some external action

(impact of an airplane, explosion,earthquake, and others). The BURST code was developed to solve this problem. The

code employs a point model of the kinetics and thermohydraulic processes in a single elementary cell of fuel elements or

fuel assemblies to solve simultaneously the equations governing the transient processes. The dependences of the reactivi-

ty on the density of the steam-water mixture and temperature are calculated beforehand and are the initial data for the

BURST code. As a self-maintained chain reaction develops,the regimes of heat removal from the fuel elements,each of

which influences keff of the system,change.

The method developed for calculating the thermophysical parameters makes it possible to calculate the space–time

distribution of the steam-water mixture during boiling and keff on the basis of this. The problem posed was implemented as

the TEHRA program. The computational results show that the quantity of fission products and actinides formed and the frac-

tion entering into the technological medium and the environment are determined by several factors, among which the deter-

mining ones are the total number of fissions over the self-maintained chain reaction,the fuel burnup,and the state of unsealed

and damaged fuel elements. A conservative model,which contains fuel assemblies with maximum burnup, i.e., maximum

accumulation of fission products,is used to calculate the worst consequences.

Analysis of a self-maintained chain reaction resulting from dropping a large number of fuel assemblies in differ-

ent compartments of near-reactor and intermediate storage sites at nuclear power plants with RBMK reactors shows the

following basic regularities: the energy release caused by the introduction of reactivity of several βeff per second is

1019–1021 fissions,the increase of the fuel-element temperature results in unsealing of the fuel elements and destruction

of the central part of the system in less than 1 sec after the self-maintaining chain reaction starts. The energy release deter-

mines the yield of radioactive inert gases and iodine. The yield of cesium and α-emitting nuclides is determined by their

production in the reactor, i.e., the degree of fuel burnup. Radioactive emissions into the atmosphere, assuming that they

do occur and bypass the ventilation system,amount to tens of Ci of 131I and about 2000–3000 Ci of cesium,which is less

than the regulatory requirements imposed on nuclear power plants for an anticipated accident involving a reactor with the

most serious consequences.

In a self-maintained chain reaction,caused by evaporation of water from containers with closer packing of

RBMK-1000 fuel assemblies with 110 × 115 mm spacing as a result of the cooling system being switched off for sever-

al days, the reactivity is due to a lowering of the water level in the containers, which is a slow process,occuring at a rate

~10 cm/h. A self-maintained chain reaction is also slow and shows small peaks in the fission intensity at 1 h which are

due to a decrease in water density and the Doppler effect. After 1.5 h the power is 4·105 W and the temperature of the

fuel-element cladding is ~430°C. The self-maintained chain reaction stops after the water in the storage site evaporates

to the level ~2 m. The time required to reach this water level is ~850 h. The total number of fissions over this period of

time is 3·1022. At such a temperature, evidently, fuel elements will remain sealed and no radionuclides will enter the envi-

ronment for 1000 h. An algorithm for calculating the temperature of the fuel,cladding, and concrete tiers was developed

for unanticipated accidents with the holding pools drying up. An example is a calculation of the drying of a compartment

of a holding pool at the storage site of the Chernobyl or Kursk nuclear power plant 26.4 m long, 5.6 m wide, 10.7 m

water level, 1580 m3 volume, containing 6038 fuel assemblies after a preliminary 1.5 yr holding period in near-reactor

pools. Considering the rate at which a single compartment is filled – 2000 fuel assemblies per year – the power of the

residual heat release in one compartment is 1000 kW. Variants where the heat is removed through the side walls and bot-

tom of the storage site and as a result of contact of the warm air, heated in the storage site, and cold air of the ventila-

tion system of the space above the water were considered. The maximum temperature becomes constant after 270 h fol-

lowing complete drying up and is 150 and 120°C for fuel element cladding and the interior surface of the side wall of

the storage site, respectively, and does not depend on the drying scenario (fast or slow). In the variant studied the fuel

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elements remain sealed; the strength of the wall of the storage site at the computed temperature of concrete is the safe-

ty determining factor.

In the present article we did not consider the nuclear safety of dry storage sites. Experience in designing such sites

predicts a substantially higher level of nuclear safety, which is achieved by constructing strong and sealed containers capa-

ble of withstanding various anticipated and unanticipated accidents,and the absence of neutron moderators in the containers.

However, evidently, the storage of fuel underwater will in the future be the required initial stage for decreasing the residual

heat release before switching to dry storage.

Conclusions.Experience in storing spent fuel under water and the results of investigations show that the system of

measures which has been developed guarantees that the storage sites are nuclear-safe. However, since closer packing of fuel

assemblies,new storage sites with new types of fuel assemblies,and stricter safety requirements are needed, additional inves-

tigations must be performed.

Computer codes for calculating keff, such as MMKFK-2,MCU-RFFI,and others,make it possible to calculate a stor-

age site for spent fuel with an error keff not greater than 1%. At the same time additional investigations are required to deter-

mine the computational errors for mixed uranium–plutonium,uranium–gadolinium,and uranium–erbium fuel.

Analysis of accidents shows that the determining factors for nuclear safety are taking account correctly of the change

in water density in the interior volume of the storage site or in individual zones of the site. The use of solid absorbers makes

it possible to increase substantially the capacity of a storage site and to decrease the danger of reaching criticality when water

density decreases. Taking account of the burnup makes it possible to have higher fuel packing density in long-time storage

sites or to substantiate safety during unanticipated accidents. However, additional investigations of the effect of complicated

isotopic composition of the fuel,a nonuniform distribution of fissioning nuclides over the height of a fuel assembly, and the

irradiation history of a fuel assembly in a reactor must be performed. Instrumental monitoring is required when fuel assem-

blies are packed in a storage site taking account of the degree of burnup.

Measurement of the subcriticality of storage sites makes it possible to determine the real value of keff of storage sites

and is an additional nuclear-safety measure. In addition, measurements make it possible to predict keff of a storage site dur-

ing unanticipated accidents. However, investigations are needed to determine the measurement error in keff with deep sub-

criticality. More detailed investigations of the consequences of unanticipated accidents are needed in order to develop mea-

sures to protect workers and the public.

REFERENCES

1. Safety Rules for Storage and Shipment of Nuclear Fuel to Nuclear Power Objects PNAÉ G-14-029-91,

TSNIIatominform, Moscow (1992).

2. E. A. Gomin and L. V. Maiorov, “The MCU-RFFI Monte-Carlo code for reactor design application,” in: Proceedings

of International Conference on Mathematics and Computation, Reactor Physics, and Environmental Analysis,

Portland, Oregon (1995).

3. MCNP – A General Monte Carlo Code for Neutron and Photon Transport, LA-7396-M (1986).

4. SCALE – A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation,

NUREG/CR-0200 (1982).

5. V. G. Polevoi, “Basic package of programs in the MMKFK-2 system for Monte Carlo solution of neutron transport

problems in reactor physics,” Preprint No. 00371, OFAP YAR (1996).

6. V. S. Vnukov, “Guarantee of computational quality of nuclear safety when working with fissioning materials,” Vopr.

Metrol., No. 4(12),61–70 (1991).

7. V. S. Vnukov and B. G. Ryazanov, “Use of degree of burnup for nuclear safety of storage sites for spent fuel,” in:

ICNC’91,Oxford (1991).

8. V. S. Vnukov, “Degree of burnup as a nuclear-safety parameter for storage sites and shipment packing containers

with spent nuclear fuel,” At. Tekh. Rubezh.,No. 12,9–11 (1990).

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9. L. Marcova, “VVER-related burnup credit calculations,” in: ICNC’99,Versailles,France (1999).

10. M. Dehart and B. Broadhead, “Investigation of burnup credit issues in BWL fuel,” ibid.

11. “Storage, handling, and movement of fuel and relation components at nuclear power plants,” Technical Report

No.189, IAEA, Vienna (1979).

12. G. A. Ponomarenko, “Analysis of criticality during shipment and storage of fuel at nuclear power plants with VVÉR

reactors,” At. Énerg., 87, No. 1,11–16 (1999).

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