NCE-2_lecture_470-534

06 Feb 2013

Decommissioning of Nuclear Facilities

11 May 2015 Dr. Muhammad Shafiq Siraj 471

1. Immediate Dismantling (or Early Site Release/'Decon' in

the US):

This option allows for the facility to be removed from regulatory control

relatively soon after shutdown or termination of regulated activities.

Final dismantling or decontamination activities can begin within a few

months or years, depending on the facility.

All components and structures that are radioactive are cleaned or

dismantled, packaged, and shipped to a low-level waste disposal site or

they are stored temporarily on site.

Once this taskwhich takes five or more yearsis completed, portion of

the site or the whole site can be reused for other purposes after

exemption from regulatory control.

Decommissioning Options


2. Safe Enclosure ('Safstor') or deferred dismantling:

This option postpones the final removal of controls for a longer period,

usually in the order of 40 to 60 years.

The facility is placed into a safe storage configuration until the eventual

dismantling and decontamination activities occur after residual

radioactivity has decayed.

For example, if a plant is allowed to sit idle for 30 years, the

radioactivity from cobalt 60 will be reduced to 1/50th of its original

level; after 50 years, the radioactivity will be about 1/1,000th of its

original level.

Once radioactivity has decayed to lower levels, the unit is taken apart,

similar to DECON.



3. Entombment (or 'Entomb'):

This option entails placing the facility into a condition that will allow the

remaining on-site radioactive material to remain on-site without ever

removing it totally.

This option usually involves reducing the size of the area where the

radioactive material is located and then encasing the facility in a long-

lived structure such as concrete, that will last for a period of time to

ensure the remaining radioactivity is no longer of concern.

The encased plant would be appropriately maintained, and surveillance

would continue until the radioactivity decays to a level that permits

termination of the plants license, with little or no additional

decontamination.


06 Feb 2013

Criticality


What is Criticality?

Fissile nuclide


A nuclear criticality accident is the occurrence of a self-sustaining neutron

chain reaction that is either unplanned or behaves unexpectedly. Only a

few special nuclear materials such as enriched uranium or plutonium are

capable of supporting a self-sustaining neutron chain reaction, hereinafter

called nuclear criticality.

Non-reactor nuclear facilities with operations, processes, storage,

handling and on-site transport of significant quantities of fissionable

materials are required to maintain a nuclear criticality safety (NCS)

program for the prevention of nuclear criticality accidents, in accordance

with ISO 1709:1995, Nuclear energy Fissile materials Principles of

criticality safety in storing, handling and processing.



When each fission leads to an average of more than one other fission,

the number of fissions and thus the ionizing radiations increase

exponentially: we then speak of a divergent chain reaction.

If such a phenomenon occurs accidentally in a nuclear facility (a plant or

a laboratory) or during the transport of fissile materials, it can expose

persons in the vicinity of the involved equipment to severe or even lethal

radiations.

Thus, we speak of a criticality accident, which moreover leads to the

production of fission products, including fission products in gaseous form.

These fission products may lead to a radioactive release into the

environment which is generally of limited extent.



The nuclear criticality risks must be considered at every stage of the fuel cycle

involving uranium, plutonium, and/or certain minor actinides (like for instance

curium, americium, etc.).

It includes uranium enrichment and conversion plants, plants for plutonium- and/or

uranium-based fuels manufacture, spent fuel reprocessing plants, research

laboratories involving fissile materials, effluent-treatment and waste-packaging

facilities and storage and transport of fissile materials (fuels, radioactive wastes,

etc.).

It is not necessary to have a complex process or large quantities of fissile materials

to initiate a divergent fission chain reaction. About 0.5 kg of plutonium 239 or 48

kg of uranium like the ones used to manufacture the fuel for PWR or BWR power

plants may be enough, in a spherical geometrical configuration with the presence

of water. By way of comparison, a 17 x 17 PWR fuel assembly contains more than

400 kg of uranium in a specially-designed geometrical configuration.



On the other hand, it is possible to handle relatively large quantities of fissile

materials as long as there is strict compliance with a set of parameters ensuring

that the criticality conditions will not be met.

The goal of nuclear criticality risks analysis is to define the necessary and sufficient

provisions (design and operational) to avoid the triggering of a divergent fission

chain reaction when fissile materials are present.

The nuclear criticality risks analysis consists of connecting (i) the possible

configurations of the fissile materials, in light of the actions that might be taken

during operations and the changes that might be caused by possible failures (error,

failures of a component, etc.) or by accidental situations (fire, earthquake, etc.),

and (ii) the margins between these configurations and potentially critical ones.

Nuclear Criticality Safety depends on the strict control of these actions.



The immediate result of a nuclear criticality accident is the production of

an uncontrolled and unpredictable radiation source that can be harmful,

even lethal, to people who are nearby.

In the workplace, nuclear criticality accidents last from a fraction of a

second up to several minutes, but may persist for much longer times,

depending upon the specific conditions.

A nuclear criticality accident itself provides various mechanisms that tend

to terminate the accident, and workplace personnel can also take actions

to terminate persistent accidents.

One accident that occurred in an experimental facility persisted for over

six days before it was terminated by facility personnel.



Neutrons emitted in the fission reaction (uranium 235, plutonium 239,

plutonium 241, etc.), after diffusion into the material, have three possible

fates:

to be absorbed by fissile nuclides and cause new fissions (can be qualified as

fissile capture);

to be absorbed by nuclides and "stay" in the nuclide, which then changes its

atomic number. In some cases, this reaction may lead to the production of a

fissile nuclide, as in the case of uranium 238, which - following several nuclear

reactions - is transformed into plutonium 239 (this is qualified as fertile

capture). In most cases, the reaction leads to the production of a non-fissile

nuclide: for example, boron 10 (20% of natural boron) which is transformed

into boron 11 (this is described as sterile capture);

to escape from the concerned system (neutron leakage), for example from the

tank containing the fissile solution.

Neutron Balance


Neutron Balance


This production of neutrons, if it is not offset by a sufficient loss (by fertile or sterile

captures and/or leakage) leads to an exponential increase in the number of

neutrons and to a criticality accident.

This balance is expressed by the neutrons effective multiplication factor (usually

denoted by keff), which indicates the factor by which the number of fissions is

multiplied from one generation of neutrons to the next one.

= =

+

where N is the number of "neutrons fathers" (generation n-1) having disappeared by

absorption or leakage and giving birth to N "neutrons sons"(generation n).

Neutron Balance


If keff < 1 (Production < Absorption + Leakage), the configuration is sub-

critical; this is the wanted safe state for nuclear facilities (excluding

reactors).

If keff = 1 (Production = Absorption + Leakage), the configuration is

critical; this is the equilibrium state encountered in a nuclear reactor

(controlled reaction), which must not be reached in other nuclear facilities.

If keff > 1 (Production > Absorption + Leakage), the configuration is

supercritical; this state corresponds to a criticality accident.

Neutron Balance


This neutron balance depends both on the characteristics of the fissile

medium (in particular the physico-chemical nature and its isotopic

composition which determine the fissile and fertile captures) and on the

geometry of the medium (which determines the proportion of neutrons

able to escape).

For example, for uranium, the limits depend on the content of isotope

235.

Thus, the minimum mass in a spherical shape that could lead to a

criticality accident (under conditions favorable to the reaction) is 0.87 kg

for highly-enriched uranium (93.5% 235U), 5.2 kg for an enrichment of

20%, and 48 kg for an enrichment of 4%.

Neutron Balance


the nuclear criticality risks are mastered by preventive provisions

implemented to control the configurations in which the fissile materials

are placed.

These provisions are expressed in practice by operational constraints

which, for example, consist of limiting the quantities of handled materials,

the dimensions of the equipment containing fissile materials, and/or the

concentrations of fissile materials in liquid media or by employing special

materials known as neutron absorbers (or poisons).

Depending on the particular nature of facilities, criticality detection and

alarm systems may be installed to enable the prompt evacuation of

personnel.

Criticality Preventive Actions


However, these systems are triggered only after the initiation of a chain

reaction and do not prevent the emission of the radiation associated with

the first moments of the accident (which may lead to lethal doses for

nearby operators).

On the other hand, the consequences for the environment of such an

accident are limited in range.

The releases of radioactive fission products comprise only a few rare gases

and very small amounts of iodine.

Furthermore, the radiations are attenuated by walls and other radiation

protection shields, and decrease when distance increases.

Criticality Preventive Actions


Implicit to the evaluated need for a nuclear criticality accident alarm

system is the requirement for the implementation of emergency

preparedness and response plans.

In consideration of such a need, ISO 11320:2011, Nuclear criticality

safety Emergency preparedness and response, was developed.

The new standard is designed to mitigate a nuclear criticality accidents

impact on human health and safety, quality of life, property and the

environment.

It was developed by ISO technical committee ISO/TC 85, Nuclear energy,

nuclear technologies, and radiological protection.

Various ISO standards exist and are developing to assist facility NCS

programs in the prevention of nuclear criticality accidents.

Rapid Response


The emergency preparedness and response plan is required to minimize

the consequences due to a nuclear criticality accident.

ISO 11320 therefore specifies the responsibilities of organizational

management, technical staff and individuals to that end.

It further requires that an evaluation of credible criticality accident

locations and characteristics be considered for establishing accident alarm

locations, immediate evacuation zones and emergency evacuation paths.

This will help personnel to avoid unnecessary radiation exposure when

exiting to predetermined emergency assembly stations.

If a nuclear criticality accident occurs at a nuclear facility, it is essential to

respond quickly, and even more important to have prepared an

emergency response.

Rapid Response


ISO 11320 provides criteria for establishing and implementing actions

that will effectively mitigate a potential accidents consequences for

human health and safety, quality of life, property and the environment.

Such emergency preparedness and response plans can also mitigate

unnecessary public angst about the hazard and its limited impacts on

operating personnel, facilities, the public and the environment in the rare

event of a nuclear criticality accident.

Rapid Response


Factors that affect criticality safety:

Fissile nuclide (233U, 235U and 239Pu)

Fraction of fertile nuclide diluting fissile nuclide (238U, 232Th or 240Pu)

Mass of fissile nuclide

Concentration of fissile nuclide

Geometry

Volume

Neutron moderators

Neutron reflectors

Neutron absorbers

Criticality Control in the PUREX Process


The preferred method of criticality control are engineered controls,

such as limiting geometry to be criticality safe under any credible

conditions

This often leads to conservative assumptions for credible conditions and adds to cost and complexity of the process

Limits equipment size and process throughput

Administrative controls have greater operational complexity

(procedures, standards, etc.), but offer greater design flexibility

and throughput

Typically, administrative controls require a double parameter failure for

a criticality to occur (no one-single control failure would cause a

criticality)



ANSI/ANS-8.1 (American National Standards Institute) states in

paragraph 4.2.2

Double Contingency Principle. Process designs should incorporate

sufficient factors of safety to require at least two unlikely, independent,

and concurrent changes in process conditions before a criticality

accident is possible.



Based on this double contingency principle, the normal case and each of the

contingencies considered one at a time must be determined to be

subcritical to establish the safety of the operation.

The method that is best used to establish this subcriticality depends on the

complexity of the parametric case.



The simplest, least complex way of establishing subcriticality is to use single

parameter limits parameters from ANSI/ANS-8.1 tables or figures. (See

example in the Tables next for aqueous solutions and metal units, respectively.)

The values in these tables represent the limiting critical values for individual

parameters, with the other parameters assumed to be at their worst possible

values for single units (including interaction with other fissile materials that is

bounded by complete water reflection.) Also covered by data in the ANSI/ANS-

8.1 standard are simple double-limit parameters, expressed as a curve.



Exempted quantity of fissionable materials

An exempted quantity of fissionable materials in the licensed site is defined as an inventory of fissionable materials, as follows:

1. less than 100 g of 233U, or 235U, or 239Pu, or of any combination of these three isotopes in fissionable material combined in any proportion; or

2. an unlimited quantity of natural or depleted uranium or natural thorium, if no other fissionable material nor significant quantities of graphite, heavy water, beryllium, or other moderators more effective than light water are allowed in the licensed site; or

3. less than 200 kg in total of natural or depleted uranium or natural thorium if some other fissionable materials are present in the licensed site, but the total amount of fissile nuclides in those fissionable materials is less than 100 g

Categorization of operations with fissionable materials


Small quantity of fissionable materials

A small quantity of fissionable materials in the licensed site is defined as an inventory of fissionable materials, which:

1. exceeds the exempt limits listed in the previous slide; but

2. does not exceed the following limits:

500 g of 233U, or 700 g of 235U, or 450 g of 239Pu, or 450 g of any combination of these three isotopes. These limits apply to operations with plutonium, 233U, or uranium enriched in 233U or 235U. These limits do not apply if significant quantities of graphite, heavy water, beryllium, or other moderators more effective than light water are present; or

80% of the appropriate smallest critical mass



Large quantity of fissionable materials

A large quantity of fissionable materials in the licensed site is defined as

an inventory of fissionable materials that exceeds the limits listed in the

previous slide.



Domain of nuclear criticality safety standards for non-reactor nuclear facilities



Parameter 233U 235U 239Pu

Mass of fissile material, g 550 760 510

Solution cylindrical diameter, cm 11.5 13.9 15.7

Solution slab thickness, cm 3.0 4.6 5.8

Solution volume, L 3.5 5.8 7.7

Concentration of fissile nuclide, g/L 10.8 11.5 7.0

Areal density of fissile nuclide, g/cm2 0.35 0.4 0.25

Uranium enrichment wt% 235U 1.0 %

Table C-1a: Single-parameter subcritical limits for uniform aqueous

solutions of fissile nuclides


Single-Parameter Limits for Uniform Aqueous Solutions of Fissile Nuclides

Table C-1b: Single-parameter subcritical limits for uniform

aqueous solutions of fissile nuclides


The areal densities of Table above are independent of the chemical

compound and are valid for mixtures that have density gradients, provided

the areal densities are uniform.

The subcritical mass limits for 233U, 235U, 239Pu in mixtures that might not

be uniform are 0.50, 0.70, and 0.45 kg, respectively, and are likewise

independent of compound.

Aqueous mixtures


The Table below contains 235U enrichment limits for uranium compounds

mixed homogeneously with water with no limitations on mass or

concentration.

Enrichment limits

Table C-2: 235U Enrichment Limits for Uranium Mixed Homogeneously with Water


The enrichment limit for uranium and the mass limits given in the next

table apply to a single piece having no concave surfaces. They may be

extended to an assembly of pieces, provided that there is no interspersed

moderation.

The 233U and 235U limits apply to mixtures of either isotope with 234U, 236U,

or 238U provided that 234U is considered to be 233U or 235U, respectively, in

computing mass.

The 239Pu limits apply to isotopic mixtures of plutonium, provided that the

concentration of 240Pu exceeds that of 241Pu and all isotopes are

considered to be 239Pu in computing mass. Density limits may be adjusted

for isotopic composition.

Metallic units


Metallic units

Table C-3: Single-Parameter Limits for Metal Units


Metallic units

Table C-4: Single-Parameter Limits for Oxides Containing no more than 1.5% Water by Weight at Full Density


Metallic units

Table C-5: Single-Parameter Limits for Oxides Containing no more than 1.5% Water by Weight at no more than Half Density(a)


Aqueous uranium solutions at low 235U enrichment

Table C-6: Limits for Uniform Aqueous Solutions of Low-Enriched Uranium


Uniform aqueous solutions of Pu(NO3)4 containing 240Pu

Table C-7: Limits for Uniform Aqueous Solutions of Pu(NO3)4 Containing 240Pu


Allowable volume of solution in a vessel packed with rings

Table C-8: Maximum Permissible Concentrations1 of Solutions2 of Fissile Materials in Vessels of Unlimited Size Packed with Borosilicate-Glass Raschig Rings


Subcritical limits for mixed-oxide heterogeneous systems

Table C-9: Subcritical Limits for Uniform Aqueous Mixtures of the Oxides of Pu and Nat. Uranium (Note: All values are upper limits except atomic ratios which are lower limits.)

C-11



Table C-10: Subcritical Limits for Single Units of Homogeneously Mixed Oxides of Plutonium and Natural Uranium at Low Moderation

(Note: The limits apply to combinations of plutonium isotopes provided 240Pu > 241Pu)



Table C-11: Subcritical Concentration Limits for Plutonium in Homogeneous Mixtures of Plutonium and Natural Uranium of Unlimited Massa

Note: These limits apply to combinations of plutonium isotopes provided 240Pu > 241Pu

06 Feb 2013

Criticality

Criticality Accidents in the World in Reprocessing Plants


Criticality Accidents in Reprocessing Plants


Chronology of process criticality accidents


Apart from reactors, with a few exceptions, all of the nuclear criticality

incidents have involved uranium or plutonium in the form of solutions.

Solutions can concentrate, leak, siphon, or be inadvertently transferred

from safe to non-safe geometry vessels or accumulate in non-safe

configurations.



Mayak (Russia) 1953

Procedural errors led to an unrecognized accumulation of 842 g of plutonium (as Pu nitrate solutions) in one vessel, which became critical and brought the vessel contents to boiling.

The operators transferred contents of another vessel to the first, ending the reaction.

Mayak (Russia) 1957

The accident occurred in a glovebox assembly within which uranium solution was precipitated into vessels.

An unexpectedly large amount of uranium precipitate accumulated in a filter receiving vessel.

The operator at the glovebox observed the filter vessel bulge prior to ejection of gas and some solution and precipitate from the vessel within the glovebox.



Mayak (Russia) 1958

Following the criticality accident at the same facility in 1957, an apparatus had been constructed to test criticality phenomena in fissile solutions.

A 400-liter tank on a platform was used for measurements involving solutions; after each experiment, the tank was drained into individual 6-liter containers of favorable geometry.

On this occasion, the tank contained uranyl nitrate solution (90% U-235) and was being drained for another experiment.

After filling several 6-liter containers, operators decided to circumvent the standard procedure to save time.

Three operators unbolted the tank and lifted it to pour directly into containers.

The presence of the operators provided sufficient neutron reflection to cause a criticality excursion, producing a flash of light and ejecting solution as high as the ceiling, 5 meters above the tank.



Oak Ridge (USA) 1958

A leak in a tank containing uranyl nitrate solution (93% U-235) was discovered on 15 June; the leak was not properly logged.

The following day other tanks were being drained into a 55-gallon drum; uranium solution from the leaking tank also entered the drum.

The operator nearest the drum noticed yellow-brown fumes rising from the drum's contents; he retreated before seeing a blue flash as the criticality excursion occurred.

Excursion power output rose for at least 3 minutes, then ended after 20 minutes.

Idaho (USA) 1959

Air sparging cylinders containing highly enriched uranyl nitrate solution initiated a siphon that transferred 200 L of solution to a 5000 gallon tank containing about 600 liters of water.

The resulting criticality lasted about 20 minutes.



Idaho (USA) 1961

Operators were attempting to clear a plugged line with air, which entered the evaporator, forcing the solution upward.

40 L of 200 g/L uranyl nitrate solution was forced up from a 5 in diameter section of an evaporator into a 24 in diameter disengagement cylinder, well above normal solution level.

Hanford (USA) 1962

Plutonium solution was spilled onto the floor of a solvent extraction hood.

Improper operation of valves allowed a mixture of plutonium solutions in a tank that became supercritical.

The excursion continued at low power levels for 37.5 hours, during which a remotely controlled robot was used to check conditions and operate valves.

Criticality was probably terminated by precipitation of plutonium in the tank to a non-critical state.



Mayak (Russia) 1968

Solutions of plutonium were being transferred from a large tank into a stainless steel vessel using a glass bottle.

While a worker was pouring a second load from the glass bottle into the vessel, a criticality excursion occurred.

Idaho (USA) 1978

A leaking valve allowed water to dilute the scrub solution used in the first cycle extraction process.

This leak was undetected because of a failed alarm system.

Because of the dilution, highly enriched uranium was stripped from the organic solvent (normally would remain in solvent).

Over the course of a month, the concentration of uranium increased in the large diameter bottom of the scrub column, resulting in a criticality.



Tokai-mura (Japan) 1999

Three operators were engaged in processes combining uranium oxide with nitric acid to produce a uranium-containing solution for shipment.

The uranium involved was 18.8% U-235.

The procedure which was used deviated from that licensed to the facility.

In particular the uranium solution was being placed in a precipitation tank for dispensing into shipment containers, not the more narrow vessel (geometrically favorable to minimizing criticality risks) prescribed by license.

While two workers were adding a seventh batch of uranium solution to the tank, a criticality excursion occurred.



Red Oil

Created by decomposition of TBP by nitric acid, under elevated temperature.

Influenced by presences of heavy metal (U or Pu), which causes higher organic solubility in aqueous solution and increases the density of the organic solution.

Decomposition of TBP is a function of nitric acid concentration and temperature.

Primary concern is in evaporators that concentrate heavy metals in the product

Red oil reactions can be very energetic, and have resulted in large explosions at reprocessing facilities

Typical safety measures include diluent washes or steam stripping of aqueous product streams to remove trace amounts of TBP before evaporation or denitration.

Major industrial accidents in Reprocessing Plants


How to avoid red oil in reprocessing facilities?

Temperature control

Maintain solutions at less than 130 C at all times

Pressure control

Adequate ventilation to avoid buildup of explosive gases

Mass control

Minimize or eliminate organics (TBP) from aqueous streams

Decanters, diluent washes, etc.

Concentration control

< 10 M HNO3

With solutions of uranyl nitrate, boiling temperature and density must be monitored

Multiple methods need to be employed so that no single parameter failure can lead to red oil formation

Controls to avoid Red Oil accidents


Other Major Accidents in Reprocessing Facilities

Mayak (Russia) 1957

Liquid high-level waste was stored in underground tanks.

The high level waste, coming from the B plant, contained sodium nitrate and acetate salts, from the acetate precipitation process.

Cooling system in one of the tanks failed, and the temperature in the tank rose to 350 C.

The tank contents evaporated to dryness, causing a massive explosion (estimated to be equivalent to 75 tons of TNT).

Over 20 MCi of radioactivity were released to the environment.



Tokai-mura (Japan) 1997

A fire occurred in the bitumen waste facility of the demonstration reprocessing plant at Tokai-mura (bitumen is used to solidify intermediate-level activity liquid radioactive waste).

The fire apparently occurred after errors were made in monitoring a chemical reaction.

The fire was not completely extinguished and about ten hours later, after chemicals had accumulated, an explosion occurred which ruptured the confinement of the facility.



Hanford (USA) 1997

Hydroxylamine* nitrate and nitric acid were stored in a tank and allowed to evaporate to dryness.

The resulting explosion destroyed the tank and blew a hole in the roof of the building.

*Hydroxylamine is a reagent used to reduce Pu valance from (IV) to (III).

THORP (Thermal Oxide Reprocessing Plant, Sellafield, UK) 2005

A pipe failure resulted in about 83,000 L of highly radioactive dissolver solution leaking into the stainless-steel lined feed clarification of the THORP facility.

This solution contained uranium and plutonium.

The leak went undetected for months before being discovered.

No injuries or exposure to radiation.

The plant is still shutdown in 2008.


Recent modifications to the PUREX Process

Industrial reprocessing firms have a high degree of confidence in the PUREX

process, however, the PUREX process has been the subject of criticism for

the past 30 years related to the separation of a pure plutonium stream.

Recall that the PUREX process co-extracts both uranium and plutonium,

then partitions them into separate streams.

Modifications to the PUREX process have recently been proposed and

developed that leave a small fraction of the uranium with the plutonium,

producing a mixed product for production of mixed oxide (MOX) fuel

These modified processes have been called COEXTM, NUEX or UREX+3 and

are all based on modified PUREX chemistry.

Calling these processes co-extraction to differentiate them from PUREX is

misleading because the PUREX process also co-extracts U and Pu.



Each specific process has its own proprietary methods of stripping

plutonium from the solvent, with a fraction of uranium.

In the PUREX process, the nitric acid concentration in the second scrub is

kept higher than ~0.5 M to keep the uranium in the organic solvent, while

the plutonium is reduced to the trivalent state and partitions to the aqueous

phase.

In the modified process, the acid concentration in the second scrub stream

is maintained at a controlled value (typically lower than 0.5 M) to allow a

small amount (~1%) of the uranium to partition to the aqueous stream

along with the plutonium (III).

After the Pu and small fraction of U are removed in the second scrub

stream, U is stripped from the solvent by using dilute (0.01 M) nitric acid.



Simplified flowsheet for U and U/Pu products

NCE-2_lecture_470-534

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