F.A.Qs About Decommissioning of Nuclear Power Reactors

adapted for educational purposes from:

European website on Decommissioning of Nuclear Installations

F.A.Qs concerning Decommissioning of Nuclear Power Reactors

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Table of Contents


Staff Responses to Frequently Asked Questions

Abstract   

Abbreviations

Introduction

Staff Responses to Frequently Asked Questions Concerning Decommissioning of Nuclear Power Reactors

1. General

2. Decommissioning Process  

3. Decommissioned Sites  

4. NRC (Nuclear Regulatory Commission) Activities  

5. Spent Fuel  

6. Radioactive Low-Level Waste  

7. Transportation  

8. License Termination And the Ultimate Disposition of the Facility  

9. Hazards Associated with Decommissioning  

10. Finances  

11. Socio-economic Issues  

12. Public Involvement  

13. Getting Additional Information  

14. Bibliography   


Staff Responses to Frequently Asked Questions


Abstract

By courtesy, we got the authorization to use these FAQ's in order to complete the information available in the present webpages. The reader has to take into account that all the information is not adapted to our countries / regulations; it is the reason why that only a part of the original document is hereafter reissued.

In order to get the whole document, links are built to the full original document and also to the official web source.

Once again we would like to be most grateful to the authors for their kindness.

In the shortcoming, the EC DB Net 2 group but also other partners will try to adapt most of the answers to the situtation met in the European Community.

The original report, through a question-and-answer format, provides "U.S. Nuclear Regulatory Commission (NRC)" staff responses to "Frequently Asked Questions" on the decommissioning process for commercial, nuclear power reactors.

The questions were taken from a variety of sources over the past several years, including written inquiries to the NRC and questions asked at public meetings and during informal discussions with the NRC staff.

In responding to the questions, the NRC staff attempted to provide the answers in a clear and non-technical form.

With the increase in the number of power reactors beginning the decommissioning process and significant changes that occurred in the regulations since 1996, the staff realized that there was a general lack of understanding of the decommissioning process and the risks associated with decommissioning. This document was developed in response to the staff's concerns. The report contains a definition of decommissioning and a discussion of alternatives. It also provides a focus on decommissioning experiences in the United States and how the NRC regulates the decommissioning process.

Questions related to spent fuel, low-level waste, and transportation related to decommissioning are answered.

Questions related to license termination, the ultimate disposition of the facility, and finances for completing decommissioning and hazards associated with decommissioning are also addressed.

This document also provides responses to questions related to public involvement in decommissioning as well as providing the public with sources for obtaining additional information on decommissioning.


Abbreviations

ADAMs Agency-Wide Documents Access and Management System
ALARA As low as is reasonably achievable
BWR Boiling Water Reactor
CEDE Committed Effective Dose Equivalent
CFR Code of Federal Regulations
DOE U.S. Department of Energy
DOT U.S. Department of Transportation
EA Environmental Assessment
EIS Environmental Impact Statement
EPA U.S. Environmental Protection Agency
EST Eastern Standard Time
FRERP Federal Radiological Emergency Response Plan
GEIS Generic Environmental Impact Statement
GE-VBWR General Electric - Vallicetos Boiling Water Reactor
GPO Government Printing Office
HEPA High-efficiency particulate air filter
HLW High level waste
ICRP International Commission on Radiation Protection
ISFSI Independent Spent Fuel Storage Installation
LLNL Lawrence Livermore National Laboratory
LLW Low-level waste
NCRP National Council on Radiation Protection and Measurements
NPDES National Pollutant Discharge Elimination System
NRC U.S. Nuclear Regulatory Commission
ODCM Offsite Dose Calculation Manual
PSDAR Post-shutdown Decommissioning Activities Report
PWR Pressurized Water Reactor
SFP Spent Fuel Pool
TEDE Total Effective Dose Equivalent
TDD Telecommunications device for the deaf
URL Universal Resource Locator
U.S. United States


Introduction

Sections 1 and 2 define decommissioning and discuss alternatives.

Section 3 focuses on decommissioning experiences within the United States.

Section 4 describes how the NRC regulates the decommissioning process.

Sections 5, 6 and 7 concern spent fuel, low-level waste during decommissioning, and transportation, respectively.

Sections 8 and 9, respectively, consider questions and answers on license termination, the ultimate disposition of the facility, and hazards associated with decommissioning.

Section 10 addresses the financial aspects of funding decommissioning.

Socio-economic issues are discussed in Section 11.

Section 12 discusses public involvement in the decommissioning process, with an emphasis on the early phases of decommissioning.

Section 13 provides the public with sources of additional information on decommissioning.

The final section contains a bibliography with relevant published materials.

As the rules / regulations are different in the European Community, Sections 4, 5, 6, 7, 8, 9, 10, 12 and 13 will only be treated on first hearing just with some adapted/relevant information (just only pointed out), or some of the definitions have been resumed.

They will be available when adapted by the different partners / regulators. http://www.eu-decom.be/faqs/faqs2.htm



Staff Responses to Frequently Asked Questions Concerning Decommissioning of Nuclear Power Reactors

CONTENTS OF THIS SECTION
1. General
2. Decommissioning Process
3. Decommissioned Sites
4. NRC Activities
5. Spent Fuel
6. Radioactive Low-Level Waste
7. Transportation
8. License Termination And the Ultimate Disposition of the Facility
9. Hazards Associated with Decommissioning
10. Finances
11. Socio-economic Issues
12. Public Involvement
13. Getting Additional Information
14. Bibliography

1. General

TABLE OF CONTENTS FOR THIS SECTION

1.1. How is decommissioning defined ?
1.2. Why do nuclear power plants shut down permanently ?
1.3. Why are power reactors decommissioned ?
1.4. How does decommissioning proceed ?
1.5. What are the benefits of decommissioning ?
1.6. What are the costs of decommissioning ?
1.7. What are the options to decommissioning ?



1.1. How is decommissioning defined ?
Decommissioning is defined as the safe removal of a facility from service and reduction of residual radioactivity to a level that permits termination of the license.


1.2. Why do nuclear power plants shut down permanently ?
Nuclear power plants cease operations for a variety of reasons. The Authorities (Regulators) grant a license for a period of 40 years. At the end of the license period, the licensee can seek to renew the operating license of the plant for another 20 years, or can cease operations and begin the decommissioning process. Some licensees choose to cease power operations before the 40-year licensing period has been completed. Reasons for this decision are usually financial. For example, the plant may require upgrades or repairs that are not economically justifiable, or the licensee may find other sources of power that are less expensive than nuclear generation. In addition to financial reasons for decommissioning, the Authorities can order the licensee to cease operations for safety reasons.

1.3. Why are power reactors decommissioned ?
As one of the conditions for an operating license, it is required to the licensee to commit to decommissioning the nuclear plant after it ceases power operations. This requirement is based on the need to reduce the amount of radioactive material at the site in order to ensure public health and safety as well as the protection of the environment.

1.4. How does decommissioning proceed ?
The regulations are written so that when a licensee announces its decision to permanently cease power operations at the nuclear power plant, the decommissioning process is automatically initiated. However, no major decommissioning activities can take place until the licensee has provided the Authorities with specific information regarding the decommissioning process as required by the decommissioning regulations discussed later. It is possible for the licensee to let the facility sit idle for a number of years before announcing its decision to permanently cease power operations (although the time could not extend beyond the duration of the operating license). However, it is not in the licensee's financial interest to delay this decision since the costs required to meet the regulations at an operating plant are much greater than the costs for a decommissioning plant.


1.5. What are the benefits of decommissioning ?
The major benefit of decommissioning for the licensee as well as the public is that the levels of radioactive material at the site are reduced to levels that permit termination of the license and use of the site for other activities, rather than leaving the radioactively contaminated material on the site so that it could adversely affect public health and safety and the environment in the future.


1.6. What are the costs of decommissioning ?
The major costs of decommissioning are the large financial costs involved in funding the project. Substantial costs are incurred in the removal, treatment, and disposal of major components of the facility that are contaminated, such as pumps, valves, piping, the steam generators and the reactor vessel. Decontamination of floors, walls and equipment also result in substantial costs. The occupational dose received by workers during decommissioning should also be considered as a cost.

1.7. What are the options to decommissioning ?

At the end of the licensing period (unless renewing and extension of the licensing), the regulations require that the facility be decommissioned. The alternative to decommissioning, at the end of the licensing period, is a "no action" alternative, implying that a licensee would simply abandon or leave a facility after ceasing operations. This is not considered to be a viable alternative to decommissioning.

The objective of decommissioning is to restore a nuclear facility to such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety or the environment. In order to ensure that at the end of its life, the risk from a facility is within acceptable bounds, some action is required. If nuclear power plants were not decommissioned, they could degrade and become radiological hazards.  



2. Decommissioning Process

TABLE OF CONTENTS FOR THIS SECTION

2.1. What terms or definitions are important to the understanding of decommissioning ?
2.2. What is the difference between radioactive contamination and activation products, and where are contaminated materials and activated materials located ?
2.3. How is a nuclear power plant decommissioned ?
2.4. Who decides how a facility should be decommissioned ?
2.6. What alternatives are currently used for decommissioning ?
2.7. What are the benefits and costs of the DECON alternative ?
2.8. What are the benefits and costs of the SAFSTOR alternative ?
2.9. What are the benefits and costs of the ENTOMB alternative ?
2.10. Is the choice of decommissioning alternatives a decision that is left entirely to the licensee, or does the NRC help make this decision ?
2.11. Must a licensee choose either DECON or SAFSTOR, or can it combine the two alternatives ?
1.12. What main factors affect a licensee's choice of a decommissioning alternative ?
2.13. How long does the dismantling phase last ?



2.1. What terms or definitions are important to the understanding of decommissioning ?

A number of terms (listed and defined below) are important to the understanding of decommissioning.

It is also important to gain an understanding of the units used for measuring radiation dose : rem and person-rem.

Activation products are radioactive materials that were created when stable substances were bombarded by neutrons.

For example, cobalt-60 is formed from the neutron bombardment of the stable isotope cobalt-59.

In a reactor facility, neutrons are created inside the reactor vessel during the fission process. These neutrons bombard (1) the metal around the reactor vessel, (2) the primary reactor coolant, and (3) the concrete near the reactor vessel and create activation products in these materials.

Alpha radiation is a positively charged particle ejected spontaneously from the nuclei of some radioactive elements.

It does not penetrate very far into material and it has a very short range even in air (a few centimeters). The most energetic alpha particle will generally fail to penetrate the dead layers of cells covering the skin and can be easily stopped by a sheet of paper. Alpha particles are hazardous when an alpha-emitting isotope is inside the body.

Background radiation means the radiation that is in the natural environment, including cosmic rays and radiation from the naturally radioactive elements, both outside and inside the bodies of humans and animals.

It also includes radon (from the ground) and global fallout (as it exists in the environment from the testing of nuclear explosive devices or from past nuclear accidents, such as Chernobyl, that contribute to background radiation and that are not under the control of the licensee).

Contamination means undesired (for example, radioactive or hazardous) material that is (1) deposited on the surface of, or internally ingrained into, structures or equipment, or (2) mixed with another material.

Dose or radiation dose is a generic term that means absorbed dose, dose equivalent, effective dose equivalent, committed dose equivalent, committed effective dose equivalent (CEDE), or total effective dose equivalent (TEDE).

In the case of radiation dose, it is energy absorbed per unit mass. Dose is measured in rads. The metric form of the rad is a gray (Gy) (1 rad = 0.01 gray). Radiation dose received by a person is measured in units called "rem," which incorporates the biological harm of the radiation dose based on the type of ionizing radiation.

A sievert (Sv) is the metric form of the rem (1 rem = 0.01 sieverts).

Greater than Class C waste is radioactive waste that is not generally acceptable for near-surface disposal. It is waste for which form and disposal methods must be different, and in general more stringent, than those specified for Class C waste. Such waste must be disposed of in a geological repository.

Half-life is the time required for half of any quantity of identical radioactive atoms to undergo radioactive decay, so that half of the atoms in the substance are no longer emitting radiation and are no longer considered to be radioactive.

Person-rem is the sum of all the radiation dose equivalents (measured in rem) that were received by an individual or by all individuals in a population group.

For example, if 1,000 people each received 1/10th of a rem (100 millirem), the corresponding population dose would be 100 person-rem.

Doses to an individual are usually measured in millirem. A sievert is the metric form of the rem (1 000 millirem = 1 rem = 0.01 sievert).

Radiation (ionizing radiation) means alpha particles, beta particles, gamma rays, x-rays, neutrons, high-speed electrons, high-speed protons, and other particles capable of producing ions.

Radiation, as used in this section, does not include non-ionizing radiation, such as radio or microwaves, or visible, infrared, or ultraviolet light.

Radioactive decay is the spontaneous natural process by which an unstable radioactive nucleus releases energy or particles.

Rem (see Dose).

Residual radioactivity means radioactive contamination or activation products that remain following decontamination and dismantling of the facility.


2.2. What is the difference between radioactive contamination and activation products, 
and where are contaminated materials and activated materials located ?

Radioactive contamination is radioactive material that is deposited on a nonradioactive surface.

The material may be deposited from the air, or it may be dissolved in water and subsequently deposited into material such as concrete. Radioactive contamination is generally located on or near the surface of materials like metal or high-density concrete or painted walls. It would travel farther into unpainted surfaces or lower density concrete. Radioactive contamination can usually be removed from surface areas by washing, scrubbing, spraying, or, in extreme cases, by removing the outer surface of the material.

Contaminated materials are transported through the facility by workers, equipment, and to some degree through the air. Although many precautions are taken to prevent the movement of contaminated material in a nuclear facility and to clean up any contaminated materials that may be found, it is most likely that contamination will occur in the reactor building, around the spent fuel pool, and around specific pieces of equipment in the auxiliary building.

The areas known to contain contamination are marked by the licensee, who routinely checks for contamination.

Activation products are radioactive materials that were created when stable substances were bombarded by neutrons. Typically these materials are the concrete and the steel that surround the fuel core. The radioactive decay of activation products is the main source of radiation exposure to plant personnel.


2.3. How is a nuclear power plant decommissioned ?

To decommission a nuclear power plant, the radioactive material on the site must be reduced to levels that would permit termination of the license.

This involves removing the spent fuel (the fuel that has been in the reactor vessel), dismantling any systems or components containing activation products (such as the reactor vessel and primary loop), and cleaning up or dismantling contaminated materials.

All activated materials generally have to be removed from the facility and shipped to a waste processing, storage or disposal facility. Contaminated materials may either be cleaned of contamination onsite, or the contaminated sections may be cut off and removed (leaving most of the component intact in the facility), or they may be removed and shipped to a waste processing, storage, or disposal facility.

The licensee decides how to decontaminate material, and the decision is usually based on the amount of contamination, the ease with which it can be removed, and the cost to remove the contamination versus the cost to ship the entire structure or component to a waste-disposal site.


2.4. Who decides how a facility should be decommissioned ?

The licensee decides how to decommission the site.

Frequently, licensees hire contractors that specialize in decommissioning sites to conduct part or most of the decommissioning. The process for decontamination and dismantling may vary from site to site.

Factors that are used to make these decisions include cost, worker exposure, availability of a waste site, and layout and structure of buildings. For example, at some sites, it may make more sense to segment the reactor vessel before removing it from the reactor building; in other cases, it would be appropriate to remove the reactor vessel intact through a hole cut in the side of the containment building and ship the reactor vessel intact.

If any major decommissioning activity does not meet the conditions specified by the regulations, the licensee is prohibited from undertaking the activity until; 1): it submits a license-amendment request that describes the proposed activity and the potential impact associated with that activity, and 2) the regulations approve the request.


2.6. What alternatives are currently used for decommissioning ?

The regulators have evaluated the environmental impacts of three general methods for decommissioning power facilities.

* DECON : The equipment, structures, and portions of the facility and site that contain radioactive contaminants are removed or decontaminated to a level that permits termination of the license shortly after cessation of operations.

* SAFSTOR : The facility is placed in a safe stable condition and maintained in that state until it is subsequently decontaminated and dismantled to levels that permit license termination.

During SAFSTOR, a facility is left intact, but the fuel has been removed from the reactor vessel, and radioactive liquids have been drained from systems and components and then processed.

Radioactive decay occurs during the SAFSTOR period, thus reducing the quantity of contaminated and radioactive material that must be disposed of during decontamination and dismantling.  

* ENTOMB :Radioactive structures, systems, and components are encased in a structurally long-lived substance, such as concrete.

The entombed structure is appropriately maintained, and continued surveillance is carried out until the radioactivity decays to a level that permits termination of the license.


2.7. What are the benefits and costs of the DECON alternative ?

The DECON option calls for prompt removal of radioactive material to permit restricted or unrestricted access.

The advantages of DECON include the following :
  • facility license is terminated quickly, and the facility and site become available for other purposes
  • availability of the operating reactor work force that is highly knowledgeable about the facility
  • elimination of the need for long-term security, maintenance, and surveillance of the facility, which would be required for the other decommissioning alternatives
  • greater certainty about the availability of low-level waste facilities that would be willing to accept the low-level radioactive waste
  • lower estimated costs compared to the alternative of SAFSTOR, largely as a result of future price escalation because most activities that occur during DECON would also occur during the SAFSTOR period, only at a later date.
  • (It is anticipated that the later the date for completion of the decommissioning, the greater the cost).

The disadvantages of DECON include the following :
  • higher worker and public doses (because there is less benefit from radioactive decay such as would occur in the SAFSTOR option)
  • a larger initial commitment of money
  • a larger potential commitment of disposal-site space than for the SAFSTOR option
  • the potential for complications if spent fuel must remain on the site until a repository for spent fuel becomes available.


2.8. What are the benefits and costs of the SAFSTOR alternative ?

The benefits of SAFSTOR include the following :
  • a substantial reduction in radioactivity as a result of the radioactive decay that results during the storage period
  • a reduction in worker dose (as compared to the DECON alternative)  
  • a reduction in public exposure because of fewer shipments of radioactive material to the low-level waste site (as compared to the DECON alternative)
  • a potential reduction in the amount of waste disposal space required (as compared to the DECON alternative)
  • lower cost during the years immediately following permanent cessation of operations
  • a storage period compatible with the need to store spent fuel onsite.

Disadvantages of SAFSTOR include the following :
shortage of personnel familiar with the facility at the time of deferred dismantling and decontamination
site unavailable for alternate uses during the extended storage period
uncertainties regarding the availability and costs of low-level radioactive waste sites in the future
continuing need for maintenance, security, and surveillance
higher total cost for the subsequent decontamination and dismantling period (assuming typical price escalation during the time the facility is stored).



2.9. What are the benefits and costs of the ENTOMB alternative ?

The benefits of the ENTOMB process are primarily related to the reduced amount of work in encasing the facility in a structurally long-lived substance, and thus, reducing the worker dose from decontaminating and dismantling the facility. In addition, public exposure from waste transported to the low-level waste site would be minimized. The ENTOMB option may have a relatively low cost.

However, because most power reactors will have radionuclides in concentrations exceeding the limits for unrestricted use even after 100 years, this option may not be feasible under the current regulations.

This option might be acceptable for reactor facilities that can demonstrate that radionuclide levels will decay to levels that will allow restricted use of the site.

Three small demonstration reactors have been entombed. Currently, no licensees have proposed the ENTOMB option for any of the power reactors undergoing decommissioning.


2.10. Is the choice of decommissioning alternatives a decision that is left entirely to the licensee, or does the Regulator help make this decision ?
The choice of the decommissioning method is left entirely to the licensee. However, the regulator would require the licensee to re-evaluate its decision if the choice (1) could not be completed as described, (2) could not be completed within a defined period after the permanent cessation of plant operations, (3) included activities that would endanger the health and safety of the public by being outside of the health and safety regulations, or (4) would result in a significant impact to the environment.


2.11. Must a licensee choose either DECON or SAFSTOR, or can it combine the two alternatives ?

A licensee need not restrict its choice of decommissioning options to either an immediate decontamination and dismantling or to a storage period of 30 to 60 years, followed by decontamination and dismantling.

Generally licensees combine the first two options. For example, after power operations stop at a facility, a licensee could use a short storage period for planning purposes, followed by removal of large components (such as the steam generators, pressurizer, and reactor vessel internals), place the facility in storage for 30 years, and eventually finish the decontamination and dismantling process.


2.12. What main factors affect a licensee's choice of a decommissioning alternative ?

The SAFSTOR alternative is often used at multi-unit sites when one or more of the units shuts down while others continue to operate.

This is especially true for facilities that share some systems. In this case, the staff from the operating unit(s) assist in the maintenance and surveillance of the unit that is in storage.

The choice of decommissioning options is also strongly influenced by potential uncertainties in low-level waste disposal costs and by concerns over the future availability of low-level waste sites. The licensee's rate regulator can also influence the choice of decommissioning alternatives.


2.13. How long does the dismantling phase last ?

The dismantling phase typically takes between 3 to 5 years to complete, although it may take longer if there are constraints on access to low-level waste burial sites, or if the licensee decides to proceed at a slower pace for programmatic reasons.




3. Decommissioned Sites

TABLE OF CONTENTS FOR THIS SECTION

3.3. What improvements have been made as a result of previous decommissioning experience ?
3.4. What research is being performed to find improved methods to be used during decommissioning ?
3.5. What differences are there in decommissioning between different types of reactor designs ?



3.3. What improvements have been made as a result of previous decommissioning experience ?

Some improvements in the process, such as the removal of large components, including the reactor vessel, and the use of a primary system chemical flush to reduce worker exposure, have resulted from the experience gained from previous plant decommissionings.

These include strippable coatings of latex or plastic that are used for decontaminating surfaces and the increased use of robotics.

In addition, most licensees have gained experience with decommissioning techniques during routine preventive maintenance programs or as part of repairs required during operations.


3.4. What research is being performed to find improved methods to be used during decommissioning ?

The following types of improvements are being investigated :

* surface-removal techniques to remove the outer surface of a contaminated structure, such as lasers or microwaves combined with vacuums, electrohydraulic scabbling (water-pressure shock waves that are electrically controlled), and electrokinetic decontamination of concrete (gel electrolytes are used with electrodes to leach ionic contaminants from deep inside porous concrete)  

* cutting techniques, such as laser cutting or oxy-gasoline torches (which work twice as fast as an acetylene torch on 1-inch steel) to remove structures  

* improved methods for worker protection, such as protective suits with liquid air-cooling apparatus and lightweight breathable suits with chemical absorption protective layers  
environmental-protection techniques, such as automated asbestos removal and in situ chemical conversion of asbestos to non-hazardous material

* survey/monitoring techniques, such as pipe-explorer internal survey/characterization systems and a remote 3-D characterization and archiving system (robotic sensor and mapping platforms analyze for hazardous organic and radioactive contaminants).  

* Commercial firms are also developing promising avenues of research into usable technologies.



3.5. What differences are there in decommissioning between different types of reactor designs ?

An analysis (of decommissioning at nuclear facilities, including nuclear power plants) looked at total estimated costs, occupational and public dose, and low-level waste volumes.

The GEIS estimates for cost for the reference facility were generally higher by about 20 % for the boiling water reactors (BWRs) than for pressurized water reactors (PWRs), depending on the decommissioning option selected.

Occupational dose estimates in the GEIS were slightly higher for the reference BWR, by 10 to 50 %, depending on the decommissioning option.

Estimates of public dose were lower for the reference BWR by up to a factor of 2, depending upon the scenario.

Burial-volume estimates for low-level waste for reference BWRs and PWRs for SAFSTOR and DECON options were very close to the same. The staff expects to issue a draft supplement in the year 2001.

One other major difference between decommissioning BWRs versus PWRs is that BWRs are designed so that the spent fuel pool is located in the reactor building, rather than in a separate building that can be isolated from the rest of the facility.

This eliminates the possibility of decontaminating and decommissioning the remainder of the facility while leaving the spent fuel pool building as a "nuclear island".




4. NRC Activities

See original FAQs from the NRC.  




5. Spent Fuel

Here also only some "general questions / answers" will be broached; indeed, in the European Community, rules differ from one land to another. It will be, when possible, paid attention to this topic and answers will be adapted.

TABLE OF CONTENTS FOR THIS SECTION

5.1. What are "high-level wastes" ?
5.2. What is meant by the term "spent fuel" ?
5.8. Spent Fuel Pools
5.8.1. Why is spent fuel stored in a pool of water ?
5.8.2. Has the spent fuel pool been analyzed to determine the limits for heat removal due to spent fuel storage ?
5.8.3. Do spent fuel pools leak, and if they do, how much radioactive material could be leaked, and where would it go ?
5.8.4. What would happen if there were a loss of heat-removal capability of water in a spent fuel pool when it was fully loaded ?
5.8.5. What would happen to the fuel in the spent fuel pool if an earthquake ruptured the pool, or if an airplane crashed into the pool ?
5.8.6. What can be done to prevent the spent fuel pool from boiling dry ?



5.1. What are "high-level wastes" ?

High-level radioactive waste (HLW), as it pertains to commercial nuclear power reactors, is mainly irradiated (spent) reactor fuel.


5.2. What is meant by the term "spent fuel" ?

Spent nuclear fuel is uranium-bearing fuel elements that have been used at commercial nuclear power reactors.

Although spent (used) fuel can no longer produce enough heat to produce electricity, it contains highly radioactive material resulting from the fission process that takes place within the reactor.

As a result, it still continues to generate radiation and heat. This heat and radiation are caused by "radioactive decay" of the products of the fission process.

The heat and radioactivity in spent fuel necessitate that any shipment be made in containers or casks that provide the necessary degree of protection.

In practice, this means that a cask must shield and contain the radioactivity and dissipate the generated heat.


5.8. Spent Fuel Pools


5.8.1. Why is spent fuel stored in a pool of water ?

Even after the nuclear reactor is shut down, the fuel continues to generate decay heat.

Decay heat results from the radioactive decay of fission products. The rate at which the decay heat is generated decreases the longer the reactor has been shut down.

So the longer the spent fuel has been out of the reactor, the less heat that it gives off.

Storing the spent fuel in a pool of water is a way to provide an adequate heat sink for the removal of heat from the irradiated fuel. In addition, the fuel is located far enough under water that the radiation emanating from the fuel is shielded by the water to adequately protect the workers from the radiation.


5.8.2. Has the spent fuel pool been analyzed to determine the limits for heat removal due to spent fuel storage ?

Yes. The regulations give criteria that must be met for fuel storage and handling. This includes designing fuel-storage systems to ensure adequate safety under normal and postulated accident conditions. The system is to be designed with suitable shielding for radiation protection, with appropriate containment, confinement, and filtering systems, and with a heat-removal capability that is reliable and that can be tested to ensure that it meets the requirements for removing the heat produced by the spent fuel.


5.8.3. Do spent fuel pools leak, and if they do, how much radioactive material could be leaked, and where would it go ?

All nuclear power plants have a reinforced-concrete spent fuel pool (SFP) structure designed to retain its function, even following the design-basis seismic event (that is, seismic Category 1 or Class 1 [earthquake]) that is anticipated for the area. The SFP also has a welded, corrosion-resistant liner.

All plants except for one have leak-detection channels positioned behind liner plate welds to collect any leakage and to direct the leakage to a point at which it can easily be monitored. Nearly all nuclear power reactors have passive features preventing draining or siphoning of the SFP to a coolant level below the top of stored, irradiated fuel. Excluding paths used for irradiated fuel transfer, passive features at nearly all nuclear reactors prevent draining or siphoning the coolant to a level that provides inadequate shielding for fuel seated in the storage racks.

In the event that SFP coolant inventory decreases significantly, several indicators are available to alert operators to that condition. The primary indication is a low-level alarm. A secondary indication of a loss of coolant is provided by area radiation alarms.

These primary and secondary alarms indicate a loss of shielding that occurs when SFP coolant inventory is lost. Except for the SFP located inside the containment building, the area radiation alarms are set to alarm at a level low enough to detect a loss of coolant inventory early enough to allow for recovery before radiation levels could make such a recovery difficult.

The level of radioactivity in the water of the spent fuel pool is low. In addition, nuclear plants have cleaning systems to maintain the purity of the water in the spent fuel pools. All nuclear plants have a groundwater monitoring system around the facility so that if a system leaks, there is a method for alerting the licensee to the problem as well as for providing information regarding the location of the contamination.


5.8.4. What would happen if there were a loss of heat-removal capability of water in a spent fuel pool when it was fully loaded ?

The consequences of losing the heat-removal capability or water (coolant) in a spent fuel pool depends on the amount of time since the fuel was last used for power operation inside the reactor.

If fuel was recently used for power operation, there may be enough decay heat to cause the spent fuel pool coolant to heat up to the boiling point if forced cooling were lost to the spent fuel pool.

If plant operators took no action, boiling would cause the level in the spent fuel pool to decrease over time. However, operators have redundant sources of water to add to the pool to maintain coolant level should a loss of forced cooling occur.

Operators are alerted to a loss of level condition by a series of alarms at the cooling system control station and in the main control room. Given the unlikely event that no operator action is taken, the pool level would decrease at a very slow rate (about one foot every several hours to weeks, depending on the age of the stored fuel).

The longer the time interval since the last batch of fuel was used to generate power in the reactor, the longer it would take for the spent fuel to boil off the spent fuel pool coolant. Boiling the spent fuel pool coolant is, however, an acceptable method for cooling spent fuel and has a minimal effect on public health and safety. In the unlikely event that a large loss of coolant uncovers the spent fuel, if sufficient time has not elapsed since the fuel was used to generate power in the reactor, the spent fuel may have enough decay heat to overheat the cladding in air and cause it to ignite.

The resulting fire could carry radioactive particles offsite and the consequences could be significant.

However, the NRC staff considers this a very low probability accident because of design features required at all spent fuel storage pools that minimize the possibility of losing all of the spent fuel pool coolant.


5.8.5. What would happen to the fuel in the spent fuel pool if an earthquake ruptured the pool, or if an airplane crashed into the pool ?

Spent fuel pools are designed to withstand earthquakes greater than any earthquake that actually occurred or is expected to occur in the area of the plant. Therefore, the probability of the spent fuel pool rupturing due to an earthquake is very low.

However, in the unlikely event that a very large earthquake does occur, one that is larger than the pool was designed to withstand, the pool structure could fail and allow the coolant to drain out. The consequences of an accident like this are discussed in the response to Question 5.8.4, above.

In the unlikely event that an aircraft crashed into the spent fuel pool, the pool structure could be severely damaged and not capable of maintaining coolant level. In this event, consequences such as those discussed in Question 5.8.4 could result. However, the staff has evaluated the possibility of an aircraft impacting the spent fuel pool and consider it a very low probability event.  


5.8.6. What can be done to prevent the spent fuel pool from boiling dry ?

A cooling system removes decay heat from the spent fuel pool. The coolant in the spent fuel pool is maintained below a specific temperature and the level of the water is maintained at a specific height over the spent fuel. Temperature indicators are installed and are either equipped with an alarm or require visual surveillance on a daily basis. High/low-water-level monitors are installed in spent fuel pools. The monitor alarms at the spent fuel pool and in the control room when the spent fuel pool's water level is not within the specified limit.



6. Radioactive Low-Level Waste

TABLE OF CONTENTS FOR THIS SECTION

6.1. What is meant by low-level radioactive waste, and how is it different from fuel ?
6.2. How is the low-level radioactive waste disposed of ?
6.8. Can the reactor vessel be disposed of at a low-level waste (LLW) site ? Can the reactor vessel be shipped intact, or does it have to be segmented ?
6.9. How are liquid wastes disposed of ?


6.1. What is meant by low-level radioactive waste, and how is it different from fuel ?

Low-level waste (LLW) is any radioactive waste that is not classified as high-level waste, spent nuclear fuel, transuranic waste (containing manmade elements heavier than uranium that emit alpha radiation -- transuranic waste is produced during reactor fuel assembly, weapons fabrication, and chemical processing operations), or uranium or thorium mill tailings.

LLW often contains small amounts of radioactivity dispersed in large amounts of material, but may also have activity levels requiring shielding and remote handling. It is generated by reactor fuel production, reactor operations, isotope production, medical procedures, and research and development activities.

LLW usually comprises the following material contaminated with radionuclides: rags, papers, filters, solidified liquids, ion-exchange resins, tools, equipment, discarded protective clothing, dirt, construction rubble, concrete, or piping.

LLW usually, but not necessarily, includes waste with relatively low concentrations of radionuclides. Although the classification of waste can be complex, "Class A waste" generally contains lower concentrations of longer half-life radioactive material than "Class B and C wastes".

Greater than Class C waste are not considered low-level radioactive waste and must be handled and disposed of differently from Class A, B and C wastes.


6.2. How is the low-level radioactive waste disposed of ?
Low-level waste is commonly disposed of by burial in near-surface shallow trenches.  
After they are filled with containers, the trenches are usually covered with a low-permeability cover (such as clay). They are then often covered with a gravel drainage layer and a layer of topsoil.

Vegetation is planted on top for erosion control. There is no intent to recover the wastes once they are disposed of.

The volume of waste that is being disposed of each year is decreasing as the result of industry efforts to compact or incinerate part of the waste.


6.8. Can the reactor vessel be disposed of at a low-level waste (LLW) site? Can the reactor vessel be shipped intact, or does it have to be segmented?

A reactor vessel can not be disposed of at a LLW site unless it meets waste classification requirements specified in the regulations and any site-specific requirements specified in the disposal facility's license.

In most cases where disposal of the reactor vessel has occurred, the reactor vessel internals have been removed before any parts of the reactor vessel were shipped to an LLW disposal site.

Licensees are required to demonstrate that the shipment meets the regulations for package integrity and that the package meets the acceptance criteria for the LLW disposal site, such as criteria for radionuclide concentration and waste form. 6.9. How are liquid wastes disposed of?

Liquid wastes are processed onsite. The liquid portion is separated from the solid portion. The solid portion is disposed of in the low-level waste site, as long as it meets the criteria for low-level waste.  

The concentration of radioactive material in the liquid portion is measured, and if the concentration is below the limits in use for release of effluents, the liquid portion may be released offsite (for instance, to sewers or a nearby body of water). Otherwise, the liquid portion is solidified (by mixing with concrete or similar solidifying or absorbing material) and disposed of as solid low-level waste.



7. Transportation

The information is at this moment not "adaptable" to our cases.



8. License Termination And the Ultimate Disposition of the Facility

Only some questions / answers of the original section are of application.


TABLE OF CONTENTS FOR THIS SECTION

8.1. How does decommissioning end, and who decides that the decommissioning is complete ?
8.2. Why is the license-termination plan filed so late in the process ?
8.3. How will the licensee know where the radioactive material or contamination is located within the plant ?
8.4. What is included in the site characterization ?
8.5. What does "suitable for release" mean ? Are there any restrictions on how the site can be used ?
8.6. Why would the licensee be allowed to restrict use of the site ?
8.7. What is "residual radioactivity", and why is it important to the termination of the license ?
8.10. What is a "total effective dose equivalent" ?
8.14. How does the dose based on the residual radioactivity levels relate to background dose levels ?
8.16. Is it possible that some isotopes are located in such a way that radiation-monitoring devices cannot accurately detect their levels of radioactivity ?
8.17. Will continued monitoring be required after the decommissioning process is complete to ensure that the radiation levels do not increase ?
8.18. What types of uses can be made of the plant site after decommissioning is completed ?
8.19. Could the licensee initiate an alternative use of the site or partial site release before the decommissioning process is completed ?



8.1. How does decommissioning end, and who decides that the decommissioning is complete ?

Licensees must submit an application for license termination at least 2 years (depends on the country, see section "Regulations") before the requested termination date of the license.

The license-termination plan must include :
* a site characterization  
* identification of remaining dismantling activities
* plans for site remediation
* detailed plans for the final survey of residual contamination on the site
* a method for demonstrating compliance with the radiological criteria for license termination.

For restricted release, the license-termination plan should include a description of the site's end use and documentation on public consultation, institutional controls, and financial assurance needed to comply with the requirements for license termination for restricted release or alternative criteria.
* an updated site-specific estimate of remaining decommissioning costs
* a supplement to the environmental report that describes any new information or significant environmental changes associated with the licensee's proposed termination activities.


8.2. Why is the license-termination plan filed so late in the process ?

The initial decommissioning activities (decontamination and dismantling) are not significantly different from routine operational activities such as replacement or refurbishment.

Because of the framework of regulatory provisions that are in place in the licensing basis for each facility, these activities do not present significant safety issues for which a detailed plan such as the license-termination plan is warranted.

Therefore, it is appropriate that the licensee be permitted to conduct these activities without the need for a license amendment. At the license-termination stage (towards the end of the decommissioning process), the Commission must consider :

(1) the licensee's plan for assuring that adequate funds will be available for final site release,

(2) the radiation-release criteria for license termination, and

(3) the adequacy of the plans for the final survey that is required to verify that the release criteria have been met.


8.3. How will the licensee know where the radioactive material or contamination is located within the plant ?

During operation, the plant is required to keep records of radiological surveillances that document where contaminations occur and locations of activation products and other sources of radiological materials.

This is the basis for a plant's initial assessment of the location of radioactive material and contamination.

Radiological surveys continue during all phases of decommissioning, and the records of these surveys are also kept. This information is used as part of the basis for the site-characterization plan; however, additional measurements are made at the site-characterization stage.

The characterization plan must be designed to demonstrate compliance with appropriate dose or risk-based regulations.


8.4. What is included in the site characterization ?

The purpose of the site characterization is to ensure that the final radiation surveys are conducted to cover all areas where contamination existed, remains, or has the potential to exist or remain as well as to provide data for planning further decommissioning activities.

The site characterization contains a description of

(1) the radiological contamination on the site before any cleanup activities associated with decommissioning took place,

(2) a historical description of site operations, spills, and accidents,

(3) a map of remaining contamination levels and contamination locations, and

(4) a description of the survey instruments and supporting quality assurance practices used in the site-characterization programme.



8.5. What does "suitable for release" mean? Are there any restrictions on how the site can be used ?

There are two broad categories of uses for the facility after the license termination.

The first is "unrestricted use", and the second is "restricted use". These will be discussed separately.

Unrestricted use means that there are no restrictions on how the site may be used. The licensee is free to continue to dismantle any remaining buildings or structures and to use the land or sell the land for any type of application.

Restricted use means that the licensee has demonstrated that further reductions in residual radioactivity would result in net public or environmental harm, or residual levels are as low as is reasonably achievable.

The licensee must have made provisions for legally enforceable institutional controls (for example, restrictions placed in the deed for the property describing what the land can and cannot be used for), which provide reasonable assurance that the radiological criteria set by the regulator will not be exceeded.

In addition, the licensee must have provided sufficient financial assurance to an amenable independent "third party" to assume and carry out responsibilities for any necessary control and maintenance of the site.

There are also regulations relating to the documentation of how the advice of individuals and institutions in the community who may be affected by the decommissioning has been sought and incorporated in the license-termination plan related to decommissioning by restricted use.

Although power reactor licensees can choose either a restricted or unrestricted option for release, the restricted option is primarily for materials licensees and would not normally be selected by reactor licensees because of the low levels of site contamination.


8.6. Why would the licensee be allowed to restrict use of the site ?

There can be situations in which restricting site use can provide protection of public health and safety by reducing the total effective dose equivalent in a more reasonable and cost-effective manner than unrestricted site use.

This protection is afforded by limiting access to the site, limiting the amount of time that an individual spends onsite, or by restricting agricultural or drinking water use.

For many facilities, the time period requiring this type of restriction can be fairly short, and need only be long enough to allow radioactive decay to reduce radioactivity to levels that permit the site to be released for unrestricted use.


8.7. What is residual radioactivity, and why is it important to the termination of the license ?

The term "residual radioactivity" means radioactivity in structures, materials, soils, groundwater, and other media at a site resulting from activities under the licensee's control.

This includes radioactivity from all licensed and unlicensed sources used by the licensee, but excludes background (natural) radiation. It also includes radioactive materials remaining at the site as a result of routine or accidental releases of radioactive material at the site and previous disposals at the site.

Criteria for the termination of the license are based on the residual radioactivity levels remaining at the site at the end of decommissioning.


8.10. What is a "total effective dose equivalent" ?

The "total effective dose equivalent" is a term that is used to express how the radiation dose is calculated to an individual.

It means that the dose from radioactive material outside of the individual (external radiation) and the dose from any radioactive material that the individual may have inhaled or ingested (internal radiation) have been considered.

For the latter case, the internal radiation dose is considered for a period of 50 years following the intake of the radioactive material.

In addition, weighting factors are used that are specific to the body organs or tissues that are irradiated. These weighting factors are used to account for the variation in sensitivity of different organs or tissues to radiation.  


8.14. How does the dose based on the residual radioactivity levels relate to background dose levels ?

This dose can be compared with the background dose which differs from country to country.

"Background radiation" means radiation from cosmic sources, naturally occurring radioactive material, including radon, and global fallout as it exists in the environment from the testing of nuclear explosive devices or from earlier nuclear accidents, such as Chernobyl, that contributes to background radiation and is not under the control of the licensee.

"Distinguishable from background" means that the detectable concentration of a radionuclide is statistically different from the background concentration of that radionuclide in the vicinity of the site.


8.16. Is it possible that some isotopes are located in such a way that radiation-monitoring devices cannot accurately detect their levels of radioactivity ?

It is unlikely that radioactive material located inside a piece of equipment or a structure is not detected during the final radiation survey.

The structures, systems, and components that have radioactive contamination exceeding "legal limits" will be decontaminated or dismantled and shipped to a low-level-waste disposal site.

The licensee must keep records of information during the operating phase of the facility that could be used to identify where spills or other occurrences involving the spread of contamination in and around the facility, equipment, or site have been located.  


8.17. Will continued monitoring be required after the decommissioning process is complete to ensure that the radiation levels do not increase ?

No. For sites that have been determined to be acceptable for unrestricted use, there are no requirements for further measurement of radiation levels.

It is not expected that these radiation levels would change -- other than to be reduced over time -- because the radioactive material will have been removed from the site, and there would be no mechanism for further contamination or radiological releases.  


8.18. What types of uses can be made of the plant site after decommissioning is completed ?

Once the license has been terminated and the site released for unrestricted use, there are no restrictions on the type of use.

Possible uses could range from restoring the natural habitat, to farming, to continued use as an industrial site (possibly leaving buildings and installing a gas-, coal-, or oil-powered generating plant).  


8.19. Could the licensee initiate an alternative use of the site or partial site release before the decommissioning process is completed ?

Requests by licensees to initiate alternative uses of the site or site partial release before the decommissioning process is completed, would be reviewed by the Commission on a case-by-case basis.




9. Hazards Associated with Decommissioning

TABLE OF CONTENTS FOR THIS SECTION

9.1. Workers
9.1.1. Where do the decommissioning workers come from ?
9.1.2. Is worker safety considered in the planning for and review of decommissioning ?
9.1.3. How much occupational dose is received by workers during decommissioning ?
9.1.4. Are there limits on the amount of occupational dose that may be received ?
9.1.5. Does the licensee have to estimate the occupational dose before the decommissioning process is initiated ?
9.2. Public and Environment
9.2.1. Is the safety of the public considered in the planning for and review of decommissioning ?
9.2.2. How much dose will the public receive during the decommissioning process ?
9.2.3. Who estimates what the doses are, and how are these estimates made ?
9.2.6. Will there be continued environmental monitoring of the site and the offsite areas to measure releases of radioactive material during the decommissioning process ?
9.2.7. Who will perform the environmental monitoring ?
9.2.11. What types of accidents at the reactor site are considered and what would be the consequences to the public ?
9.3. General
9.3.1. In general, how safe is a decommissioning plant in contrast to an operating plant ?  
9.3.2. Will there still be emergency preparedness plans and warning sirens in the vicinity of the plant ?
9.3.3. What measures are taken to prevent vandalism and sabotage during decommissioning ?




9.1. Workers

9.1.1. Where do the decommissioning workers come from ?

The majority of workers for an immediate decontamination and dismantling programme will likely be people who worked on the operating plant. These workers are most familiar with the facility and its history.

Some jobs, however, may be contracted out to companies that have gained experience at other plants in specialized areas of decommissioning or dismantling.

There will be very few employees during the storage phase in facilities that are placed in a storage mode.

A new group of workers will likely need to be hired who are most likely unfamiliar with the plant, but who will probably have had some decommissioning experience at other facilities.


9.1.2. Is worker safety considered in the planning for and review of decommissioning ?

Yes. Worker safety is considered both in terms of the radiological hazard (their exposure to radiation) and in terms of industrial safety.


9.1.3. How much occupational dose is received by workers during decommissioning ?

The amount of occupational dose received during the decommissioning process will depend on the design and size of the facility as well as on the plans for decommissioning.

A greater amount of occupational dose is anticipated to be incurred for an immediate decontamination and dismantling than for a storage period followed by dismantling.

The person-rem numbers are the doses that are received by all the workers.  


9.1.4. Are there limits on the amount of occupational dose that may be received ?

Yes. The regulations state that the licensee shall control the occupational dose to individual adults to an annual limit, depending on the regulation in the country, -- total effective dose equivalent to the entire body -- or to an organ dose.

There are also dose limits to the eyes, the skin, and the extremities.  


9.1.5. Does the licensee have to estimate the occupational dose
before the decommissioning process is initiated ?

No. However, at the time that the license-termination plan is updated, the licensee is required to update its environmental report as appropriate to reflect any new information or significant environmental change associated with the applicant's proposed decommissioning activities.

The environmental report contains an estimate of occupational dose, so the licensee needs to estimate the occupational dose for decommissioning to determine if the estimates are within the range given in the environmental report for routine operations.

However, it becomes also a challenge to perform "decommissioning strategies" leading to the lowest dose uptake by workers, based on one hand on the history of the plant and on the other hand mostly on the experience gained by "decommissioners", following the so-called "Alara Principle".  




9.2. Public and Environment

9.2.1. Is the safety of the public considered in the planning for and review of decommissioning ?

Yes. The safety of the public is a major concern, even though the potential for hazards to the public from the decommissioning process and potential accidents is much less than it is when the facility is operating.


9.2.2. How much dose will the public receive during the decommissioning process ?

The major source of exposure to the public is from the shipment of low-level waste from the reactor site to the low-level waste disposal site.

The radiation dose is received by people who travel along the same route as the trucks that are transporting low-level waste if not on a private way.

However, because of the variability in the timing of each shipment, the short period of time that any person would be near any of the trucks, and the small dose that is allowed 6 feet from the side of a truck, the dose that is received by any one person traveling down the highway or stopped at a rest stop is a very small fraction of the annual dose that the person would receive from background radiation.

Minor sources of exposure to the public include radioactive effluent releases during the decommissioning process as discussed in the response to "Question 9.2.4.".  


9.2.3. Who estimates what the doses are, and how are these estimates made ?

The licensee estimates the doses.

The doses are estimated using assumptions about the amount of radioactive material that will be released or the proximity of the public to the source of radiation.

The doses are calculated using approved scenarios, assumptions, parameter values, and conceptual models. The regulator reviews the licensee's estimates of the doses and often recalculates the doses using its own assumptions for activities with the potential for significant worker exposure.  


9.2.6. Will there be continued environmental monitoring of the site and the offsite areas to measure releases of radioactive material during the decommissioning process ?

Yes. The radiological environmental monitoring programme that was in place at the nuclear plant will continue even after the plant is shut down.

The programme will be modified to appropriately monitor the types of releases that may occur during decommissioning and to monitor results at appropriate intervals of time. Not all measurements will be made on a continuous basis.

The licensee uses the results of the environmental monitoring programme to calculate the dose to the public.  


9.2.7. Who will perform the environmental monitoring ?

The Radiological Environmental Monitoring Programme is conducted by the licensee. The procedures and results of the Radiological Environmental Monitoring Programme are inspected and reviewed by the regulator to ensure that all requirements are being met.


9.2.11. What types of accidents at the reactor site are considered and what would be the consequences to the public ?

Once the reactor permanently shuts down, the risk to the public is greatly reduced; however, there are still several accidents that may occur with consequences offsite.

The accidents that have the potential for the greatest offsite doses are those that involve the spent fuel that has recently been moved from the reactor to the spent fuel pool. Over time, the hazard from the spent fuel diminishes as the radioactive material in the fuel decays.

Licensees are required to examine their sites and decommissioning plans to identify postulated accidents that could occur during decommissioning. An analysis of these accidents is required in their "Final Safety Analysis Report", which is part of the licensing basis for the plant.

Except for the fuel-related accidents in the first year(s) after the facility ceases operation, the offsite consequences of these accidents are very small and do not require offsite emergency response.

Examples of the types of accidents that are considered by the licensees include :
  • Cask or heavy load-handling accident with a subsequent drop into spent fuel pool
  • Loss of cooling for the spent fuel pool or loss of water from the spent fuel pool
  • Materials handling event (non-fuel)
  • Radioactive liquid waste releases
  • Accidents from handling spent resin
  • Fire
  • Explosions
  • "External events"
  • Transportation accidents.

If a licensee requests an exemption to a regulation because they believe it no longer applies due to the decommissioning state of the plant, they must show that the regulation is not needed including consideration of the risk to the public. Additional information regarding the consequences of spent fuel pool accidents is given in the response to "Question 5.8.4.".  



9.3. General


9.3.1. In general, how safe is a decommissioning plant in contrast to an operating plant ?

At the time that the plant permanently ceases operations and the fuel is removed from the reactor, the risk to the public from an accident drops significantly.  


9.3.2. Will there still be emergency preparedness plans and warning sirens in the vicinity of the plant ?

For some period of time after the licensee ceases reactor operations, the offsite emergency preparedness planning will be maintained.

This period of time depends on when the reactor was last critical as well as onsite-specific considerations. Offsite emergency planning may be eliminated when the fuel has been removed from the reactor and placed in the spent fuel pool, and sufficient time has elapsed, and there are no longer any postulated accidents that would result in offsite dose consequences that are large enough to require offsite emergency planning.

There would be no requirement to maintain offsite systems to warn the public. Onsite emergency plans will be required for both the spent fuel pool and the Independent Spent Fuel Storage Installations, but offsite plans will not be required. If, however, an operating plant is located at the same site as the decommissioning plant, the emergency preparedness plans will still be in effect for the operating plant.  


9.3.3. What measures are taken to prevent vandalism and sabotage during decommissioning ?

The facility is required to have a security plan when a plant is being decommissioned; however, as the hazards are removed from the nuclear reactor site, security requirements are modified.

Many plants reduce the area that they keep very secured to the area around where the spent fuel is stored. This is known as the "nuclear island".

Reducing the size of the area that has strict security measures allows for better control of the material that must be safeguarded. Security measures for the nuclear island are designed to prevent sabotage or removal of the nuclear material.  



10. Finances

In order to avoid any kind of misunderdstanding, this topic will be treated more in detail, following if possible the FAQ's, when the "right information" will be got from each involved company/institution/country.

Anyway, in any case, the total cost of decommissioning depends on many factors, including the sequence and timing of the various stages of the program, location of the facility, current radioactive waste burial costs, and plans for spent fuel storage.

The costs are adjusted annually, as further specified in the regulations. Actual site-specific costs incurred and estimated costs of decommissioning give a better indication of what the process costs.

Decommissioning costs vary, based on plant size and design, local labor and radiological waste burial costs, and the specific process that is being used for decommissioning.

The licensee makes the site specific estimates of the decommissioning costs or hires a contractor who has extensive experience in making these estimates.

An estimate is made at or about 5 years preceeding the projected end of operations. At this time, power reactor licensees shall submit a preliminary decommissioning cost estimate, which includes an up-to-date assessment of the major factors that could affect the cost to decommission.

If the amount of money available is inadequate, the licensee has approximately 5 years to adjust the money in the decommissioning trust fund to ensure that appropriate funds are available for decommissioning.  




11. Socio-economic Issues


TABLE OF CONTENTS FOR THIS SECTION

11.1. What impact would each of the alternatives have on the economy of the surrounding area, including work-force requirements ?
11.2. How many people are needed onsite during the decommissioning process ? Is this more than during operations ?


11.1. What impact would each of the alternatives have on the economy of the surrounding area, including work-force requirements ?

The biggest socio-economic impact occurs before decommissioning starts, at the time the plant ceases operations, and the tax income created by the plant is substantially reduced.

The surrounding communities may find their property tax base reduced in half or more, depending on the presence or absence of other industries in the area : it is the biggest "economic impact".

Typically, additional public services are not required during decommissioning because the plant staff will be smaller than the operating staff : it is the biggest "social impact".  


11.2. How many people are needed onsite during the decommissioning process ? Is this more than during operations ?

After cessation of operations, the number of workers in the plant will be reduced.

Plants that are currently being decommissioned using the "DECON" alternative have work forces in the range of approximately one-third to one-tenth the number of persons who were employed at the plant during its operation.

These personnel are periodically supplemented with contract personnel during major decommissioning activities, such as the removal of large components like the steam generators and pressurizer.

If the plant were placed in "SAFSTOR", the number of workers would be further reduced. Decommissioning plants that are located at the same site as operating facilities generally have a staff of 20 or fewer during SAFSTOR. Single-unit plants (not located next to operating units) require a larger staff and may have 20 to 70 employees during SAFSTOR. After the SAFSTOR period, the number of workers would increase to the range of one-third to one-tenth the number of persons who were employed at the plant during its operation, and would be further supplemented with contractor personnel for the final cleanup of the site.



12. Public Involvement

The following section has to be checked with people involved in this field.



13. Getting Additional Information

The following section has to be checked with people involved in this field.

NRC's website contains information of interest to the public (http://www.nrc.gov). This site also has a link to the "Public Electronic Reading Room". Copies can also be read online or downloaded electronically from the NRC's website.



14. Bibliography

See "original F.A.Qs" from the NRC.

References

1. Decommissioning Nuclear Facilities, World Nuclear Association, (updated June 2011).

2. Comparative analysis of the Oskarshamn 3 and Barsebäck site decommissioning studies. Bertil Hansson, Bewon Lars-Olof Jönsson, Barsebäck Kraft AB January 2009. ISSN 1402-3091. SKB Rapport R-09-55. Commissioned by Svensk Kärnbränslehantering AB, Swedish Nuclear Fuel and Waste Management Co. Box 250, SE-101 24 Stockholm SWEDEN. Phone +46 8 459 84 00. http://www.skb.se  



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