April 2006 6-1 NUREG-1817

6.0 Fuel Cycle, Transportation, and Decommissioning

This chapter addresses the environmental impacts from (1) the uranium fuel cycle and solid

waste management, (2) transportation of radioactive material, and (3) decommissioning for the

proposed Grand Gulf early site permit (ESP) site. Distinctions between the impacts of

advanced light-water reactor (LWR) designs and the gas-cooled reactor designs are discussed. |

In its evaluation of uranium fuel cycle impacts for the Grand Gulf ESP site, System Energy

Resources, Inc. (SERI) used the plant parameter envelope (PPE) approach for the advanced

LWR designs but not for the two gas-cooled reactors. In its evaluation of the impacts from

transportation of radioactive materials, SERI did not use the PPE approach, but rather

evaluated each reactor design individually. An applicant for a construction permit (CP) or a |

combined license (COL) referencing the Grand Gulf ESP would, therefore, have to perform a |

new evaluation if a different design is proposed at that stage. |

6.1 Fuel Cycle Impacts and Solid Waste Management

This section discusses the environmental impacts from the uranium fuel cycle and solid waste

management for both the advanced LWR designs and gas-cooled reactor designs. The

impacts of the two types of design are presented separately because Title 10 of the Code of

Federal Regulations (CFR), Section 51.51 (10 CFR 51.51) provides specific criteria for

evaluating the environmental impacts for only LWR designs. Consequently, issues related to |

fuel cycle impacts and solid waste management are not resolved because of the lack of data to |

validate impacts from gas-cooled designs.

6.1.1 Light Water Reactors

The regulations in 10 CFR 51.51(a) state that |

Every environmental report prepared for the construction permit stage of a light water |

cooled nuclear power reactor, and submitted on or after September 4, 1979, shall take

Table S–3, Table of Uranium Fuel Cycle Environmental Data, as the basis for evaluating

the contribution of the environmental effects of uranium mining and milling, the

production of uranium hexafluoride, isotopic enrichment, fuel fabrication, reprocessing of

irradiated fuel, transportation of radioactive materials and management of low level

wastes and high level wastes related to uranium fuel cycle activities to the |

environmental costs of licensing the nuclear power reactor. Table S–3 shall be included

in the environmental report and may be supplemented by a discussion of the

environmental significance of the data set forth in the table as weighed in the analysis

for the proposed facility.
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-2 April 2006

| The PPE for the new unit or units at the Grand Gulf ESP site uses the bounding input

parameters from the following LWR designs:

| C Advanced CANDU (CANada Deuterium Uranium Reactor) (ACR-700) – This reactor,

developed by Atomic Energy Canada Limited, is an evolutionary extension of CANDU 6

plant using very slightly enriched uranium fuel and light water coolant.

 C Advanced Boiling Water Reactor (ABWR) – This reactor, developed by General Electric

Company, is a standardized plant that has been certified under the U.S. Nuclear

| Regulatory Commission (NRC) requirements in 10 CFR Part 52 (Appendix A). The

ABWR is fueled with slightly enriched uranium and uses a light water cooling system.

 C Advanced Pressurized Water Reactor (AP1000) – This is an earlier version of the

AP1000 reactor final design developed by Westinghouse Electric Company and

subsequently approved by the NRC, using slightly enriched uranium and a light water

cooling system. This design is not the AP1000 that has received final design approval

from the NRC; therefore, this design will be referred to as the “surrogate AP1000.”

 C Economic Simplified Boiling Water Reactor (ESBWR) – This reactor, developed by

General Electric Company, is fueled with slightly enriched uranium and uses a light 

water cooling system.

 C International Reactor Innovative and Secure (IRIS) next-generation pressurized water

reactor (PWR) – This reactor, under development by a consortium led by Westinghouse

Electric Company, is a modular LWR.

| These light water designs all use uranium dioxide fuel; therefore, Table S–3 (10 CFR 51.51(b))

can be used to assess environmental impacts. Table S–3 values are normalized for a

| reference 1000-MW(e) LWR at an 80 percent capacity factor. The 10 CFR 51.51(b) Table S–3

values are reproduced in Table 6-1. The PPE power rating for the Grand Gulf ESP site is

| 8600 MW(t) with a net electrical output of up to 3000 MW(e) (SERI 2005a).

Specific categories of natural resource use are included in Table S–3 (see Table 6-1). These

categories relate to land use, water consumption and thermal effluents, radioactive releases,

burial of transuranic and high-level and low-level wastes, and radiation doses from

| transportation and occupational exposures. In developing Table S–3, the staff considered two

fuel cycle options–no recycle and uranium-only recycle–that differed in the treatment of spent

fuel removed from a reactor. “No recycle” treats all spent fuel as waste to be stored at a

| Federal waste repository; “uranium-only recycle” involves reprocessing spent fuel to recover

unused uranium and return it to the system. Neither cycle involves the recovery of plutonium. 

The contributions in Table S–3 resulting from reprocessing, waste management, and

| transportation of wastes are maximized for both of the two fuel cycles (uranium-only and no
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-3 NUREG-1817

recycle); that is, the identified environmental impacts are based on the cycle that results in the

greater impact. The uranium fuel cycle is defined as the total of those operations and

processes associated with provision, utilization, and ultimate disposition of fuel for nuclear

power reactors. 

Table 6-1. Table S–3 from 10 CFR 51.51(b), Table of Uranium Fuel Cycle Environmental |

Data1

Environmental considerations Total Maximum effect per annual fuel requirement or

reference reactor year of model 1,000 MWe LWR

Natural Resource Use 

Land (acres):

Temporarily committed2............ ..... 100

Undisturbed area ............ ........ 79

Disturbed area ............ .......... 22 Equivalent to a 110 MWe coal-fired power plant.

Permanently committed ............ ..... 13

Overburden moved (millions of MT) ........ 2.8 Equivalent to 95 MWe coal-fired power plant.

Water (millions of gallons):

Discharged to air ............ .......... 160 =2 percent of model 1,000 MWe LWR with cooling 

 tower.

Discharged to water bodies ............ . . 11,090

Discharged to ground ............ ....... 127

Total ............ ......... 11,377 <4 percent of model 1,000 MWe LWR with once-through

 cooling.

Fossil fuel:

Electrical energy (thousands of MW-hr) ..... 323 <5 percent of model 1,000 MWe LWR output. 

Equivalent coal (thousands of MT) ......... 118 Equivalent to the consumption of a 45 MWe coal-fired

power plant.

Natural gas (millions of standard cubic feet) . 135 <0.4 percent of model 1,000 MWe energy output. 

Effluents–Chemical (MT)

Gases (including entrainment):3

SOx............ ............ ......... 4,400

NOx

4

............ ............ ........ 1,190 Equivalent to emissions from 45 MWe coal-fired plant 

 for a year.

Hydrocarbons ............ ............ . 14

CO ............ ............ ......... 29.6

Particulates ............ ............ ... 1,154
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-4 April 2006

Table 6-1. (contd)

Environmental considerations Total Maximum effect per annual fuel requirement or

reference reactor year of model 1,000 MWe LWR

Other gases: 

F ............ ............ ........... .67 Principally from UF6 production, enrichment, and

reprocessing. Concentration within range of state

standards—below level that has effects on human

health. 

HCI ............ ............ ......... .014

Liquids:

SO–

4

............ ............ ........ 9.9 From enrichment, fuel fabrication, and reprocessing

steps. Components that constitute a potential for

adverse environmental effect are present in dilute

concentrations and receive additional dilution by

receiving bodies of water to levels below permissible

standards. The constituents that require dilution and

the flow of dilution water are: NH3–600 cfs.,

NO3–20 cfs., Fluoride–70 cfs.

NO–

3

............ ............ ........ 25.8

Fluoride ............ ............ ...... 12.9

Ca++............ ............ ......... 5.4

Cl-............ ............ .......... 8.5

Na+............ ............ ......... 12.1

NH3............ ............ ......... 10.0

Fe ............ ............ .......... .4

Tailings solutions (thousands of MT) ....... 240 From mills only—no significant effluents to environment.

Solids ............ ............ ....... 91,000 Principally from mills-no significant effluents to

environment.

Effluents—Radiological (curies)

Gases (including entrainment):

Rn–222 ............ ............ ...... ....... Presently under reconsideration by the Commission.

Ra–226 ............ ............ ...... .02

Th–230 ............ ............ ...... .02

Uranium ............ ............ ..... .034

Tritium (thousands) ............ ......... 18.1

C–14 ............ ............ ........ 24

Kr–85 (thousands) ............ ......... 400

Ru–106 ............ ............ ...... .14 Principally from fuel reprocessing plants.

I–129 ............ ............ ........ 1.3

I–131 ............ ............ ........ .83

Tc–99 ............ ............ ....... ....... Presently under consideration by the Commission.

Fission products and transuranics ......... .203

Liquids:

Uranium and daughters ............ ..... 2.1 Principally from milling—included tailings liquor and

returned to ground—no effluents; therefore, no effect

on environment.

Ra–226 ............ ............ ...... .0034 From UF6 production. 

Th–230 ............ ............ ...... .0015

Th–234 ............ ............ ...... .01 From fuel fabrication plants—concentration 10 percent of

10 CFR 20 for total processing 26 annual fuel

requirements for model LWR. 

Fission and activation products ........... 5.9x10-6
Fuel Cycle, Transportation, and Decommissioning

Table 6-1. (contd)

Environmental considerations Total Maximum effect per annual fuel requirement or

reference reactor year of model 1,000 MWe LWR

April 2006 6-5 NUREG-1817

Solids (buried on site): 

Other than high level (shallow) ............ 11,300 9,100 Ci comes from low level reactor wastes and

1,500 Ci comes from reactor decontamination and

decommissioning—buried at land burial facilities.

600 Ci comes from mills—included in tailings returned

to ground. Approximately 60 Ci comes from conversion

and spent fuel storage. No significant effluent to the

environment. 

TRU and HLW (deep) ............ ....... 1.1x107 Buried at Federal Repository.

Effluents—thermal (billions of British thermal

units) ............ ............ ...........

4,063 <5 percent of model 1,000 MWe LWR

Transportation (person-rem):

Exposure of workers and general public ..... 2.5

Occupational exposure (person-rem) ....... 22.6 From reprocessing and waste management.

1

 In some cases where no entry appears it is clear from the background documents that the matter was addressed and

that, in effect, the Table should be read as if a specific zero entry had been made. However, there are other areas that

are not addressed at all in the Table. Table S–3 does not include health effects from the effluents described in the

Table, or estimates of releases of Radon-222 from the uranium fuel cycle or estimates of Technetium-99 released from

waste management or reprocessing activities. These issues may be the subject of litigation in the individual licensing

proceedings. 

 Data supporting this table are given in the “Environmental Survey of the Uranium Fuel Cycle,” WASH–1248, April

1974; the “Environmental Survey of the Reprocessing and Waste Management Portion of the LWR Fuel Cycle,”

NUREG–0116 (Supp.1 to WASH–1248, NRC 1976); the “Public Comments and Task Force Responses Regarding the

Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle,” NUREG–0216

(Supp. 2 to WASH–1248) (NRC 1977b); and in the record of the final rulemaking pertaining to Uranium Fuel Cycle

Impacts from Spent Fuel Reprocessing and Radioactive Waste Management, Docket RM–50–3. The contributions

from reprocessing, waste management and transportation of wastes are maximized for either of the two fuel cycles

(uranium only and no recycle). The contribution from transportation excludes transportation of cold fuel to a reactor and

of irradiated fuel and radioactive wastes from a reactor which are considered in Table S–4 of §51.20(g). The

contributions from the other steps of the fuel cycle are given in columns A-E of Table S–3A of WASH-1248. 

2

 The contributions to temporarily committed land from reprocessing are not prorated over 30 years, since the

complete temporary impact accrues regardless of whether the plant services one reactor for one year or 57 reactors for

30 years. 

3

 Estimated effluents based upon combustion of equivalent coal for power generation. 

4

 1.2 percent from natural gas use and process. 

During the Carter administration, the Nuclear Nonproliferation Act of 1978, Pub. L. No. 95-242 |

(22 USC 3201 et seq.), was enacted; it significantly impacted the disposition of spent nuclear |

fuel by indefinitely deferring the commercial reprocessing and recycling of plutonium produced |

in the U.S. commercial nuclear power program. While the ban on the reprocessing of spent |

fuel was lifted during the Reagan administration, economic circumstances changed, reserves of |

uranium ore increased, and the stagnation of the nuclear power industry provided little incentive |

for industry to resume reprocessing. During the 109

th Congress, the Energy Policy Act of 2005, |

Pub. L. No. 109-58 (119 Stat. 594 [2005]), was enacted. It authorized the U.S. Department of |

Energy (DOE) to conduct an advanced fuel recycling technology research and development
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-6 April 2006

| program to evaluate proliferation-resistant fuel recycling and transmutation technologies that

| minimize environmental or public health and safety impacts. Consequently, while Federal policy

| does not prohibit reprocessing, additional DOE efforts would be required before commercial

| reprocessing and recycling of spent nuclear fuel produced from U.S. commercial nuclear power

| plants could commence. 

| The no-recycle option is presented schematically in Figure 6-1. Natural uranium is mined in

| either open-pit or underground mines or by an in situ leach solution mining process. In situ

leach mining, the primary form of mining in the United States today, involves injecting a lixiviant

solution into the uranium ore body to dissolve uranium and then pumping the solution to the

| surface for further processing. The ore or in situ leach solution is transferred to mills where it is

processed to produce uranium oxide or “yellowcake.” A conversion facility prepares the

| uranium oxide by converting it to uranium hexafluoride, which is then processed at an

| enrichment facility to increase the percentage of the more fissile isotope uranium-235 and

decrease the percentage if the non-fissile isotope uranium-238. At a fuel-fabrication facility, the

enriched uranium, which is approximately 5 percent uranium-235, is then converted to uranium

dioxide (UO2). The UO2 is pelletized, sintered, and inserted into tubes to form fuel assemblies. 

The fuel assemblies are placed in the reactor to produce power. When the content of the

uranium-235 reaches a point where the nuclear reactor has become inefficient with respect to

neutron economy, the fuel assemblies are withdrawn from the reactor. After onsite storage for

sufficient time to allow for short-lived fission product decay and to reduce the heat generation

| rate, the fuel assemblies would be transferred to a waste repository. Disposal of spent fuel

elements in a repository constitutes the final step in the no-recycle option.

Figure 6-1. The Uranium Fuel Cycle: No-Recycle Option (derived from NRC 1999)
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-7 NUREG-1817

The following assessment of the environmental impacts of the fuel cycle as related to the

operation of the proposed project is based on the values given in Table S–3 (see Table 6-1)

and the staff’s analysis of the radiological impact from radon and technetium. In the Generic |

Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS) (NRC 1996), |

the staff provides a detailed analysis of the environmental impacts from the uranium fuel cycle. 

Although the GEIS is specific to the impacts related to license renewal, the information is |

relevant to this review because the advanced LWR designs considered here use the same type

of fuel; the staff’s analyses in Section 6.2.3 of the GEIS are summarized and incorporated by |

reference here.

The fuel cycle impacts in Table S–3 are based on a reference 1000-MW(e) LWR operating at

an annual capacity factor of 80 percent for a net electric output of 800 MW(e). In the following

review and evaluation of the environmental impacts of the fuel cycle, the staff used the stated |

capacity factor in the SERI PPE of 96 percent with a total net electric output of 3000 MW(e) for |

the ESP site (SERI 2005a); this results in approximately four times the impact values in |

Table S–3 (see Table 6-1). Throughout Chapter 6, this will be referred to as the 1000-MW(e) |

LWR scaled model, reflecting 3000 MW(e) for the site.

Recent changes in the fuel cycle may have some bearing on environmental impacts; however,

as discussed below, the staff is confident that the contemporary fuel cycle impacts below are

those identified in Table S–3.

The values in Table S–3 were calculated from industry averages for the performance of each

type of facility or operation within the fuel cycle. Recognizing that this approach meant that

there would be a range of reasonable values for each estimate, the staff followed the policy of

choosing the assumptions or factors to be applied so that the calculated values would not be

underestimated. This approach was intended to ensure that the actual environmental impacts

would be less than the quantities shown in Table S–3 for all LWR nuclear power plants within

the widest range of operating conditions. Many subtle fuel cycle parameters and interactions

were recognized by the staff as being less than the precision of the estimates and were not

considered or were considered but had no effect on the Table S–3 calculations. For example,

to determine the quantity of fuel required for a year’s operation of a nuclear power plant in

Table S–3, the staff defined the model reactor as a 1000-MW(e) light water reactor operating at |

80-percent capacity with a 12-month fuel reloading cycle and an average fuel burnup of

33,000 MWd/MTU. This is a “reactor reference year” or “reference reactor year” depending on

the source (either Table S–3 or the GEIS), but it has the same meaning. The sum of the initial |

fuel loading plus all of the reloads for the lifetime of the reactor can be divided by the now more

likely 60-year (40-year initial operating license term and 20-year license renewal term) lifetime |

to obtain an average annual fuel requirement. This was done for both boiling water reactors

(BWRs) and PWRs; the higher annual requirement, 35 metric tonnes (MT) of uranium made

into fuel for a BWR, was chosen in the GEIS as the basis for the reference reactor year. A |

number of fuel management improvements have been adopted by nuclear power plant
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-8 April 2006

| managers to achieve higher performance and to reduce fuel and separative work (enrichment)

requirements. Since Table S–3 was promulgated, these improvements have reduced the

annual fuel requirement.

Another change is the elimination of the U.S. restrictions on the importation of foreign uranium. 

| The economic conditions of the uranium market favor utilization of foreign uranium at the

expense of the domestic uranium industry. These market conditions have forced the closing of

most U.S. uranium mines and mills, substantially reducing the environmental impacts in the

| United States from these activities. Factoring in changes to the fuel cycle suggests that the

| environmental impacts on mining and tail millings could drop levels below those given in

| Table S–3; however, Table S–3 estimates have not been reduced.

| Section 6.2 of the GEIS discusses the sensitivity to recent changes in the fuel cycle on the

environmental impacts in greater detail.

6.1.1.1 Land Use

The total annual land requirement for the fuel cycle supporting the 1000-MW(e) LWR scaled

| model is about 183 ha (452 ac). Approximately 20 ha (52 ac) are permanently committed land,

| and 162 ha (400 ac) are temporarily committed. A “temporary” land commitment is a

commitment for the life of the specific fuel cycle plant (e.g., a mill, enrichment plant, or

succeeding plants). Following completion of decommissioning, such land can be released for

unrestricted use. “Permanent” commitments represent land that may not be released for use

after plant shutdown and/or decommissioning because decommissioning activities do not result

in removal of sufficient radioactive material to meet the limits in 10 CFR Part 20, Subpart E for

| release of that area for unrestricted use. Of the 162 ha (400 ac) of temporarily committed land,

| 113 ha (280 ac) are undisturbed and 49 ha (120 ac) are disturbed (SERI 2005a). In

| comparison, a coal-fired power plant with the same MW(e) output as the LWR scaled model

| and that uses strip-mined coal requires the disturbance of about 324 ha (800 ac) per year for

| fuel alone. The staff concludes that the impacts on land use to support the 1000-MW(e) LWR

scaled model would be small.

6.1.1.2 Water Use

| The principal water use for the fuel cycle supporting a 1000-MW(e) LWR scaled model is that

required to remove waste heat from the power stations supplying electrical energy to the

enrichment step of this cycle. Scaling from Table S–3, of the total annual water use

of 1.72 x 108 m3 (4.55 x 1010 gal), about 1.7 x 108 m3 (4.44 x 1010 | gal) are required for the

removal of waste heat, assuming that these plants use once-through cooling. Other water
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-9 NUREG-1817

uses involve the discharge to air (e.g., evaporation losses in process cooling) of about

2.42 x 106 m3/yr (6.40 x 108 gal/yr) and water discharged to ground (e.g., mine drainage) of |

about 1.92 x 106 m3/yr (5.08 x 108 gal/yr). |

On a thermal effluent basis, annual discharges from the nuclear fuel cycle are about 4 percent

of the 1000-MW(e) LWR scaled model using once-through cooling. The consumptive water

use of 2.42 x 106 m3/yr (6.40 x 108 gal/yr) is about 2 percent of the 1000-MW(e) LWR scaled |

model using cooling towers. The maximum consumptive water use (assuming that all plants

supplying electrical energy to the nuclear fuel cycle use cooling towers) would be about

6 percent of the 1000-MW(e) LWR scaled model using cooling towers. Under this condition,

thermal effluents would be negligible. The staff concludes that the impacts on water use for

these combinations of thermal loadings and water consumption would be small relative to the

water use and thermal discharges.

6.1.1.3 Fossil Fuel Impacts

Electric energy and process heat are required during various phases of the fuel cycle process. 

The electric energy is usually produced by the combustion of fossil fuel at conventional power

plants. Electric energy associated with the fuel cycle represents about 5 percent of the annual

electric power production of the reference 1000-MW(e) LWR. Process heat is primarily

generated by the combustion of natural gas. This gas consumption, if used to generate

electricity, would be less than 0.4 percent of the electrical output from the model plant. The |

staff concludes that the fossil fuel impacts from the direct and indirect consumption of electric

energy for fuel cycle operations would be small relative to the net power production of the

proposed project.

6.1.1.4 Chemical Effluents

The quantities of chemical, gaseous, and particulate effluents from fuel cycle processes are

given in Table S–3 (see Table 6-1) for the reference 1000-MW(e) LWR. The quantities of

effluents would be approximately four times greater for the reference 1000-MW(e) LWR scaled

model. The principal effluents are SOx, NOx, and particulates. Based on data in The Seventh

Annual Report of the Council on Environmental Quality, these emissions constitute a small

additional atmospheric loading in comparison with the emissions from the stationary fuel

combustion and transportation sectors in the United States. The fuel cycle emissions constitute

about 0.08 percent of the annual national releases for each of these effluents (CEQ 1976).

Liquid chemical effluents produced in fuel cycle processes are related to fuel enrichment and

fabrication and may be released to receiving waters. These effluents are usually present in

dilute concentrations such that only small amounts of dilution water are required to reach levels

of concentration that are within established standards. Table S–3 (see Table 6-1) specifies the
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-10 April 2006

amount of dilution water required for specific constituents. Additionally, all liquid discharges into

the navigable waters of the United States from plants associated with the fuel cycle operations

will be subject to requirements and limitations set by an appropriate Federal, State, regional,

| local, or affected Native American Tribal regulatory agency.

Tailings solutions and solids are generated during the milling process and are not released in

quantities sufficient to have a significant impact on the environment.

The staff determined that the impacts of these chemical effluents would be small.

6.1.1.5 Radioactive Effluents

Radioactive effluents estimated to be released to the environment from waste management

activities and certain other phases of the fuel cycle process are set forth in Table S–3 (see

| Table 6-1). Using these data, the staff calculated for 1 year of operation of the 1000-MW(e)

| LWR scaled model, the 100-year involuntary environmental dose commitment to the

| U.S. population from the LWR-supporting fuel cycle. These calculations estimate that the

overall whole body gaseous dose commitment to the U.S. population from the fuel cycle

(excluding reactor releases and the dose commitments resulting from radon-222 and

technetium-99) would be approximately 16 person-Sv (1600 person-rem) per year of operation

of the 1000-MW(e) LWR scaled model; this reference reactor year is scaled to reflect the total

electric power rating for the site for a year.

The additional whole body dose commitment to the U.S. population from radioactive liquid

effluents from all fuel cycle operations other than reactor operation would be approximately

| 8 person-Sv (800 person-rem) per year of operation. Thus, the estimated 100-year

environmental dose commitment to the U.S. population from radioactive gaseous and liquid

releases because of these portions of the fuel cycle is approximately 24 person-Sv

(2400 person-rem) to the whole body per reference reactor year.

Currently, the radiological impacts associated with radon-222 and technetium-99 release are

not addressed in Table S–3. Principal radon releases occur during mining and milling

operations and as emissions from mill tailings, whereas principal technetium-99 releases occur

from gaseous diffusion enrichment facilities. SERI provided an assessment of radon-222 and

technetium-99 in its response to a request for additional information on February 3, 2005

| (SERI 2005b). The staff’s evaluation in this environmental impact statement (EIS) relied on the

| information discussed in the GEIS.

| In Section 6.2 of the GEIS, the staff estimated the radon-222 releases from mining and milling

operation, and from mill tailings for each year of operations of the reference 1000-MW(e) LWR. 

The estimated releases of radon-222 for the reference reactor year for the 1000-MW(e) LWR
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-11 NUREG-1817

scaled model, or for the total electric power rating for the site for a year, is approximately

7.7 x 1014 Bq (20,800 Ci). Of this total, about 78 percent would be from mining, 15 percent from

milling operations, 7 percent from inactive tails prior to stabilization. For radon releases from

stabilized tailings, the staff assumed that the scaled model would result in an emission of

1.5 x 1011 Bq (4 Ci) per site year; i.e., four times the GEIS estimate for the reference reactor |

year. The major risks from radon-222 are from exposure to the bone and the lung, although

there is a small risk from exposure to the whole body. The organ-specific dose weighting

factors from 10 CFR 20.1003 were applied to the bone and lung doses to estimate the 100-year

dose commitment from radon-222 to the whole body. The estimated 100-year environmental

dose commitment from mining, milling, and tailings prior to stabilization for each site year

(assuming the 1000-MW(e) LWR scaled model) would be approximately 37 person-Sv

(3700 person-rem) to the whole body. From stabilized tailings piles, the estimated 100-year

environmental dose commitment would be approximately 0.71 person-Sv (71 person-rem) to |

the whole body. Additional insights regarding Federal policy/resource perspectives concerning |

institutional controls comparisons with routine radon-222 exposure and risk and long-term |

releases from stabilized tailings piles are discussed in the GEIS. SERI provided an assessment |

of radon-222 and technetium-99 in its response to a request for additional information on

February 3, 2005 (SERI 2005b). The evaluation in this EIS relied on the information discussed |

in the GEIS. |

Also as discussed in the GEIS, the staff considered the potential health effects associated with |

the releases of technetium-99. The estimated releases of technetium-99 for the reference

reactor year for the 1000-MW(e) LWR scaled model are 1.1 x 109 Bq (0.03 Ci) from chemical |

processing of recycled uranium hexafluoride before it enters the isotope enrichment cascade

and 7.4 x 108 Bq (0.02 Ci) into the groundwater from a potential repository. The major risks |

from technetium-99 are from exposure of the gastrointestinal tract and kidney, although there is

a small risk from exposure to the whole body. Applying the organ-specific dose weighting

factors from 10 CFR 20.1003 to the gastrointestinal tract and kidney doses, the total-body

100-year dose commitment from technetium-99 to the whole body was estimated to be

4 person-Sv (400 person-rem) for the 1000-MW(e) LWR scaled model. |

Although radiation may cause cancers at high doses and high dose rates, currently there are no

data that unequivocally establish the occurrence of cancer following exposure to low doses

below about 100 mSv (10,000 mrem) and at low dose rates. However, radiation protection |

experts conservatively assume that any amount of radiation may pose some risk of causing |

cancer or a severe hereditary effect and that the risk is higher for higher radiation exposures. 

Therefore, a linear, no-threshold dose response model is used to describe the relationship |

between radiation dose and detriments such as cancer induction. A recent report by the |

National Research Council (2006), the BEIR VII report, supports the linear, no-threshold dose |

response model. Simply put, the theory states that any increase in dose, no matter how small, |
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-12 April 2006

results in an incremental increase in health risk. This theory is accepted by the NRC as a

conservative model for estimating health risks from radiation exposure, recognizing that the

model probably overestimates those risks.

Based on this model, the staff estimated the risk to the public from radiation exposure using the

nominal probability coefficient for total detriment (730 fatal cancers, nonfatal cancers, and

| severe hereditary effects per 10,000 person-Sv (1,000,000 person-rem)) from International

| Commission on Radiological Protection Publication 60 (ICRP 1991). This coefficient was

multiplied by the sum of the estimated whole body population doses discussed above,

| approximately 66 person-Sv/yr (6600 person-rem/yr), to calculate that the U.S. population

would incur a total of approximately 4.8 fatal cancers, nonfatal cancers, and severe hereditary

effects annually. This risk is very small compared to the number of fatal cancers, nonfatal

cancers, and severe hereditary effects that would be estimated to the U.S. population annually

from exposure to natural sources of radiation using the same risk estimation method.

Radon releases from tailings are indistinguishable from background radiation levels at a few

| kilometers from the tailings pile (at less than 1 km in some cases) (NRC 1978). The public

| dose limit in the U.S. Environmental Protection Agency’s (EPA’s) regulation, 40 CFR 190.10(a),

is 0.25 mSv/yr (25 mrem/yr) to the whole body from the entire fuel cycle, but most NRC

licensees have airborne effluents resulting in doses of less than 0.01 mSv/yr (1 mrem/yr)

(61 FR 65120).

| In addition, at the request of the U.S. Congress, the National Cancer Institute conducted a

| study and published “Cancer in Populations Living Near Nuclear Facilities” in 1990 (NCI 1990). 

This report included an evaluation of health statistics around all nuclear power plants, as well as

several other nuclear fuel cycle facilities, in operation in the United States in 1981 and found

“no evidence that an excess occurrence of cancer has resulted from living near nuclear

| facilities” (NCI 1990). The contribution to the annual average dose received by an individual

| from the fuel-cycle related radiation and other sources as reported in the National Council on

| Radiation Protection and Measurements (NCRP) Report 93 (NCRP 1987) is listed in Table 6-2. 

| The nuclear fuel cycle contribution to an individual’s annual average radiation is extremely small

| (less than 0.01 mSv (1 mrem) per year).

Based on the analyses presented above, the staff concludes that the environmental impacts of

radioactive effluents from the fuel cycle are small.

6.1.1.6 Radioactive Waste

The quantities of buried radioactive waste material (low-level, high-level, and transuranic

wastes) are specified in Table S–3 (see Table 6-1). For low-level waste disposal at land burial

facilities, the Commission notes in Table S–3 that there will be no significant radioactive 
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-13 NUREG-1817

Table 6-2. Comparison of Annual Average Dose Received by an Individual from All Sources

Source

Dose

(mSv/yr)(a) Percent of Total

Natural

Radon 2 55

Cosmic 0.27 8

Terrestrial 0.28 8

Internal (body) 0.39 11

Total natural sources 3 82

Artificial

Medical x-ray 0.39 11

Nuclear medicine 0.14 4

Consumer products 0.10 3

Total artificial sources 0.63 18

Other

Occupational 0.009 <0.30

Nuclear fuel cycle <0.01 <0.03

Fallout <0.01 <0.03

Miscellaneous sources <0.01 <0.03

(a) Multiply millisievert (mSv)/yr by 100 to obtain millirem/yr. |

Source: NCRP 1987 |

releases to the environment. For high-level and transuranic wastes, the Commission notes that

these are to be buried at a repository, such as the candidate repository at Yucca Mountain,

Nevada, and that no release to the environment is expected to be associated with such |

disposal, because it has been assumed that all of the gaseous and volatile radionuclides |

contained in the spent fuel are released to the atmosphere before the disposal of the waste. In

NUREG-0116 (NRC 1976), which provides background and context for the high-level and

transuranic Table S–3 values established by the Commission, the staff indicates that these

high-level and transuranic wastes will be buried and will not be released to the environment.

On February 15, 2002, subsequent to receipt of a recommendation by Secretary Abraham,

DOE, the President recommended the Yucca Mountain site for the development of a repository |

for the geologic disposal of spent nuclear fuel and high-level nuclear waste (White House Press |

Release 2002). |

The EPA developed Yucca Mountain-specific repository standards, which were subsequently |

adopted by the NRC in 10 CFR Part 63. In an opinion, issued July 9, 2004, the U.S. Court of

Appeals for the District of Columbia Circuit (the Court) vacated EPA’s radiation protection

standards for the candidate repository, which required compliance with certain dose limits over

a 10,000-year period (U.S. Court of Appeals 2004). The Court’s decision also vacated the |

compliance period in NRC’s licensing criteria for the candidate repository in 10 CFR Part 63. In |
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-14 April 2006

| response to the Court’s decision, EPA issued its proposed revised standards on August 22,

| 2005, that would revise the radiation protection standards for the candidate repository

| (70 FR 49014). In order to be consistent with EPA’s revised standards, NRC proposed

| revisions to 10 CFR Part 63 on September 8, 2005 (70 FR 53313). The 10 CFR Part 63

| rulemaking is titled “Implementation of a Dose Standard after 10,000 years,” and the comment

| period was extended to December 7, 2005. The proposed standards are 0.15 mSv (15 mrem)

| per year for 10,000 years following disposal and 3.5 mSv (350 mrem) per year for 10,000 years

| through 1 million years after disposal. RIN 3150 will not be finalized by the time this EIS is

| issued.

| Consequently at this time, for the high-level waste and spent fuel disposal component of the

fuel cycle, there is some uncertainty with respect to regulatory limits for offsite releases of

radionuclides for the current candidate repository site. However, prior to promulgation of the

affected provisions of the Commission’s regulations, the staff assumed that limits are

developed along the line of the 1995 National Academy of Sciences report, Technical Bases for

Yucca Mountain Standards, and that in accordance with the Commission’s Waste Confidence

| Decision, 10 CFR 51.23, a repository can and likely will be developed at some site, which would

comply with such limits, with peak doses to virtually all individuals of 1 mSv (100 mrem) per

year or less (NAS 1995; NRC 1996).

| Despite any uncertainty with respect to these rules, some judgement as to National

Environmental Policy Act of 1969 (NEPA) implications of offsite radiological impacts of spent

fuel and high-level waste disposal should be made. For the reasons stated above, the staff

concludes that the environmental impacts of radioactive waste disposal would be small.

6.1.1.7 Occupational Dose

In the review and evaluation of the environmental impacts of the fuel cycle, the staff considered

| the higher capacity factor in the PPE of 96 percent with a total net electric output of

| 3000 MW(e) for the ESP site (SERI 2005a). This case is referred to as the 1000-MW(e) LWR

scaled model. The annual occupational dose attributable to all phases of the fuel cycle for the

| 1000-MW(e) LWR scaled model is about 24 person-Sv (2400 person-rem). This is based on a

| 6 person-Sv (600 person-rem) occupational dose estimate attributable to all phases of the fuel

| cycle for the model 1000 MW(e) LWR (NRC 1996). The environmental impact from this

occupational dose is considered small because the dose to any individual worker is maintained

| within the dose limits of 10 CFR Part 20, which is 0.05 Sv/yr (5 rem/yr). 

6.1.1.8 Transportation

| The transportation dose to workers and the public totals about 0.025 person-Sv

| (2.5 person-rem) annually for the reference 1000-MW(e) LWR per Table S–3 (see Table 6-1). 
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-15 NUREG-1817

This corresponds to a dose of 0.1 person-Sv (10 person-rem) for the 1000-MW(e) LWR scaled |

model. For comparative purposes, the estimated collective dose from natural background

radiation to the population within 80 km (50 mi) of the Grand Gulf ESP site is 1020 person-Sv/yr

(102,000 person-rem/yr). On this basis of this comparison, the staff concludes that

environmental impacts of transportation would be small.

6.1.1.9 Conclusion

The staff evaluated the environmental impacts of the uranium fuel cycle as given in Table S–3

(see Table 6-1), considered the effects of radon-222 and technetium-99, and appropriately

scaled the impacts for the 1000-MW(e) LWR scaled model. Based on this evaluation, the staff

concludes that the impacts would be SMALL, and mitigation would not be warranted. The staff |

will verify the continued applicability of all assumptions should an applicant for a CP or COL |

reference the Grand Gulf ESP. |

6.1.2 Gas-Cooled Reactors

As noted earlier, issues related to reactors based on non-LWR designs are not resolved |

because of the lack of information to validate values and impacts. However, the following |

analyses were performed using data from SERI for the purposes of estimation only. |

The gas-cooled reactors analyzed for the uranium fuel cycle are:

 C Gas Turbine Modular Helium Reactor (GT-MHR) – This reactor, developed by General

Atomics, is a modular helium-cooled graphite-moderated reactor.

 C Pebble Bed Modular Reactor (PBMR) – This reactor, developed by PBMR (Pty) Ltd., is a

modular graphite-moderated helium-cooled gas turbine reactor.

Table S–3 from 10 CFR 51.51(a) can be used as a basis for bounding the environmental

impacts from the uranium fuel cycle only for LWRs. SERI performed an assessment of the

environmental impacts of the fuel cycle for gas-cooled reactor designs by comparing key

parameters for these reactor designs to those used to generate the impacts in Table S–3

(SERI 2005a). Key parameters are energy usage, material involved, and number of shipments |

for each major fuel cycle activity (i.e., mining, milling, conversion, enrichment, fuel fabrication,

and radioactive waste disposal). SERI sought to demonstrate in its environmental report that |

the impacts for the gas-cooled reactor designs were comparable to the environmental impacts

identified in the technical basis document, WASH-1248, Environmental Summary of the

Uranium Fuel Cycle, (AEC 1974) and its Supplement 1 (NUREG-0116) (NRC 1976) for |

Table S–3.
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-16 April 2006

| As discussed in Section 6.1.1, the fuel cycle impacts in Table S–3 (see Table 6-1) were based

on a reference 1000-MW(e) LWR operating at an annual capacity factor of 80 percent for a net

electric output of 800 MW(e). This is termed the “reference reactor year.” For the purposes of

evaluating fuel cycle impacts for the Grand Gulf ESP site, it was assumed that the additional

LWR site-wide fuel impacts would be based on a total net electric output of 3000 MW(e) at

| 96 percent annual capacity factor. This was termed the 1000-MW(e) LWR scaled model and

resulted in a factor about four times (i.e., 3000/800) the impacts in Table S–3.

One of the other-than-LWRs considered by SERI, the GT-MHR, is a four-module, 2400-MW(t),

nominal 1140-MW(e) unit assumed to operate at an annual capacity factor of 88 percent for a

| net electric output of 1003 MW(e). Therefore, the maximum number of GT-MHR units that

| could be sited at the Grand Gulf ESP site and remain near or below the 3000 MW(e) total net

| electric output PPE for the site is three (i.e., 3 x 1003). 

The second other-than-LWR considered by SERI, the PBMR, is an eight module, 3200-MW(t),

nominal 1320-MW(e) unit assumed to operate at an annual capacity factor of 95 percent for a

net electric output of 1253 MW(e). Therefore, the comparable number of PBMR units to remain

| below the 3000 MW(e) total net electric output PPE for the site is two (i.e., 2 x 1253).

| SERI (2005a) compared the impacts in Table S–3 for LWRs with those of the gas-cooled

reactor designs. The comparison used an annual fuel loading as a starting point and then

proceeded in reverse direction through the fuel cycle (fuel fabrication, enrichment, conversion,

milling, mining, radioactive waste). Table 6-3 provides an estimate of the impacts for each

| phase of the uranium fuel cycle assuming that the Grand Gulf ESP site would host three

| modular GT-MHR units or two modular PBMR units.

6.1.2.1 Fuel Fabrication

The quantity of UO2 required for reactor fuel is a key parameter. The more UO2 required, the

greater the environmental impacts (meaning, more energy, greater emissions, and increased

water usage). The 1000-MW(e) LWR scaled model described in Section 6.1.1 would require

the equivalent of 160 MT of enriched UO2| annually. This compares to 18 to 19 MT of enriched

UO2 annually for the gas-cooled reactor technologies (see Table 6-3).

GT-MHR fuel consists of microspheres of uranium oxycarbide coated with multiple layers of

pryrocarbon and silicon carbide referred to as TRISO coating. Two types of microspheres are

used in the GT-MHR fuel, one enriched to 19.8 percent uranium-235 and one with natural

uranium. The microspheres and graphite shims are bound together into a rod-shaped compact,

which is stacked into graphite blocks referred to as fuel elements. A reactor core consists of

1020 fuel elements. 
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-17 NUREG-1817

Table 6-3. Fuel Cycle Environmental Impacts from Gas-Cooled Reactor Designs for the Grand

Gulf Early Site Permit Site

Reactor Technology Facility/Activity

GT-MHR

4 Modules

(2400 MW(t) total

.1140 MW(e) total

88 percent capacity:

multiplier=3)

PBMR

8 Modules

(3200 MW(t) total

.1320 MW(e) total

95 percent capacity:

multiplier=2) |

Mining Operations

 Annual ore supply (million MT) 1.01 1.01 |

Milling Operations

 Annual yellowcake (MT) 909 606 |

UF6 Production

 Annual UF6 (MT) 1137 758 |

Enrichment Operations

 Enriched UF6 (MT) 24 25 |

 Annual separative work units (MT) 612 388 |

Fuel Fabrication Plant Operations

 Enriched UO2 (MT) 18 19 |

 Annual fuel loading (MT Uranium) 16 17 |

Solid Radioactive Waste

 Annual low-level waste from reactor operations 3300 Ci(a); 400 m3131 Ci; 2400 drums(a)

 Low-level waste from reactor decontamination

 and decommissioning (Ci per reference

 reactor year)

Data not available Data not available |

(a) Multiply curies (Ci) by 3.7 x 1010 to obtain becquerel (Bq). |

Notes:

- The enrichment separative work units (SWU) calculation was performed using the United States Enrichment

Corporation, Inc. (USEC) SWU calculator and assumes a 0.30 percent tails assay.

- The information on the reference reactor (mining, milling, UF6, enrichment, fuel fabrication values) was taken

from NUREG-0116, Table 3.2, (NRC 1976) no recycling.

- The information on the reference reactor (solid radioactive waste) was taken from 10 CFR 51.51, Table S-3.

- The calculated information on the reference reactor uses the same methodology as for the reactor

technologies.

- The normalized information is based on 1000 MW(e) and the reactor vendor-supplied unit capacity factor.

- For the new reactor technologies, the annual fuel loading was provided by the reactor vendor.

- The USEC SWU calculator also calculated the kgs of U feed. This number was multiplied by 1.48 to get the

necessary amount of UF6.

- The annual yellowcake number was generated using the relationship 2.61285 lb. of U3O8 to 1 kg U of UF6;

1.185 kgs of U3O8 to 1.48 kg.

- The annual ore supply was generated assuming an 0.1 percent ore body and a 90 percent recovery

efficiency.

- Co-60 with a 5.26 year half-life and Fe-55 with a 2.73 year half-life are the main nuclides listed for the PBMR

decontamination and decommissioning waste.

Source: 10 CFR 51.51(b), Table S-3 Table of Uranium Fuel Cycle Environmental Data |
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-18 April 2006

PBMR fuel consists of UO2 kernels (enriched to 12.9 percent uranium-235) that are TRISOcoated, similar to the GT-MHR fuel. The TRISO-coated particles are imbedded into a graphite

matrix to form a fuel sphere that is 60 mm in diameter. Each fuel sphere contains

approximately 15,000 TRISO-coated particles. Approximately 260,000 fuel spheres make up a

core of a single reactor module.

The fuel described above for gas-cooled reactors is fabricated differently than fuel for LWRs. 

There are no currently operating large-scale fuel fabrication facilities producing gas-cooled

reactor fuels in the United States; thus, a direct comparison of environmental impacts is not

possible. Based on some environmental impacts from a small-scale fuel fabrication facility

producing gas-cooled reactor fuel, SERI concluded that the environmental impacts from

producing gas-cooled reactor fuel would be “not inconsistent” with those of LWRs

| (SERI 2005a). By comparison with the fuel fabrication impacts for LWR technologies, the staff

concludes that the environmental impacts from producing gas-cooled reactor fuel likely would

| be small. However, these impacts would need to be assessed at the CP or COL stage, when

the staff would consider the environmental data that are available on a large-scale, fuel-

| fabrication facility for gas-cooled reactors should an applicant select one of these designs. 

6.1.2.2 Enrichment

| SERI (2005a) identified two quantities of interest for enrichment. These were (1) the amount of

energy required to enrich the fuel measured in separative work units (SWUs), and (2) the

amount of UF6 needed. A SWU is a measure of energy required to enrich the fuel. The major

environmental impacts for the entire uranium fuel cycle are from the emissions of the fossil fuel

plants used to supply energy for the gaseous diffusion plants that enrich the uranium. An

enrichment technology developed since the impacts in Table S–3 (see Table 6-1) were

developed and evaluated includes the gas centrifuge process that uses 90 percent less energy

| than the gaseous diffusion process (NRC 1996).

To produce 160 MT of enriched UO2 for the 1000-MW(e) LWR scaled model, the enrichment

plant needs to produce about 208 MT of UF6| , which requires approximately 500 MT of SWUs

(SERI 2005a). For gas-cooled reactor technologies, the needed enriched UF6| ranges from 24

to 25 MT of UF6. The amount of energy to produce these quantities of enriched UF6| for the

| gas-cooled reactor designs range from 388 to 612 MT of SWU. The upper range is up to

| 20 percent higher than the energy required for the reference LWR. SERI (2005a) concluded

that the large reduction in energy associated with using an alternate enrichment technology (for

example, centrifuge) and its associated environmental impacts would more than offset the

increase in SWUs. The staff concludes that, on balance, the environmental impacts of

enriching gas-cooled fuels by comparison with the impacts of enriching LWR fuel would likely
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-19 NUREG-1817

be small. However, these impacts would need to be assessed at the CP or COL stage, when |

the staff would consider impacts from the enrichment technology in use at that time, should an |

applicant select one of these designs. |

6.1.2.3 Uranium Hexafluoride Production – Conversion

There are two uranium conversion processes: a wet and a dry process. In the GEIS |

(NRC 1996), the NRC stated that environmental releases from the conversion facilities are

small compared to the overall fuel cycle impacts and that changing from 100 percent use of one

process to 100 percent use of the other would make no significant difference in the overall

impacts. Similar conversion technologies would be used today to produce UF6 as were

considered when determining the environmental impacts that were part of Table S–3 of

10 CFR 51.51(a) (see Table 6-1). |

The conversion facility would need to produce 1440 MT of UF6 annually for the reference |

1000-MW(e) LWR scaled model compared to 758 to 1137 MT of UF6 for the gas-cooled |

reactors based on the SWU calculator (SERI 2005a); see Table 6-3, footnote (a) above. The |

other-than-LWR values are less than the amount of UF6 required for the 1000-MW(e) LWR |

scaled model; therefore, the associated environmental impacts are expected to be less. On this |

basis, the staff concludes that the environmental impacts from producing UF6 for gas-cooled

reactors would be small.

6.1.2.4 Uranium Milling

Annual yellowcake (U3O8) production is the metric of interest for uranium milling. Plants

requiring less yellowcake production than the reference plant would require less energy, have

fewer emissions, and use less water.

The uranium mill for the 1000-MW(e) LWR scaled model would produce about 1200 MT of

yellowcake. The uranium mill for the gas-cooled reactor technologies would need to produce

606 to 909 MT of yellowcake, which is less than the amount of yellowcake needed for the |

1000 MW(e) LWR scaled model (SERI 2005a). On this basis, the staff concludes that the |

environmental impacts from uranium milling for the gas-cooled reactors would be small.

6.1.2.5 Uranium Mining

Annual ore supply is the metric of interest for uranium mining. The less ore mined, the smaller

the environmental impacts (i.e., less energy used, fewer emissions, less water usage). For the

1000-MW(e) LWR scaled model, 1.09 million MT of raw ore would be required to produce |

1200 MT of yellowcake. For the gas-cooled reactor technologies, the scaled ore requirements

range from 0.67 to 1.01 million MT of ore, a range that is comparable to the amount of ore |
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-20 April 2006

required for the reference 1000-MW(e) LWR scaled model. For this reason, the staff concludes

that the environmental impacts from uranium mining for the gas-cooled reactors would

be small.

6.1.2.6 Solid Low-Level Radioactive Waste – Operations

Table S–3 (see Table 6-1) of 10 CFR 51.51(a) states that there are 3.4 x 1014 | Bq (9100 Ci) of

| low-level waste generated annually from operations of the reference LWR; the 1000-MW(e)

LWR scaled model would result in 1.35 x 10

15 Bq (36,400 Ci) of low-level waste annually. 

Gas-cooled reactor technologies are projected to generate 4.8 x 1012 Bq to 1.2 x 1014 | Bq

| (131 to 3300 Ci) of low-level waste scaled annually, far below the amounts generated by the

| reference LWR (SERI 2005a). For this reason, the staff concludes that the environmental

impacts from low-level radioactive waste operations for gas-cooled reactors would be small.

6.1.2.7 Solid Low-Level Radioactive Waste – Decontamination and Decommissioning

In Table S–3 (see Table 6-1), the Commission states that 5.6 x 1013 Bq (1500 Ci) per reference

reactor year “...comes from reactor decontamination and decommissioning - buried at land

| burial facilities.” SERI (2005a) noted that gas-cooled reactor technologies would (1) generate

less waste than the reference 1000-MW(e) LWR, and (2) produce less heavy metal radioactive

waste because of the higher thermal efficiency and higher fuel burnup. The gas-cooled reactor

designs are also more compact than the reference LWR design, which would be expected to

| result in less decontamination and decommissioning waste (SERI 2005a). SERI concluded that

low-level waste impact from decontamination and decommissioning will be comparable to or

| less than that of the reference LWR (SERI 2005a). On this basis, the staff concludes that the

environmental impacts from solid low-level radioactive waste generated during decontamination

| and decommissioning for gas-cooled reactors would likely be small, but these impacts would

| need to be assessed again at the CP or COL stage if an applicant selects a gas-cooled design.

6.1.2.8 Conclusions

| The staff expects that the environmental impacts from the uranium fuel cycle activities and solid

| waste management activities for the proposed gas-cooled reactors likely would be small. 

However, because of the uncertainty in the final design of the gas-cooled reactors and the

| change in technology that could be applied to uranium fuel cycle activities, this issue is not

| resolved. Should an applicant reference one of these designs, additional reviews would be

| needed at the CP or COL stage in the following areas: fuel fabrication, enrichment, and solid

low-level waste operation during decontamination and decommissioning.
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-21 NUREG-1817

6.2 Transportation of Radioactive Materials

This section addresses both the radiological and nonradiological environmental impacts from

normal operating and accident conditions resulting from (1) shipment of unirradiated fuel to new

nuclear units at the Grand Gulf ESP site, (2) shipment of spent fuel to a monitored retrievable

storage facility or a permanent repository, and (3) shipment of low-level radioactive waste and

mixed waste to offsite disposal facilities. Distinctions between transportation impacts of

advanced LWR designs and gas-cooled reactor designs are discussed.

The NRC evaluated the environmental effects of transportation of fuel and waste for light water

nuclear power reactors in WASH-1238 (AEC 1972) and NUREG-75/038 (NRC 1975) and found |

the impact to be SMALL. These documents provided the basis for Table S–4 in 10 CFR 51.52,

which summarizes the environmental impacts of transportation of fuel and waste to and from

one LWR of 3000 to 5000 megawatts thermal (MW(t)) (1000 to 1500 MW(e)). Impacts are |

provided for normal conditions of transport and accidents in transport for a reference

1100-MW(e) LWR. 

Dose to transportation workers during normal transportation operations was estimated to result

in a collective dose of 0.04 person-Sv (4 person-rem) per reference reactor year. The |

combined dose to the public along the route and dose to onlookers were estimated to result in a |

collective dose of 0.03 person-Sv (3 person-rem) per reference reactor year. Environmental |

risks (radiological) during accident conditions were determined to be small. Nonradiological |

impacts during accident conditions were estimated as one fatal injury in 100 reference reactor

years and one nonfatal injury in 10 reference reactor years. Subsequent reviews of

transportation impacts in NUREG-0170 (NRC 1977a) and Sprung et al. (2000) concluded that

impacts were bounded by Table S–4 of 10 CFR 51.52. |

In accordance with 10 CFR 51.52(a), a full description and detailed analysis of transportation

impacts is not required when licensing an LWR (i.e., impacts are assumed bounded by

Table S–4) if an LWR meets the following criteria:

 C The reactor has a core thermal power level not exceeding 3800 MW(t).

 C Fuel is in the form of sintered UO2 pellets having a uranium-235 enrichment not

exceeding 4 percent by weight, and pellets are encapsulated in zirconium-clad fuel rods.

 C Average level of irradiation of the fuel from the reactor does not exceed

33,000 MWd/MT, and no irradiated fuel assembly is shipped until at least 90 days after it

is discharged from the reactor.
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-22 April 2006

 C With the exception of irradiated fuel, all radioactive waste shipped from the reactor is

packaged and in solid form.

 C Unirradiated fuel is shipped to the reactor by truck; irradiated fuel is shipped from the

reactor by truck, rail, or barge; and radioactive waste other than irradiated fuel is

shipped from the reactor by truck or rail.

The environmental impacts of the transportation of fuel and radioactive wastes to and from

nuclear power facilities were resolved generically in 10 CFR 51.52, provided that the specific

conditions in the rule (see above) are met; if not, then a full description and detailed analysis is

required for initial licensing. Once licensed, the NRC may consider requests to operate at

conditions above those in the facility’s licensing basis, for example, higher burnups,

enrichments, or thermal power levels above 33,000 MWd/MTU, 4 percent, and 3800 MW(t),

| respectively. Departures from the conditions itemized in 10 CFR 51.52(a) must be supported

| by a full description and detailed analysis of the environmental effects, as specified by

| 10 CFR 51.52(b).

SERI has not identified a specific reactor design for the Grand Gulf ESP site but used bounding

parameters from seven reactor designs. Five of the designs are LWRs and include the

| ACR-700 (3964 MW(t)/unit); the ABWR (4300 MW(t)/unit); the surrogate AP1000

| (3400 MW(t)/unit); the ESBWR (4000 MW(t)/unit), and the IRIS (3000 MW(t)/unit). For the

ACR-700 reactor design, two reactors make up a unit. For the IRIS design, three reactors

| (modules) make up a unit. For the remaining LWR designs, one reactor makes up a unit. 

None of the proposed LWR designs meet all the conditions in 10 CFR 51.52(a); therefore, a full

description and detailed analysis are required for each LWR design. This conclusion is based

on the following:

| C ACR-700, ABWR, and ESBWR designs exceed the 3800-MW(t) core thermal powerlevel limit.

 C ABWR, surrogate AP1000, ESBWR, and IRIS designs require fuel that exceeds the

uranium-235 enrichment of 4 percent.

 C ABWR, surrogate AP1000, ESBWR, and IRIS designs are expected to exceed the

average irradiation level of 33,000 MWd/MTU.

The remaining two designs are gas-cooled reactors: the GT-MHR and the PBMR. Each

GT-MHR unit is a four-module, 2400-MW(t), 1140-MW(e) gas-cooled reactor designed to

operate at a unit capacity factor of 88 percent. Each PBMR is an eight-module, 3200-MW(t),

| 1320-MW(e) gas-cooled reactor designed to operate at a unit capacity factor of 96 percent. 

These compare to the reference reactor in WASH-1238 (AEC 1972), which is a single-unit,
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-23 NUREG-1817

1100-MW(e) LWR with a unit capacity factor of 80 percent. The gas-cooled reactor designs do

not meet the conditions in 10 CFR 51.52(a) because these reactors are not LWR designs upon |

which Table S–4 impacts were based. Therefore, a full description and detailed analysis was

required for each gas-cooled reactor design. This was provided by SERI in its response to a

request for additional information on September 30, 2004 (SERI 2004f). |

SERI used a sensitivity analysis to show that transportation impacts from advanced LWR

designs would be bounded by the criteria identified in Table S–4 (SERI 2005a). The GEIS, |

Addendum 1 (NRC 1999) was referenced as the basis for exceeding 4 percent uranium-235 |

enrichment and 33,000 MWd/MTU. However, the GEIS, Addendum 1 applies to reactors that |

are listed in the GEIS, Appendix A, which does not address advanced reactors. |

SERI also used a sensitivity analysis to show that transportation impacts from the advanced

gas-cooled reactor designs would be bounded by the criteria identified in Table S–4

(SERI 2005a); however, as discussed previously, this type of analysis does not adequately |

meet the requirements of 10 CFR 51.52. SERI (2005a) identified the major contributors to |

transportation risk to be the number and type of shipment (shipment risk) and the kind of

material being shipped (material risk). Its evaluation of shipment risk showed fewer shipments

of unirradiated fuel, spent fuel, and low-level waste would be required for the advanced

gas-cooled reactors compared to the reference LWR when averaged over 40 years of

operation. Regarding material risk, SERI (2004f) concluded the following:

 C The estimated total spent fuel radioactive inventory and fission product inventory was

less for the gas-cooled reactors when compared to the reference LWR.

 C Actinide inventories would be greater for the gas-cooled reactors (59 to 64 percent |

greater) because of the increased burnup for these types of reactors; however, because |

the GT-MHR was assumed to ship about one-third less spent fuel on a MTU basis, |

SERI (2005a) determined the actinide inventory per shipment would be about one-half |

of that in the reference LWR shipment. The PBMR is assumed to ship the same |

amount of spent fuel in a spent fuel shipping cask as the reference LWR so there is |

about a 60 percent increase in per-shipment actinide inventories from PBMR spent fuel |

shipments relative to the reference LWR. |

 C Gas-cooled reactors would generate fewer kilowatts of decay heat per MTU and fewer

kilowatts of decay heat per truck cask at the time of shipment.
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-24 April 2006

6.2.1 Transportation of Unirradiated Fuel

The staff performed an independent review of the environmental impacts of transporting

unirradiated (fresh) fuel to the Grand Gulf ESP site. Environmental impacts of normal operating

conditions and transportation accidents are discussed in this section. Appendix H provides the

details of the analysis.

6.2.1.1 Normal Conditions

Normal conditions, sometimes referred to as “incident-free” transportation, are transportation

activities in which shipments reach their destination without releasing any radioactive cargo to

the environment. Impacts from these shipments would be from the low levels of radiation that

penetrate the unirradiated fuel shipping casks.

Truck Shipments

Table 6-4 provides an estimate of the number of truck shipments of unirradiated fuel for each

advanced reactor design compared to those of the reference 1100-MW(e) reactor specified in

WASH-1238 (AEC 1972). Estimates are normalized for an equivalent 1100-MW(e) electric

generating capacity. The basis for the shipment estimates can be found in Appendix H of this

| EIS. Only the ACR-700, PBMR, and GT-MHR reactor designs would exceed the number of

truck shipments of unirradiated fuel estimated for the reference LWR in WASH-1238

| (AEC 1972). The largest number of shipments, in excess of 700 shipments over 40 years, is

| for the GT-MHR. However, the combined annual shipments of unirradiated fuel, spent fuel, and

| radioactive waste equates to far less than the one truck shipment per day specified in

| Table S–4 of 10 CFR 51.52 for all reactor types. 

Shipping Mode and Weight Limits

| In 10 CFR 51.52(a), a condition is identified that states all unirradiated fuel be shipped to the

reactor by truck. In information provided by SERI, SERI specifies that unirradiated fuel will be

shipped to the reactor site by truck for all reactor designs that it references (INEEL 2003). In

| addition, 10 CFR 51.52(c) includes a condition that the truck shipments not exceed 33,100 kg

(73,000 lbs), as governed by Federal or State gross vehicle weight restrictions. All the

advanced reactor designs would meet this weight restriction for unirradiated fuel (INEEL 2003).
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-25 NUREG-1817

Table 6-4. Numbers of Truck Shipments of Unirradiated Fuel for Each Advanced Reactor Type

Reactor Type

Num ber of Shipments

per Reactor Unit Unit Electric

Generation,

MW (e)(c)

Capacity

Factor(c)

|

Normalized, |

Shipments per

1100 MW(e)(d,e)

Initial

Core(a)

Annual

Reload Total(b)

Reference LWR

(WASH-1238)

18 6 252 1100 0.8 252

ABW R/ESBWR 30 6.1 267 1500 0.95 165 |

Surrogate AP1000 14 3.8 161 1150 0.95 130 |

ACR-700 30 15.4 628 1462(f) 0.9 420 |

IRIS 34 4.3 201 1005(g) 0.96 184 |

GT-MHR 51 20 831 1140(h) 0.88 729 |

PBMR 44 20 824 1320(i) 0.95 579 |

(a) Shipments of the initial core have been rounded up to the next highest whole number. |

(b) Total shipments unirradiated fuel over a 40-year plant lifetime (i.e., initial core load plus 39 years of average |

annual reload quantities).

(c) Unit capacities and capacity factors were taken from INEEL (2003).

(d) Normalized to net electric output for WASH-1238 reference LWR; i.e., 1100-MW(e) plant at 80 percent or net |

electrical output of 880 MW(e).

(e) Ranges of capacities are given in INEEL (2003) for these unirradiated fuel shipments. The unirradiated fuel |

shipment data for these reactors were derived using the upper limit of the ranges. |

(f) The ACR-700 unit includes two reactors at 731 MW(e) per reactor. |

(g) The IRIS unit includes three reactors at 335 MW(e) per reactor. |

(h) The GT-MHR unit includes four reactors at 285 MW(e) per reactor. |

(i) The PBMR unit includes eight reactors at 165 MW(e) per reactor. |

WASH-1238 = ACE 1972 |

Note: The reference LWR shipment values have all been normalized to 880-MW(e) net electrical generation. |

Radiological Doses to Transport Workers and the Public

10 CFR 51.52, Table S–4, includes conditions related to radiological dose to transport workers

and members of the public along transport routes. These doses are a function of many 

variables, including the radiation dose rate emitted from the unirradiated fuel shipments, the

number of exposed individuals and their locations relative to the shipment, the time in transit

(including travel and stop times), and the number of shipments to which the individuals are

exposed. For this EIS, the radiological dose impacts of the transportation of unirradiated fuel |

were calculated for the worker and the public using the RADTRAN 5 computer code

(Neuhauser et al. 2003). Details of the calculations are found in Appendix H.

Table 6-5 presents the radiological impacts to workers, public onlookers (persons at stops and |

sharing the road), and members of the public along the route (i.e., residents within 800 m |

(0.5 mi) of the highway) for the advanced reactor designs. The cumulative annual dose |

estimates in Table 6-5 were normalized to 1100 MW(e). The NRC staff performed an |
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-26 April 2006

independent review and determined that all dose estimates are bounded by the Table S–4

conditions of 0.04 person-Sv/yr (4 person-rem/yr) to transportation workers, 0.03 person-Sv/yr

(3 person-rem/yr) to onlookers, and 0.03 person-Sv/yr (3 person-rem/yr) to members of the

public along the route.

Table 6-5. Radiological Impacts of Transporting Unirradiated Fuel to Advanced Reactor Sites

Plant Type

Normalized

Average

Annual

Shipments

Cumulative Annual Dose, person-Sv/yr per

1100 MW(e)(a)

Workers

Public -

Onlookers

Public - Along

Route

| Reference LWR

(WASH-1238)

6.3 1.1 x 10-4 4.2 x 10-4 1.0 x 10-5 |

| ABW R/ESBWR 4.1 7.1 x 10-5 2.7 x 10-4 6.6 x 10-6 |

| Surrogate AP1000 3.3 5.6 x 10-5 2.2 x 10-4 5.2 x 10-6

ACR-700 10.5 1.8 x 10-4 7.0 x 10-4 1.7 x 10-5

IRIS 4.6 7.9 x 10-5 3.1 x 10-4 7.4 x 10-6

GT-MHR 18.2 3.1 x 10-4 1.2 x 10-3 2.9 x 10-5

PBMR 14.5 2.5 x 10-4 9.6 x 10-4 2.3 x 10-5

10 CFR 51.52, Table S–4

Condition

<1 per day 4.0 x 10-2 3.0 x 10-2 3.0 x 10-2

| (a) Multiply person-sievert (Sv)/yr by 100 to obtain doses in person-rem/yr.

WASH-1238 = AEC 1972

Although radiation may cause cancers at high doses and high dose rates, currently there are no

| data that unequivocally establish the occurrence of cancer following exposure to low doses,

below about 100 mSv (10,000 mrem). However, radiation protection experts conservatively

assume that any amount of radiation may pose some risk of causing cancer or a severe

hereditary effect and that the risk is higher for higher radiation exposures. Therefore, a linear,

| no-threshold dose response model is used to describe the relationship between radiation dose

| and detriments such as cancer induction. A recent report (National Research Council 2006),

| the BEIR VII report, supports the linear, no-threshold dose response theory. Simply put, this

| theory states that any increase in dose, no matter how small, results in an incremental increase

| in health risk. This theory is accepted by the NRC as a conservative model for estimating

health risks from radiation exposure, recognizing that the model probably overestimates those

risks.

| Based on this model, the staff estimates the risk to the public from radiation exposure using the

| nominal probability coefficient for total detriment (730 fatal cancers, nonfatal cancers, and
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-27 NUREG-1817

severe hereditary effects per 10,000 person-Sv (1,000,000 person-rem)) from International |

Commission on Radiological Protection Publication 60 (ICRP 1991). All the public doses |

presented in Table 6-5 are less than or equal to 0.0012 person-Sv/yr (0.12 person-rem/yr); |

therefore, the total detriment estimates associated with these doses would all be less than |

1 x 10-4 fatal cancers, nonfatal cancers, and severe hereditary effects per year. These risks are |

very small compared to the fatal cancers, nonfatal cancers, and severe hereditary effects that |

would be expected to occur annually in the same population from exposure to natural sources |

of radiation.

Maximally Exposed Individuals Under Normal Transport Conditions |

A scenario-based analysis was conducted to develop estimates of incident-free radiation doses |

to maximally exposed individuals (MEI). The analysis is based on information in DOE (2002) |

and incorporates information about exposure times, dose rates, and the number of times an |

individual may be exposed to an offsite shipment. Adjustments were made where necessary to |

reflect the fuel and waste shipments addressed in this EIS. In all cases, it was assumed that |

the dose rate emitted from the shipping containers is 0.1 mSv/hr (10 mrem/hr) at 2 m (6.6 ft) |

from the side of the transport vehicle, the maximum dose rate allowed by U.S. Department of |

Transportation regulations, even though unirradiated fuel and radioactive waste will have much |

lower dose rates than the regulations allow. An MEI is a person who may receive the highest |

radiation dose from a shipment to and/or from the advanced reactor site. The analysis is |

described below.

Truck crew member. Truck crew members would receive the highest radiation doses during |

incident-free transport because of their proximity to the loaded shipping container for an |

extended period of time. The analysis assumed that crew member doses are limited to |

0.02 Sv (2 rem) per year, which is the DOE administrative control level (DOE 2002). This limit |

is anticipated to apply to spent nuclear fuel shipments to a disposal facility, as DOE will take |

title to the spent fuel at the reactor site. Spent nuclear fuel represents the bulk of the fuel and |

waste shipments to/from advanced reactor sites, and those with the highest radiation dose |

rates, so crew doses from unirradiated fuel and radioactive waste shipments will be lower |

than the spent nuclear fuel shipments. The NRC limit for occupational exposures is |

0.05 Sv/yr (5 rem/yr). |

Inspectors. Radioactive shipments are inspected by Federal or state vehicle inspectors, for |

example, at state ports of entry. DOE (2002) assumed that inspectors would be exposed for |

1 hour at a distance of 1 m (3.3 ft) from the shipping containers. The dose rate at 1 m (3.3 ft) is |

about 0.14 mSv/hr (14 mrem/hr), so the dose per shipment is about 0.14 Sv (14 mrem). This is |

independent of the location of the advanced reactor site. Based on this conservative value, the |

annual doses to vehicle inspectors were calculated to be in the range of 9 to 18 mSv/yr (900 to |

1800 mrem/yr), assuming the same person inspects all shipments of fuel and waste to and from |
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-28 April 2006

| the advanced reactor sites. The high end of the range is the ACR-700 and the low end is the

| surrogate AP1000. All of the values are less than the 20 mSv/yr (2000 mrem/yr) administrative

| control level on individual doses.

| Resident. The analysis assumed that a resident lives 30 m (100 ft) from the point where a

| shipment would pass and would be exposed to all shipments along a particular route.

| Exposures to residents on a per-shipment basis were extracted from RADTRAN 5 output files. 

| These dose estimates are based on an individual located 30 m (100 ft) from the shipments that

| are traveling 24 km/hr (15 mph). The potential radiation doses to maximally-exposed residents,

| which are independent of the location of the advanced reactor site, ranged from about

| 0.00027 mSv/yr (0.027 mrem/yr) for the surrogate AP1000 to 0.00055 mSv/yr (0.055 mrem/yr)

| for the ACR-700.

| Individual stuck in traffic. This scenario addresses potential traffic interruptions that could lead

| to a person being exposed to a loaded shipment for one hour at a distance of 1.2 m (4 ft). The

| analysis assumed this exposure scenario would occur only one time to any individual. The

| dose to the MEI was calculated in DOE (2002) to be 0.016 mSv (1.6 mrem). 

| Person at a truck service station. This scenario estimates doses to an employee at a service

| station where all truck shipments to/from the advanced reactors would stop. DOE (2002)

| assumed this person is exposed for 49 minutes at a distance of 16 m (52 ft) from the loaded

| shipping container. This results in a dose of about 0.0007 mSv/shipment (0.07 mrem/shipment)

| and an annual dose in the range from 0.044 mSv (4.4 mrem) for the surrogate AP1000 to

0.09 mSv/yr (9 mrem/yr) for the ACR-700.

6.2.1.2 Accidents

| Accident risks are a combination of accident frequency and consequence. Accident

| frequencies for transportation of fuel to and from future reactors are expected to be lower than

those used in the analysis in WASH-1238 (AEC 1972), which forms the basis for Table S–4 of

10 CFR 51.52, because of improvements in highway safety and security and an expected

| decrease in traffic accident, injury, and fatality rates. There is no significant difference in

| consequences of accidents severe enough to result in a release of unirradiated fuel particles to

| the environment between advanced LWRs and current-generation LWRs because the fuel

| form, cladding, and packaging are similar to those analyzed in WASH-1238. Consequently, the

| impacts of accidents during transport of unirradiated fuel for advanced LWRs to the Grand

| Gulf ESP site are expected to be smaller than the impacts listed in Table S–4 for current

| generation LWRs.

With respect to the advanced gas-cooled reactors, accident rates (accidents per unit distance)

and associated accident frequencies (accidents per year) would be expected to follow the same
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-29 NUREG-1817

trends as for LWRs (i.e., overall reduction relative to the accident rates used in the WASH-1238

analysis). The consequences of accidents involving gas-cooled reactor unirradiated fuel,

however, are more uncertain. The staff assumed that the gas-cooled reactor unirradiated fuel

shipments would have the same abilities as LWR unirradiated fuel to maintain functional

integrity following a traffic accident. This assumption is considered to be conservative because

gas-cooled reactor fuel operates at significantly higher temperatures, and thus maintains

integrity under more severe thermal conditions than LWR fuel. Detailed information about the

behavior of the gas-cooled reactor fuel under impact conditions was not available. However,

packaging systems for unirradiated gas-cooled reactor fuel will need to meet the same |

performance requirements as unirradiated LWR fuel packages, including fissile material |

controls to prevent criticality during normal and accident conditions. Consequently, it is |

expected that packaging systems for unirradiated gas-cooled reactor fuels would provide

equivalent protection to those designed for unirradiated LWR fuels. In addition, the fuel forms |

for the gas-cooled reactors are similar to those for LWRs (i.e., UO2 for the PBMR and uranium

oxycarbide for the GT-MHR versus UO2 for LWRs) thus, the inherent failure resistance provided |

by unirradiated gas-cooled reactor fuels should be similar to that provided by LWRs. Based on |

the assumption that unirradiated gas-cooled and LWR fuels and associated packaging systems

would provide similar resistance to various environmental conditions, the staff estimates that |

the impacts of accidents involving unirradiated gas-cooled reactor fuel likely would not be |

significantly different from impacts involving unirradiated LWR fuel and would be within the |

impacts listed in Table S–4 for current-generation LWRs. However, these impacts are not |

considered to be resolved, and would need to be assessed at the CP or COL stage when |

specific information is available regarding other-than-LWR fuel performance, if the applicant |

references such designs.

6.2.2 Transportation of Spent Fuel

The staff performed an independent review of the environmental impacts of transporting spent

fuel from the proposed new nuclear unit or units at the Grand Gulf ESP site to a spent fuel |

disposal repository. The Yucca Mountain, Nevada, site is a possible location for a geologic

repository. The staff considers that an estimate of the impacts of the transportation of spent

fuel to a possible repository at Yucca Mountain, Nevada to be a reasonable bounding estimate |

of the transportation impacts to a storage or disposal facility because of the distances involved

and the representativeness of the distribution of members of the public in urban, suburban, and |

rural areas (i.e., population distributions) along the shipping routes. Environmental impacts of |

normal operating conditions and transportation accidents are discussed in this section.

This analysis is based on shipment of spent fuel by legal-weight trucks in casks with

characteristics similar to casks currently available (i.e., massive, heavily shielded, cylindrical

metal pressure vessels). Each shipment is assumed to consist of a single shipping cask loaded

on a modified trailer. These assumptions are consistent with assumptions made in the
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-30 April 2006

| evaluation of the environmental impacts of transportation of spent fuel in Addendum 1 to the

| GEIS (NRC 1999). These assumptions are conservative because the alternative assumptions

involve rail transportation or heavy-haul trucks, which would reduce the overall number of spent

fuel shipments (NRC 1999).

Environmental impacts of transportation of spent fuel were calculated using the RADTRAN 5

computer code (Neuhauser et al. 2003). Routing and population data used in the RADTRAN 5

for truck shipments were obtained from the TRAGIS routing code (Johnson and

| Michelhaugh 2000). The population data in the TRAGIS code are based on the 2000 U.S.

census.

| The staff’s evaluation reviewed the impacts of spent fuel shipments originating from the Grand

| Gulf ESP site and the alternative sites: James A. FitzPatrick Nuclear Power Plant and Pilgrim

Nuclear Station. Another alternative site, River Bend Station, was considered by SERI in its

environmental report, but was not evaluated by the staff because the route characteristics of

distance and population would not be sufficiently different to produce results different from the

| Grand Gulf ESP site. Appendix H provides the details of the analysis.

6.2.2.1 Normal Conditions

Normal conditions, sometimes referred to as “incident-free” transportation, are transportation

| activities in which shipments reach their destination without an accident occurring en route.

Impacts from these shipments would be from the low levels of radiation that penetrate the

| heavily shielded spent fuel shipping cask. Radiation exposure would occur to (1) persons

| residing along the transportation corridors between the Grand Gulf ESP site and the proposed

| repository; (2) persons in vehicles traveling on the same route as the spent fuel shipment;

| (3) persons at vehicle stops for vehicle inspections, refueling, and rest; and (4) transportation

| crew workers.

Shipping casks have not been designed for the advanced reactor designs. Information in

INEEL (2003) indicated that advanced LWR fuel designs would not be significantly different

from existing LWR designs; therefore, the characteristics of current shipping cask designs were

used for the analysis for advanced LWR designs. No information is available on spent fuel

| shipping cask designs for the gas-cooled reactors. For purposes of this Chapter 6 analysis,

| their design was assumed to be the same as those used for the existing LWRs. Spent fuel

| shipping cask designs for gas-cooled reactors have not been defined and, therefore, impacts

| are not resolved. Impacts would be evaluated at the CP or COL stage if the applicant

references such designs.

Radiation doses are a function of many parameters, including vehicle speed, traffic count, dose

rate at 1 m from the vehicle, packaging dimensions, number of persons in the truck crew, stop
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-31 NUREG-1817

time, and population density at stops. For a listing of the values for these and other

parameters, refer to Appendix H. Table 6-6 presents radiation dose estimates to the transport

workers and the public for the primary and alternative ESP sites. Doses are presented on a

per-shipment basis. The per-shipment dose estimates are independent of reactor technology

because they were calculated based on an assumed external radiation dose rate emitted from

the cask, which was fixed at the regulatory maximum limit for the advanced reactor designs

(i.e., 0.1 mSv/hr (10 mrem/hr) at 2 m).

Table 6-6. Routine (Incident-Free) Radiation Doses to Transport Workers and the Public |

from Shipping Spent Fuel from Potential Early Site Permit Sites to a Spent Fuel |

Disposal Facility |

Reactor Site |

Population Dose, person-Sv/shipment(a) |

Crew Onlookers Along Route |

Grand Gulf(b) 8.7 x 10 |

-4 2.8 x 10-3 7.0 x 10-5 |

FitzPatrick 9.8 x 10 |

-4 3.5 x 10-3 9.5 x 10-5 |

Pilgrim 1.1 x 10 |

-3 3.9 x 10-3 1.2 x 10-4 |

(a) Multiply person-sievert (Sv)/yr by 100 to obtain doses in person-rem/yr. |

(b) Doses for the River Bend alternative site can be assumed to be bounded by the values for the proposed |

Grand Gulf ESP site because differences in route characteristics are minimal. |

Population dose estimates per reactor year are presented in Table 6-7 for specific advanced |

reactor designs. Population doses were calculated by multiplying the number of spent fuel

shipments per year for each advanced reactor design times the dose per shipment from

Table 6-6. Population doses were normalized to the reference LWR design in WASH-1238

(880 net MW(e)) (AEC 1972). This corresponds to an 1100-MW(e) LWR operating at |

80 percent capacity. Appendix H provides the basis upon which the number of spent fuel

shipments was derived for each advanced reactor design.

The bounding cumulative doses to the exposed population given in Table S–4

(10 CFR 51.51(c)) are |

 C 0.04 person-Sv (4 person-rem) per reactor-year to transport workers

 C 0.03 person-Sv (3 person-rem) per reactor-year to general public (onlookers) and |

members of the public along the route. |

Population doses to the crew and the onlookers for all the reactor types, including the reference

reactor designated in Table 6-7, exceed Table S–4 values. Two key reasons for the higher |

population doses relative to Table S–4 are the higher number of spent fuel shipments estimated

for some of the reactor technologies and the longer shipping distances assumed for the

analyses (i.e., to a possible repository in Nevada) than were used in WASH-1238. WASH-1238
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-32 April 2006

used a “typical” distance for a spent fuel shipment of 1600 km (1000 mi), whereas the shipping

distances used in this assessment ranged from about 3000 km (1800 mi) to 4700 km (2900 mi). 

The higher numbers of shipments are based on spent fuel shipping casks designed to transport

| shorter-cooled fuel (150 days out of the reactor). It was assumed in this Chapter 6 analysis that

the shipping cask capacities are 0.5 MTU/shipment, roughly equivalent to one PWR or two

BWR spent fuel assemblies per shipment.

Newer shipping cask designs are based on longer-cooled spent fuel (5 years out of reactor)

and have larger capacities than those used in this assessment. DOE (2002) spent fuel shipping

cask capacities were approximately 1.8 MTU/shipment, or up to four PWR or nine BWR fuel

| assemblies per shipment. Use of the newer shipping cask designs would reduce the number of

spent fuel shipments and the associated environmental impacts. On balance, if the population

doses are adjusted for the shipping distance and shipping cask capacity, the routine population

doses from spent fuel shipments from all reactor types and all sites fall within Table S–4

requirements.

Other conservative assumptions in the staff’s calculation include:

| C Use of the regulatory maximum dose rate (0.1 mSv/hr or 10 mrem/hr at 2 m) in the

RADTRAN 5 calculations. The shipping casks assumed in the EIS prepared in support

of the application for a geologic repository at the proposed Yucca Mountain site

(DOE 2002) were designed to transport spent fuel that has cooled for five years. In

reality, most spent fuel will have cooled for much longer than five years before it is

shipped to a possible geologic repository. Sprung et al. (2000) developed a probabilistic

distribution of dose rates based on fuel cooling times that indicates that approximately

three-fourths of the spent fuel to be transported to a possible geologic repository will

have dose rates less than half of the regulatory limit. Consequently, the estimated

population doses in Table 6-7 could be divided in half if more realistic dose rate

projections are used.

 C Use of 30 minutes as the average time at a truck stop in the calculations. Many stops

| made for actual spent fuel shipments are short-duration stops (e.g., 10 minutes) for brief

| visual inspections of the cargo (e.g., checking the cask tie-downs). These stops

typically occur in minimally populated areas, such as an overpass or freeway ramp in an

unpopulated area. Furthermore, empirical data provided in Griego et al. (1996) indicate

that a 30-minute stop is toward the high end of the stop-time distribution. Average stop

times observed by Griego et al. (1996) are on the order of 18 minutes. Based on these

observations, it was concluded that the stop model assumptions used in this study

overestimate public doses at stops by at least a factor of two. Consequently, the doses

to onlookers given in Table 6-7 could be reduced by a factor of two to reflect more

realistic truck shipping conditions.
April 2006 6-33 NUREG-1817 Table 6-7. Routine (Incident-Free) Population Doses from Spent Fuel Transportation, Normalized to Refe

Water Reactor Reactor Type 

Reference LWR (WASH-1238) ABWR/ESBWR Surrogate AP1000 AC

No Shipments per Year 60 41 40 

Environmental Effects, person-Sv(a) per reference reactor year Reactor Site Crew Onlookers Along Route Crew Onlookers Along Route Crew Onlookers Along Route Crew Onlo

Grand Gulf 5.2 x 10-2 1.7 x 10-1 4.2 x 10-3 3.5 x 10-2 1.2 x 10-1 2.8 x 10-3 3.4 x 10-2 1.1 x 10-1 2.7 x 10-3 7.8 x 10-2 2.5 x 10FitzPatrick 5.9 x 10-2 2.1 x 10-1 5.7 x 10-3 4.0 x 10-2 1.4 x 10-1 3.9 x 10-3 3.9 x 10-2 1.4 x 10-1 3.8 x 10-3 8.8 x 10-2 3.1 x 10Pilgrim 6.5 x 10 

-2 2.3 x 10-1 7.0 x 10-3 4.4 x 10-2 1.6 x 10-1 4.8 x 10-3 4.3 x 10-2 1.5 x 10-1 4.6 x 10-3 9.8 x 10-2 3.5 x 10

Reactor Type IRIS GT-MHR PBMR No Shipments per Year 35 34 12 Environmental Effects, person-Sv(a) per reference reactor year Reactor Site Crew Onlookers Along Route Crew Onlookers Along Route Crew Onlookers Along Route Grand Gulf(b) 3.0 x 10 

-2 9.8 x 10-2 2.4 x 10-3 2.9 x 10-2 9.4 x 10-2 2.3 x 10-3 9.7 x 10-3 3.2 x 10-2 7.8 x 10-4 FitzPatrick 3.4 x 10-2 1.2 x 10-1 3.3 x 10-3 3.3 x 10-2 1.2 x 10-1 3.2 x 10-3 1.1 x 10-2 3.9 x 10-2 1.1 x 10-3 Pilgrim 3.8 x 10 -2 1.3 x 10-1 4.0 x 10-3 3.6 x 10-2 1.3 x 10-1 3.9 x 10-3 1.2 x 10-2 4.3 x 10-2 1.3 x 10-3 (a) Multiply person-sievert (Sv)/yr by 100 to obtain dose in mrem/yr. 

(b) Doses for the River Bend alternative site can be assumed to be bounded by the values for the proposed Grand Gulf ESP site 

because differences in route characteristics are minimal. 

WASH-1238 = AEC 1972
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-34 April 2006

SERI performed its own RADTRAN 5 calculations looking at the impact of “incident-free”

transport of spent fuel to a spent fuel disposal facility. The assumed transport of spent fuel

originated from the Maine Yankee Nuclear Plant (a distance further than the Grand Gulf ESP

site) and terminated at a disposal facility assumed to be at Yucca Mountain, Nevada. Dose

estimates per shipment were similar to those calculated by the staff.

Although radiation may cause cancers at high doses and high dose rates, currently there are no

| data that unequivocally establish the occurrence of cancer following exposure to low doses,

| below about 100 mSv (10,000 mrem), and at low dose rates. However, radiation protection

experts conservatively assume that any amount of radiation may pose some risk of causing

cancer or a severe hereditary effect and that the risk is higher for higher radiation exposures. 

| Therefore, a linear, no-threshold dose response model is used to describe the relationship

| between radiation dose and detriments such as cancer induction. A recent report (National

| Research Council 2006), the BEIR VII Report, supports the linear, no threshold dose response

| theory. Simply put, this theory states that any increase in dose, no matter how small, results in

an incremental increase in health risk. This theory is accepted by the NRC as a conservative

model for estimating health risks from radiation exposure, recognizing that the model probably

overestimates those risks.

| Based on this model, the staff estimated the risk to the public from radiation exposure using the

nominal probability coefficient for total detriment (730 fatal cancers, nonfatal cancers, and

severe hereditary effects per 10,000 person-Sv (1,000,000 person-rem)) from International

| Commission on Radiological Protection Publication 60 (ICRP 1991). All the population doses

presented in Table 6-7 are less than one person-Sv/yr (100 person-rem/yr); therefore, the total

detriment estimates associated with these population doses would all be less than 0.1 fatal

cancers, nonfatal cancers, and severe hereditary effects per year. These risks are very small

compared to the fatal cancers, nonfatal cancers, and severe hereditary effects that would be

| expected to occur annually in the same population from exposure to natural sources

of radiation.

| Dose estimates to the MEI from transport of unirradiated fuel, spent fuel, and wastes under

| normal conditions are presented in Section 6.2.1.1.

6.2.2.2 Accidents

As discussed previously, the staff used the RADTRAN 5 computer code to estimate impacts of

transportation accidents involving spent fuel shipments. RADTRAN 5 considers a spectrum of

potential transportation accidents, ranging from those with high frequencies and low

consequences (e.g., “fender benders”) to those with low frequencies and high consequences

(i.e., accidents in which the shipping container is exposed to severe mechanical and thermal

conditions). Details of the analysis are discussed in Appendix H.
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-35 NUREG-1817

Radionuclide inventories are important parameters in the calculation of accident risks. The

radionuclide inventories used in this Chapter 6 analysis are from Early Site Permit |

Environmental Report Sections and Supporting Documentation (INEEL 2003). This report

included hundreds of radionuclides for each advanced reactor type. A screening analysis was

conducted to select the dominant contributors to accident risks to simplify the RADTRAN 5

calculations. The screening identified the radionuclides that would contribute more than

99.999 percent of the dose from inhalation of radionuclides released following a transportation

accident. The dominant radionuclides are similar regardless of the fuel type (i.e., advanced

LWR fuel or gas-cooled reactor fuel). Spent fuel inventories used in the staff analysis are

presented in Table 6-8. Note that the list of radionuclides provided in the table includes all of |

the radionuclides that were included in the analysis conducted by Sprung et al. (2000), which

validates the screening process used in this EIS. Also note that the INEEL (2003) analysis |

relied upon by SERI in its application did not provide radionuclide source terms for radioactive |

material deposited on the external surfaces of LWR spent fuel rods (commonly called “crud”). |

In addition, data on activation products was provided for only the advanced BWR. The |

advanced BWR spent fuel transportation risks were calculated assuming the entire cobalt-60 |

inventory is in the form of crud. This is very conservative as the source term because it is about |

two orders of magnitude greater than that given in Sprung et al. (2000). Because crud is |

deposited from corrosion products generated elsewhere in the reactor cooling system and the |

complete reactor design and operating parameters are uncertain, the quantities and |

characteristics of crud deposited on advanced reactor spent fuel are unknown at this time. |

Consequently, the impacts of crud and activation products on spent fuel transportation accident |

risks are not resolved and would need to be examined at the CP or COL stage. No |

radionuclide inventory data were presented in INEEL (2003) for the ACR-700 and IRIS

advanced reactors. Because transportation accident risks were not quantified for these reactor |

types, these accident risks are not resolved and would need to be assessed at the CP or COL |

stage if the applicant references either of these designs. |

Robust shipping casks are used to transport spent fuel because of the radiation shielding and |

accident resistance required by 10 CFR Part 71. Spent fuel shipping casks must be certified

Type B packaging systems, which means the casks must withstand a series of severe |

postulated accident conditions with essentially no loss of containment or shielding capability. |

These casks are also designed with fissile material controls to ensure the spent fuel remains |

subcritical under normal and accident conditions. According to Sprung et al. (2000), the |

likelihood of encountering accident conditions that would lead to shipping cask failure is less |

than 0.01 percent (i.e., more than 99.99 percent of all accidents would result in no release of

radioactive material from the shipping cask). The staff assumed that shipping casks for

advanced reactor spent fuels will provide equivalent mechanical and thermal protection of the

spent fuel cargo. 
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-36 April 2006

Table 6-8. Radionuclide Inventories Used in Transportation Accident Risk Calculations

for Each Advanced Reactor Type, Bq/MTU

(a)

Radionuclide

ABWR and ESBWR

Inventory

Surrogate AP1000

Inventory

GT-MHR

Inventory

PBMR

Inventory

Am-241 4.96 x 1013 2.69 x 1013 8.18 x 1013 7.55 x 1013

Am-242m 1.24 x 1012 4.85 x 1011 5.03 x 1011 8.51 x 1011

Am-243 1.20 x 1012 1.24 x 1012 5.14 x 1011 4.77 x 1012

Ce-144 4.22 x 1014 3.28 x 1014 2.15 x 1015 1.19 x 1015

Cm-242 2.04 x 1012 1.05 x 1012 1.51 x 1012 2.78 x 1012

Cm-243 1.37 x 1012 1.14 x 1012 2.02 x 1011 1.96 x 1012

Cm-244 1.80 x 1014 2.87 x 1014 2.83 x 1013 5.48 x 1014

Cm-245 2.43 x 1010 4.48 x 1010 1.65 x 1085.29 x 1010

Co-60 1.01 x 1014 (b) (b) (b)

Cs-134 1.78 x 1015 1.78 x 1015 2.21 x 1015 4.03 x 1015

Cs-137 4.59 x 1015 3.44 x 1015 1.08 x 1016 1.41 x 1016

Eu-154 3.81 x 1014 3.38 x 1014 3.23 x 1014 3.74 x 1014

Eu-155 1.93 x 1014 1.71 x 1014 8.77 x 1013 1.08 x 1014

Pm-147 1.25 x 1015 6.51 x 1014 6.92 x 1015 5.07 x 1015

Pu-238 2.27 x 1014 2.25 x 1014 1.17 x 1014 4.55 x 1014

Pu-239 1.43 x 1013 9.44 x 1012 2.25 x 1013 1.11 x 1013

Pu-240 2.28 x 1013 2.01 x 1013 3.96 x 1013 3.32 x 1013

Pu-241 4.51 x 1015 2.58 x 1015 8.33 x 1015 7.18 x 1015

Pu-242 8.29 x 1010 6.73 x 1010 1.56 x 1011 4.51 x 1011

Ru-106 6.07 x 1014 5.74 x 1014 1.48 x 1015 1.68 x 1015

Sb-125 1.99 x 1014 1.42 x 1014 2.21 x 1014 2.51 x 10 14

Sr-90 3.27 x 1015 2.29 x 1015 8.95 x 1015 1.08 x 1016

Y-90 3.27 x 1015 2.29 x 1015 8.95 x 1015 1.08 x 1016

(a) Divide bequerel (Bq)/metric tons uranium (Bq/MTU) by 3.7 x 1010 | to obtain curies/MTU.

| (b) Cobalt-60 is an activation product. Only the ABWR and ESBWR reactor types identified in INEEL (2003)

included inventory data for activation products.

The RADTRAN 5 accident risk calculations were performed using unit radionuclide inventories

(Bq/MTU) for the spent fuel shipments from the various reactor types. The resulting risk

estimates were then multiplied by assumed annual spent fuel shipments (MTU/yr) to derive
Fuel Cycle, Transportation, and Decommissioning

(a) Internal dose from ingestion of contaminated food was not considered as the staff assumed |

evacuation and subsequent interdiction of foodstuffs following a potential transportation accident. |

April 2006 6-37 NUREG-1817

estimates of the annual accident risks associated with spent fuel shipments from each potential

advanced reactor site. As was done for routine exposures, the staff assumed that the numbers |

of shipments of spent fuel per year are equivalent to the annual discharge quantities.

For this assessment, release fractions for current-generation LWR fuels were used to

approximate the impacts from the advanced reactor spent fuel shipments. This assumes that

the fuel materials and containment systems (i.e., cladding, fuel coatings) behave similarly to

current LWR fuel under applied mechanical and thermal conditions. Because of the lack of

experimental data on gas-cooled reactor fuels, it is currently not known if this approach is

bounding. However, gas-cooled reactors operate at much higher temperatures than LWRs;

therefore, high temperature conditions anticipated in transportation accident fires should have

less of an effect on radionuclide releases than they do for LWR fuels. Thus, smaller release

fractions are anticipated for advanced gas-cooled reactor fuels than for LWR fuels subjected to

thermal transients. However, this issue is not resolved because of the lack of information on |

these designs. 

The NRC staff used RADTRAN 5 to calculate the population dose from the radioactive material |

released to the environment and assessed for four of five(a) possible exposure pathways. |

These pathways are:

(1) External dose from exposure to the passing cloud of radioactive material (cloudshine) |

(2) External dose from the radionuclides deposited on the ground by the passing plume

(groundshine). The staff’s analysis included the radiation exposure from this pathway even |

though the area surrounding a potential accidental release would be evacuated and

decontaminated, thus preventing long-term exposures from this pathway. |

(3) Internal dose from inhalation of airborne radioactive contaminants (inhalation) |

(4) Internal dose from resuspension of radioactive materials that were deposited on the ground

(resuspension). The staff’s analysis included the radiation exposures from this pathway |

even though evacuation and decontamination of the area surrounding a potential accidental

release would prevent long-term exposures.

Table 6-9 presents the environmental consequences of transportation accidents when shipping

spent fuel from the Grand Gulf ESP site and alternative sites to the proposed Yucca Mountain |

repository. The shipping distances and population distribution information for the routes were |

the same as those used for the normal “incident-free” conditions (for details, see Appendix H). |
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-38 April 2006

| Table 6-9 presents estimates of population dose (person-Sv/ reactor year) for several of the

advanced reactor designs. These values are normalized to the WASH-1238 reference reactor

(880-MW(e) net electrical generation, 1100-MW(e) reactor operating at 80 percent capacity)

| (AEC 1972). 

Table 6-9. Annual Spent Fuel Transportation Accident Impacts for Advanced

Reactors, Normalized to Reference 1000-MW(e) Light Water Reactor

Net Electrical Generation

MTU/yr

Advanced Reactor Type

| ABWR and

ESBWR

Surrogate

| AP1000 GT-MHR PBMR

20.3 19.7 6.0 5.8

Population Dose, person-Sv/per reference reactor year(a)

Grand Gulf(b) | 4.1 x 10-6 3.7 x 10-7 1.7 x 10-7 2.7 x 10-7 |

FitzPatrick 3.8 x 10-6 3.3 x 10-7 1.5 x 10-7 2.5 x 10-7

Pilgrim 8.1 x 10-6 7.2 x 10-7 3.3 x 10-7 5.4 x 10-7 |

(a) Multiply person-Sv/yr times 100 to obtain person-rem/yr.

(b) Doses for the River Bend alternative site can be assumed to be bounded by the values

for the proposed Grand Gulf ESP site because differences in route characteristics are

minimal.

Although radiation may cause cancers at high doses and high dose rates, currently there are no

data that unequivocally establish the occurrence of cancer following exposure to low doses

| below about 100 mSv (10,000 mrem) and at low dose rates. However, radiation protection

experts conservatively assume that any amount of radiation may pose some risk of causing

cancer or a severe hereditary effect, and that the risk is higher for higher radiation exposures. 

| Therefore, a linear, no-threshold dose response model is used to describe the relationship

| between radiation dose and detriments such as cancer induction. A recent report (National

| Research Council 2006), the BEIR VII report, supports the linear, no-threshold dose response

| theory. Simply put, this theory states that any increase in dose, no matter how small, results in

an incremental increase in health risk. This theory is accepted by the NRC as a conservative

model for estimating health risks from radiation exposure, recognizing that the model probably

overestimates those risks.

Based on this model, the staff estimates the risk to the public from radiation exposure using the

nominal probability coefficient for total detriment – 730 fatal cancers, nonfatal cancers, and

severe hereditary effects per 10,000 person-Sv (1,000,000 person-rem) – from International

| Commission on Radiological Protection Publication 60 (ICRP 1991). All the population doses

presented in Table 6-9 are less than 1.0 x 10-5 person-Sv (1.0 x 10-3 person-rem) per reference

| reactor year; therefore, the total detriment estimates associated with these population doses

would all be less than 1.0 x 10-6 fatal cancers, nonfatal cancers, and severe hereditary effects
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-39 NUREG-1817

per reference reactor year. These risks are quite small compared to the fatal cancers, nonfatal

cancers, and severe hereditary effects that would be expected to occur annually in the same |

population from exposure to natural sources of radiation. |

6.2.2.3 Conclusion

The values determined by this analysis represent the contribution of such effects to the |

environmental costs of licensing the reactor. Because of the conservative approaches and data |

used to calculate doses, actual environmental effects are not likely to exceed those calculated |

in the EIS. Thus, the staff concludes that the overall transportation accident risks associated |

with advanced LWR reactor spent fuel shipments are SMALL and are consistent with the risks |

associated with transportation of spent fuel from current-generation reactors presented in

Table S–4 of 10 CFR 51.52. The fuel performance characteristics, shipping casks, and |

accident risks for other-than-LWR designs are not resolved and would need to be assessed at |

the CP or COL stage if the applicant references such designs. |

6.2.3 Transportation of Radioactive Waste

This section discusses the environmental effects of transporting waste from advanced reactor |

sites. The environmental conditions listed in 10 CFR 51.52(a) that apply to shipments of |

radioactive waste include the following:

 C Radioactive waste (except spent fuel) is packaged in solid form.

 C Radioactive waste (except spent fuel) is shipped from the reactor by truck or rail.

 C The weight limitation is 33,100 kg (73,000 lb) per truck and 90,700 kg (100 tons) per |

cask per railcar.

 C The traffic density limitation is less than one truck shipment per day or three railcars per

month.

In INEEL (2003), it is stated that all the radioactive waste will be transported by truck. SERI |

plans to solidify and package its waste regardless of which advanced reactor technology it |

chooses. In addition, waste from any of the advanced reactor technologies will be subject to |

NRC (10 CFR Part 71) and U.S. Department of Transportation (49 CFR Parts 171, 172, 173, |

and 178) regulations for the shipment of radioactive material. Radioactive waste from any of |

the advanced reactor technologies are expected to be capable of being shipped in compliance

with Federal or State weight restrictions.
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-40 April 2006

Table 6-10 presents estimates of annual waste volumes and annual waste shipment numbers

for the advanced reactor types normalized to the reference 1100-MW(e) LWR defined in

WASH-1238 (AEC 1972). Annual waste volumes and waste shipments for the advanced

reactor technologies were less than the 1100-MW(e) reference reactor that was the basis for

Table S–4 for all designs except the PBMR. 

Table 6-10. Summary of Radioactive Waste Shipments for Advanced Reactors

Reactor Type

INEEL (2003)

Waste

Generation

Information

Annual Waste

Volume, m3/yr

per unit 

Electrical

Output, MW(e)

per unit

Normalized

Rate, m3/1100

MW(e) reactor

(880 MW(e)

net)(a)

Shipments/

1100 MW(e)

(880 MW(e)

net) Electrical

Output(b) |

Reference LWR

(WASH-1238)

100 m3/yr 

per unit

108 1100 108 46 

|

ABWR 100 m3/yr

per unit

100 1500 62 27 

|

ESBWR 100 m3/yr

per unit

100 1500 62 27 

|

Surrogate

| AP1000

55 m3/yr 

per unit

56 1150 45 20

| 

ACR-700 47.5 m3/yr 

per unit

95 1462(c) 64 28

| 

| IRIS 25 m3/yr

per unit

74

(3 units)

1005(d) 67 29 

| GT-MHR 98 m3/yr

(4 unit plant)

98

(4 units)

1140(e) 86 37(g) |

| PBMR 100 drums/yr

per unit

168

(8 units)

1320(f) 118 51(g) |

(a) Capacity factors used to normalize the waste generation rates to an equivalent electrical generation output are

| given in Table 6-3 for each reactor type. All are normalized to 880-MW(e) net electrical output (1100-MW(e)

plant with an 80-percent capacity factor).

(b) The number of shipments per 1100 MW(e) was calculated assuming the WASH-1238 average waste shipment

capacity of 2.34 m3 per shipment (108 m3/yr divided by 46 shipments/yr).

| (c) The ACR-700 unit includes two reactors at 731 MW(e) per reactor.

| (d) The IRIS unit includes three reactors at 335 MW(e) per reactor.

| (e) The GT-MHR unit includes four reactors at 285 MW(e) per reactor.

| (f) The PBMR unit includes eight reactors at 165 MW(e) per reactor.

(g) SERI states in INEEL (2003) that 90 percent of the waste could be shipped on trucks carrying 28 m3 (1000 ft3| )

of waste and the remaining 10 percent in shipments carrying 5.7 m3 (200 ft3) of radioactive waste. This would

| result in five to six shipments per year after normalization to the reference LWR electrical output.

Conversions: 1 m3 = 35.31 ft3, drum volume = 210 liters (0.21 m3| )

| WASH-1238 = AEC 1972
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-41 NUREG-1817

As shown in the table, only the PBMR would be expected to generate a larger volume of

radioactive waste than the reference LWR in WASH-1238 (AEC 1972). However, the GT-MHR |

and PBMR information in INEEL (2003) assumed that the applicant would ship wastes using |

two different packaging systems: one that hauls 28 m3 per shipment (1000 ft3 per shipment) |

and one that hauls 5.7 m3 per shipment (200 ft3 per shipment). Under those conditions, the

number of shipments of radioactive waste per year, normalized to 1100 MW(e) electric

generation capacity, would be about six shipments per year per 1100 MW(e) (880 net MW(e))

for the GT-MHR and seven shipments per year per 1100 MW(e) for the PBMR. These

estimates are well below the reference LWR (46 shipments per year per 1100 MW(e)). |

However, impacts from other than LWR designs are not resolved because of the lack of |

verifiable information. 

The sum of the daily shipments of unirradiated fuel, spent fuel, and radioactive waste is well

below the one truck shipment per day condition given in 10 CFR 51.52, Table S–4 for all reactor |

types. Doubling the shipment estimates to account for empty return shipments of fuel and |

waste is still well below the one-truck-shipment-per-day condition. |

Dose estimates to the maximally-exposed individual from transport of unirradiated fuel, spent

fuel, and waste under normal conditions are presented in Section 6.2.1.1. |

6.2.4 Conclusions

An analysis was conducted of the impacts under normal operating and accident conditions of |

transporting unirradiated fuel to advanced reactor sites and spent fuel and wastes from

advanced reactor sites to disposal facilities. To make comparisons to Table S–4, the

environmental impacts are normalized to a reference reactor year. The reference reactor is an

1100-MW(e) reactor that has an 80-percent capacity factor, for a total electrical output of

880 MW(e) per year. The environmental impacts can be adjusted to calculate impacts per site

by multiplying the normalized impacts by the ratio of the total electric output for the advanced

reactor sites to the electric output of the reference reactor.

Because of the conservative approaches and data used to calculate doses, actual |

environmental effects are not likely to exceed those calculated in the EIS. Thus, the staff |

concludes that the environmental impacts of transportation of fuel and radioactive wastes to

and from advanced LWR designs would be SMALL, and would be consistent with the risks

associated with transportation of fuel and radioactive wastes from current-generation reactors

presented in Table S–4 of 10 CFR 51.52. For gas-cooled designs, the impacts are likely to be

small, but this issue is not resolved because of the lack of verifiable information on these |

designs. At the CP or COL stage, an applicant referencing these designs would need to |

provide the necessary data and the staff would need to validate the assumptions used in this |

transportation analysis. 
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-42 April 2006

| Assumptions that will need validation if a gas-cooled is selected include:

| C Verifying that unirradiated and spent fuel from gas-cooled reactors have the same

| abilities as LWR unirradiated and spent fuel to maintain fuel and cladding integrity

following a traffic accident.

 C Verifying that shipping cask design assumptions (for example, cask capacities) are

equal to or bounded by the assumptions in this analysis.

| C Verifying that unirradiated fuel initial core/refueling requirement, spent fuel generation

rates, and radioactive waste generation rate assumptions are equal to or bounded by

the assumptions in this analysis.

 C Verifying that shipping cask capacities and accident source terms, including spent fuel

inventories, severity fractions, and release fractions, are equal to or bounded by the

assumptions in this analysis.

| Should the ACR-700 or IRIS reactors be chosen for the ESP site, a transportation accident

| analysis will be performed as spent fuel inventories were not available for this analysis. 

6.3 Decommissioning Impacts

At the end of the operating life of a power reactor, the NRC regulations require that the facility

undergo decommissioning. Decommissioning is the removal of a facility safely from service

and the reduction of residual radioactivity to a level that permits termination of the NRC license. 

The regulations governing decommissioning of power reactors are found in 10 CFR 50.75

and 50.82.

Environmental impacts from the activities associated with the decommissioning of any LWR

before or at the end of an initial or renewed license are evaluated in the Generic Environmental

Impact Statement on Decommissioning of Nuclear Facilities, Supplement 1, Regarding the

| Decommissioning of Nuclear Power Reactors, NUREG-0586, (NRC 2002). If an applicant for a

| CP or COL referencing the Grand Gulf ESP applies for a license to operate one or more

| additional units at the Grand Gulf ESP site, there is a requirement to provide a report containing

a certification that financial assurance for radiological decommissioning will be provided. At the

time an application is submitted, the requirements in 10 CFR 50.33, 50.75, and 52.77 (and any

| other applicable requirements) would have to be met.

| At the ESP stage, applicants are not required to submit information regarding the process of

decommissioning, such as the method chosen for decommissioning, the schedule, or any other

| aspect of planning for decommissioning. SERI did not provide this information in its application. 
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-43 NUREG-1817

For the new nuclear unit or units, if LWR designs are chosen or if other-than-LWRs that were |

considered in NUREG-0586, Supplement 1 are chosen, the impacts from decommissioning are

expected to be within the bounds described in NUREG-0586, Supplement 1. In such cases, the

staff expects the impact from decommissioning are likely to be small. However, for whatever |

design that is selected, the impacts from decommissioning are not resolved and would have to |

be assessed at the CP or COL stage. |

6.4 References

10 CFR Part 20. Code of Federal Regulations, Title 10, Energy, Part 20, “Standards for

Protection Against Radiation.”

10 CFR Part 50. Code of Federal Regulations, Title 10, Energy, Part 50, “Domestic Licensing

of Production and Utilization Facilities.”

10 CFR Part 51. Code of Federal Regulations, Title 10, Energy, Part 51, “Environmental

Protection Regulations for Domestic Licensing and Related Regulatory Functions.”

10 CFR Part 52. Code of Federal Regulations, Title 10, Energy, Part 52, “Early Site Permits,

Standard Design Certifications, and Combined Licenses for Nuclear Power Plants.”

10 CFR Part 63. Code of Federal Regulations, Title 10, Energy, Part 63, “Disposal of High- |

Level Radioactive Wastes in a Geologic Repository at Yucca Mountain, Nevada.”

10 CFR Part 71. Code of Federal Regulations, Title 10, Energy, Part 71, “Packaging and

Transportation of Radioactive Material.”

40 CFR Part 190. Code of Federal Regulations, Title 40, Protection of Environment, Part 190,

“Environmental Radiation Protection Standards for Nuclear Power Operations.”

49 CFR Part 171. Code of Federal Regulations, Title 49, Transportation, Part 171, “General |

Information, Regulations, and Definitions.” |

49 CFR Part 172. Code of Federal Regulations, Title 49, Transportation, Part 172, “Hazardous |

Materials Table, Special Provisions, Hazardous Materials Communications, Emergency |

Response Information, and Training Requirements.” |

49 CFR Part 173. Code of Federal Regulations, Title 49, Transportation, Part 173, “Shippers -

General Requirements for Shipments and Packaging.”
Fuel Cycle, Transportation, and Decommissioning

NUREG-1817 6-44 April 2006

49 CFR Part 178. Code of Federal Regulations, Title 49, Transportation, Part 178,

“Specifications for Packaging.”

61 FR 65120. December 10, 1996. “10 CFR Part 20. Resolution of Dual Regulation of

| Airborne Effluents of Radioactive Materials; Clean Air Act.” Federal Register.

| 70 FR 49014. August 22, 2005. “Public Health and Environmental Radiation Protection

| Standards for Yucca Mountain, NV.” Federal Register. 

| 70 FR 53313. September 8, 2005. “Implementation of a Dose Standard after 10,000 Years.” 

| Federal Register. 

Council on Environmental Quality (CEQ). 1976. The Seventh Annual Report of the Council on

Environmental Quality. Executive Office of the President, Administrative Operations Branch,

Washington, D.C.

| Energy Policy Act of 2005. Public Law 109-58.

Griego, N.R., J.D. Smith, and K.S. Neuhauser. 1996. “Investigation of RADTRAN Stop Model

| Input Parameters for Truck Stops.” Conference Proceedings - Waste Management 96,

CONF-960212-44. Tucson, Arizona.

Idaho National Engineering and Environmental Laboratory (INEEL). 2003. Early Site Permit

Environmental Report Sections and Supporting Documentation. Engineering Design File

Number 3747, Idaho Falls, Idaho.

| International Commission on Radiological Protection (ICRP). 1991. 1990 Recommendations of

| the International Commission on Radiological Protection. ICRP Publication 60, Pergamon

| Press, New York.

Johnson, P.E. and R.D. Michelhaugh. 2000. Transportation Routing Analysis Geographic

Information System (WebTRAGIS) User’s Manual. ORNL/TM-2000/86, Oak Ridge National

Laboratory, Oak Ridge, Tennessee. Available at

http://www.ornl.gov/~webworks/cpr/v823/rpt/106749.pdf.

National Academy of Science (NAS). 1995. Technical Bases for Yucca Mountain Standards.

National Academy Press, Washington, D.C.

|

National Cancer Institute (NCI). 1990. Cancer in Populations Living Near Facilities, National

Institutes of Health, Washington, D.C.
Fuel Cycle, Transportation, and Decommissioning

April 2006 6-45 NUREG-1817

National Council on Radiation Protection and Measurements (NCRP). 1987. Ionizing Radiation

Exposure of the Population of the United States. Report No. 93, Bethesda, Maryland.

National Environmental Policy Act of 1969, as amended (NEPA). 42 USC 4321, et seq.

National Research Council. 2006. Health Risks from Exposure to Low Levels of Ionizing |

Radiation: BEIR VII – Phase 2. Committee to Assess Health Risks from Exposure to Low |

Levels of Ionizing Radiation, National Research Council, National Academies Press, |

Washington, D.C. |

Neuhauser, K.S., F.L. Kanipe, and R.F. Weiner. 2003. RADTRAN 5 User Guide. 

SAND2003-2354, Sandia National Laboratories, Albuquerque, New Mexico. Available at

http://infoserve.sandia.gov/sand_doc/2003/032354.pdf. |

Sprung, J.L., D.J. Ammerman, N.L. Breivik, R.J. Dukart, F.L. Kanipe, J.A. Koski, G.S. Mills, |

K.S. Neuhauser, H.D. Radloff, R.F. Weiner, and H.R. Yoshimura. 2000. Reexamination of |

Spent Fuel Shipment Risk Estimates. NUREG/CR-6672, U.S. Nuclear Regulatory Commission, |

Washington, D.C. |

System Energy Resources, Inc. (SERI). 2004f. Letter dated September 30, 2004 from

G.A. Zinke to Nuclear Regulatory Commission, “Response to Supplemental Request for

Additional Information (RAI) Regarding the Environmental Portion of the Early Site Permit

Application by System Energy Resources, Inc. (SERI) for the Grand Gulf ESP Site.” Available

at http://www.nrc.gov/reading-rm/adams.html, Accession No. ML042810132.

System Energy Resources, Inc. (SERI). 2005a. Grand Gulf Site Early Site Permit Application. |

Revision 2, Jackson, Mississippi. Available at http://www.nrc.gov/reading-rm/adams.html,

Accession No. ML052780449. |

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