Appendix G

Authorizations and Consultations

April 2006 G-1 NUREG-1817

Appendix G

Authorizations and Consultations

Table G-1 contains a list of the environmental-related authorizations, permits, and certifications |

potentially required by Federal, State, local, and affected Native American tribal agencies

related to the construction and operation of one or more new nuclear units at the proposed |

Grand Gulf early site permit site.

Table G-1. Federal, State, and Local Authorizations

Agency Authority Requirement Activity Covered

U.S. Nuclear Regulatory

Commission

10 CFR Part 50 Domestic Licensing of

Production and

Utilization Facilities

Construction permit for a new nuclear |

power plant

U.S. Nuclear Regulatory

Commission

10 CFR Part 52 Combined License Issuance of a combined license for |

new nuclear power plants

U.S. Fish and Wildlife

Service and National

Oceanic and Atmospheric

Administration Fisheries

Endangered

Species Act 

16 USC 1536

Consultation Consultation concerning potential

impacts to threatened and |

endangered species |

16 USC 1539 Incidental Take Permit Project related mortality and

modification of critical habitat of |

Federal threatened and endangered

species

U.S. Fish and Wildlife

Service

Migratory Bird

Treaty Act

16 USC 703

Consultation Consultation concerning potential

impacts to migratory birds

|

U.S. Army Corps of

Engineers

Clean Water Act

33 USC 1251

Section 404 Permit Aquatic resource alteration permit

(wetland filling, stream alteration) |

33 CFR Part 209 Dredge and Fill

Discharge Permit

Permit for discharge of dredged spoils |

U.S. Coast Guard 14 USC 81, 83, 85,

633/49

USC 1655(b)

Navigation markers - authorization to |

protect river navigation from hazards

connected with temporary |

construction activities in the river.

Federal Aviation

Administration

Federal Aviation

Act 14 CFR 77.13

Notice Notice to the Federal Aviation |

Administration for structures over |

200 ft in height (e.g., construction |

cranes and cooling towers) |
Appendix G

NUREG-1817 G-2 April 2006

Table G-1. (contd)

Agency Authority Requirement Activity Covered

Mississippi Department of

Environmental Quality

Regulation

APC-S-2

Permit to Construct

Permit to Operate

Permit for the construction and/or

operation of air emissions equipment

Mississippi Department of

Environmental Quality

Regulation

APC-S-5

Permit Mississippi regulations for the

prevention of significant deterioration

of air quality

Mississippi Department of

Environmental Quality

Regulation

APC-S-6

Permit Air operating permit under Title V of

the Federal Clean Air Act

Mississippi Department of

Environmental Quality

Regulation

HW-1

Permit Hazardous waste management

regulations

Mississippi Department of

Environmental Quality

Regulation

LW-1

Permit Surface water and groundwater use

and protection regulations

Mississippi Department of

Environmental Quality

Regulation

SW-2

Permit Non-hazardous solid waste

management regulations and criteria

Mississippi Department of

Environmental Quality

Regulation

UST-2

Permit Underground storage tank regulations

Mississippi Department of

Environmental Quality

Regulation

WPC-1

National Pollutant

Discharge Elimination

System Storm Water

Permit

Waste water regulations for National

Pollutant Discharge Elimination

System permits, water quality based

effluent limitations, and water quality

certification

Mississippi Department of

Environmental Quality

Regulation WPC-2 Water quality criteria for intrastate,

interstate, and coastal waters

Mississippi Department of

Environmental Quality

Regulation WPC-3 Certification Regulations for the certification of

municipal and domestic waste water

facility operators

Mississippi Department of

Wildlife, Fisheries, and

Parks

Natural Heritage

Program

Scientific Collection

Permit

Ecological monitoring programs

Mississippi Public Service

Commission

MS Code of 1972 

SEC. 77-3-11

Certificate of Public

Convenience and

Necessity

Certificate that the present and future

public convenience and necessity

require or will require the operation of

such equipment for facility

Louisiana Department of

Wildlife and Fisheries

Natural Heritage

Program

Scientific Collection

Permit

Ecological monitoring programs
Appendix H

Data and Information to Support Specific Analyses

April 2006 H-1 NUREG-1817

Appendix H

Data and Information to Support Specific Analyses

The data and information used by the U.S. Nuclear Regulatory Commission (NRC) staff to |

support specific analyses in the course of evaluating the proposed Grand Gulf early site permit

(ESP) site and included in this appendix are:

 C Section H.1 - Support Information for Projected Populations (see Chapter 2 for

discussion)

 C Section H.2 - Environmental Impacts of Transportation, which discusses the effects of

both incident-free transportation, transportation accidents, and the environmental effects |

of radioactive waste shipments (see Section 6.2 for further discussion)

 C Section H.3 - Support Information for Radiological Dose Assessment, which compares |

the System Energy Resources, Inc.’s assessment of the radiological impact of the |

proposed Grand Gulf ESP site with the NRC staff’s independent assessment of the |

radiological impacts of normal operation for a new nuclear unit

 C Section H.4 - References. |

H.1 Support Information for Projected Populations (Chapter 2)

The projected resident population within 16 km (10 mi) of the proposed Grand Gulf ESP facility |

is shown in Table H-1 and discussed in Chapter 2. The projected resident population within |

80 km (50 mi) of the proposed Grand Gulf ESP facility is shown in Table H-2 and also

discussed in Chapter 2.

Table H-1. Projected Resident Population within 16 Kilometers (10 Miles) of the Proposed |

Grand Gulf Early Site Permit Facility

Sector/Year

0-2 km 2-3 km 3-5 km 5-6 km 6-8 km 8-16 km

(0-1 mi) (1-2 mi) (2-3 mi) (3-4 mi) (4-5 mi) (5-10 mi) Total

North

2002 0 3 0 0 0 10 13

2030 0 3 0 0 0 10 13

2040 0 3 0 0 0 10 13

2050 0 3 0 0 0 10 13

2060 0 3 0 0 0 10 13

2070 0 3 0 0 0 10 13
Appendix H

NUREG-1817 H-2 April 2006

Table H-1. (contd)

Sector/Year

0-2 km 2-3 km 3-5 km 5-6 km 6-8 km 8-16 km

(0-1 mi) (1-2 mi) (2-3 mi) (3-4 mi) (4-5 mi) (5-10 mi) Total

N-NE

2002 0 11 0 0 0 3 14

2030 0 11 0 0 0 3 14

2040 0 11 0 0 0 3 14

2050 0 11 0 0 0 3 14

2060 0 11 0 0 0 3 14

2070 0 11 0 0 0 3 14

NE

2002 0 0 0 0 29 3 32

2030 0 0 0 0 31 3 34

2040 0 0 0 0 32 3 35

2050 0 0 0 0 33 3 36

2060 0 0 0 0 34 3 37

2070 0 0 0 0 34 4 38

E-NE

2002 0 14 0 45 27 102 188

2030 0 15 0 48 29 110 202

2040 0 15 0 50 30 112 207

2050 0 16 0 51 30 115 212

2060 0 16 0 52 31 118 218

2070 0 17 0 53 32 121 223

East

2002 0 17 0 84 68 173 342

2030 0 18 0 90 73 186 368

2040 0 19 0 93 75 191 377

2050 0 19 0 95 77 195 386

2060 0 20 0 97 79 200 396

2070 0 20 0 100 81 205 406

E-SE

2002 0 0 0 0 0 851 851

2030 0 0 0 0 0 915 915

2040 0 0 0 0 0 938 938

2050 0 0 0 0 0 961 961

2060 0 0 0 0 0 985 985

2070 0 0 0 0 0 1010 1010
Appendix H

Table H-1. (contd)

Sector/Year

0-2 km 2-3 km 3-5 km 5-6 km 6-8 km 8-16 km

(0-1 mi) (1-2 mi) (2-3 mi) (3-4 mi) (4-5 mi) (5-10 mi) Total

April 2006 H-3 NUREG-1817

SE

2002 0 0 10 0 212 3312 3534

2030 0 0 11 0 228 3560 3799

2040 0 0 11 0 234 3649 3894

2050 0 0 11 0 239 3741 3991

2060 0 0 12 0 245 3834 4091

2070 0 0 12 0 252 3930 4193

S-SE

2002 0 6 8 0 42 513 569

2030 0 6 9 0 45 551 612

2040 0 7 9 0 46 565 627

2050 0 7 9 0 47 579 643

2060 0 7 9 0 49 594 659

2070 0 7 9 0 50 609 675

South

2002 0 0 4 0 0 96 100

2030 0 0 4 0 0 99 103

2040 0 0 4 0 0 100 104

2050 0 0 4 0 0 101 105

2060 0 0 4 0 0 102 106

2070 0 0 4 0 0 103 107

S-SW

2002 0 0 0 0 0 1362 1362

2030 0 0 0 0 0 1464 1464

2040 0 0 0 0 0 1501 1501

2050 0 0 0 0 0 1538 1538

2060 0 0 0 0 0 1577 1577

2070 0 0 0 0 0 1616 1616

SW

2002 0 0 0 0 0 6 6

2030 0 0 0 0 0 6 6

2040 0 0 0 0 0 7 7

2050 0 0 0 0 0 7 7

2060 0 0 0 0 0 7 7

2070 0 0 0 0 0 7 7
Appendix H

Table H-1. (contd)

Sector/Year

0-2 km 2-3 km 3-5 km 5-6 km 6-8 km 8-16 km

(0-1 mi) (1-2 mi) (2-3 mi) (3-4 mi) (4-5 mi) (5-10 mi) Total

NUREG-1817 H-4 April 2006

W-SW

2002 0 0 0 0 0 98 98

2030 0 0 0 0 0 101 101

2040 0 0 0 0 0 102 102

2050 0 0 0 0 0 103 103

2060 0 0 0 0 0 104 104

2070 0 0 0 0 0 105 105

West

2002 0 0 0 0 0 101 101

2030 0 0 0 0 0 104 104

2040 0 0 0 0 0 105 105

2050 0 0 0 0 0 106 106

2060 0 0 0 0 0 107 107

2070 0 0 0 0 0 108 108

W-NW

2002 0 0 0 0 0 6 6

2030 0 0 0 0 0 6 6

2040 0 0 0 0 0 6 6

2050 0 0 0 0 0 6 6

2060 0 0 0 0 0 6 6

2070 0 0 0 0 0 6 6

NW

2002 0 0 0 0 0 35 35

2030 0 0 0 0 0 35 35

2040 0 0 0 0 0 35 35

2050 0 0 0 0 0 35 35

2060 0 0 0 0 0 35 35

2070 0 0 0 0 0 35 35

N-NW

2002 0 0 0 0 0 0 0

2030 0 0 0 0 0 0 0

2040 0 0 0 0 0 0 0

2050 0 0 0 0 0 0 0

2060 0 0 0 0 0 0 0

2070 0 0 0 0 0 0 0
Appendix H

Table H-1. (contd)

Sector/Year

0-2 km 2-3 km 3-5 km 5-6 km 6-8 km 8-16 km

(0-1 mi) (1-2 mi) (2-3 mi) (3-4 mi) (4-5 mi) (5-10 mi) Total

April 2006 H-5 NUREG-1817

Totals

2002 0 51 22 129 378 6671 7251

2030 0 54 23 139 406 7154 7776

2040 0 55 24 142 417 7327 7964

2050 0 56 25 146 427 7504 8157

2060 0 57 25 149 438 7686 8355

2070 0 58 26 153 449 7872 8557

Sources: SERI 2005a |

Table H-2. Projected Resident Population within 80 Kilometers (50 Miles) of the Proposed |

Grand Gulf Early Site Permit Facility

Sector/Year

16-32 km 32-48 km 48-64 km 64-80 km 0-16 km

(10-20 mi) (20-30 mi) (30-40 mi) (40-50 mi) Subtotal (0-10 mi) Total

North

2002 726 470 653 392 2,241 13 2,254

2030 770 498 692 416 2,375 13 2,388

2040 785 508 706 424 2,423 13 2,436

2050 801 518 720 432 2,471 13 2,484

2060 817 529 735 441 2,521 13 2,534

2070 833 539 749 450 2,571 13 2,584

N-NE

2002 20,890 17,721 6,377 200 45,188 14 45,202

2030 22,770 19,316 6,951 218 49,255 14 49,269

2040 23,453 19,895 7,159 225 50,733 14 50,747

2050 24,157 20,492 7,374 231 52,255 14 52,269

2060 24,882 21,107 7,595 238 53,822 14 53,836

2070 25,628 21,740 7,823 245 55,437 14 55,451

NE

2002 6000 6,132 2,005 680 14,817 32 14,849

2030 6450 6,592 2,155 731 15,928 34 15,962

2040 6611 6,757 2,209 749 16,326 35 16,361

2050 6777 6,926 2,264 768 16,735 36 16,771

2060 6946 7,099 2,321 787 17,153 37 17,190

2070 7120 7,276 2,379 807 17,582 38 17,620
Appendix H

NUREG-1817 H-6 April 2006

Table H-2. (contd)

Sector/Year

16-32 km 32-48 km 48-64 km 64-80 km 0-16 km

(10-20 mi) (20-30 mi) (30-40 mi) (40-50 mi) Subtotal (0-10 mi) Total

E-NE

2002 836 2,213 27,800 48,984 79,833 188 80,021

2030 901 2,386 29,968 52,805 86,060 202 86,262

2040 925 2,448 30,748 54,178 88,298 207 88,505

2050 949 2,511 31,547 55,586 90,593 212 90,805

2060 973 2,577 32,367 57,032 92,949 218 93,167

2070 999 2,644 33,209 58,514 95,365 223 95,588

East

2002 1,238 1,456 10,900 8,039 21,633 342 21,975

2030 1,355 1,594 11,930 8,799 23,677 368 24,045

2040 1,398 1,644 12,306 9,076 24,423 377 24,800

2050 1,442 1,696 12,693 9,362 25,192 386 25,578

2060 1,487 1,749 13,093 9,657 25,986 396 26,382

2070 1,534 1,804 13,506 9,961 26,805 406 27,211

E-SE

2002 995 1,160 7,000 8,020 17,175 851 18,026

2030 1,085 1,264 7,630 8,742 18,721 915 19,636

2040 1,117 1,302 7,859 9,004 19,282 938 20,220

2050 1,151 1,341 8,095 9,274 19,861 961 20,822

2060 1,185 1,382 8,338 9,552 20,457 985 21,442

2070 1,221 1,423 8,588 9,839 21,070 1,010 22,080

SE

2002 1,200 1,613 4,151 18,987 25,951 3,534 29,485

2030 1,308 1,758 4,525 20,696 28,287 3,799 32,086

2040 1,347 1,811 4,660 21,317 29,135 3,894 33,029

2050 1,388 1,865 4,800 21,956 30,009 3,991 34,000

2060 1,429 1,921 4,944 22,615 30,910 4,091 35,001

2070 1,472 1,979 5,092 23,293 31,837 4,193 36,030

S-SE

2002 700 483 1,764 4,226 7,173 569 7,742

2030 753 519 1,896 4,543 7,711 612 8,323

2040 771 532 1,944 4,657 7,904 627 8,531

2050 791 546 1,992 4,773 8,101 643 8,744

2060 810 559 2,042 4,892 8,304 659 8,963

2070 831 573 2,093 5,015 8,511 675 9,186
Appendix H

Table H-2. (contd)

Sector/Year

16-32 km 32-48 km 48-64 km 64-80 km 0-16 km

(10-20 mi) (20-30 mi) (30-40 mi) (40-50 mi) Subtotal (0-10 mi) Total

April 2006 H-7 NUREG-1817

South

2002 3,900 2,222 1,242 1,087 8,451 100 8,551

2030 4,017 2,289 1,279 1,120 8,705 103 8,808

2040 4,057 2,312 1,292 1,131 8,792 104 8,896

2050 4,098 2,335 1,305 1,142 8,879 105 8,984

2060 4,139 2,358 1,318 1,154 8,968 106 9,074

2070 4,180 2,382 1,331 1,165 9,058 107 9,165

S-SW

2002 1,069 8,026 16,095 10,600 35,790 1,362 37,152

2030 1,101 8,267 16,578 10,918 36,864 1,464 38,328

2040 1,112 8,349 16,744 11,027 37,232 1,501 38,733

2050 1,123 8,433 16,911 11,137 37,605 1,538 39,143

2060 1,134 8,517 17,080 11,249 37,981 1,577 39,558

2070 1,146 8,602 17,251 11,361 38,361 1,616 39,977

SW

2002 500 1,712 5,700 8,034 15,946 6 15,952

2030 530 1,815 6,042 8,516 16,903 6 16,909

2040 541 1,851 6,163 8,686 17,241 7 17,248

2050 551 1,888 6,286 8,860 17,586 7 17,593

2060 562 1,926 6,412 9,037 17,937 7 17,944

2070 574 1,964 6,540 9,218 18,296 7 18,303

W-SW

2002 1,230 1,400 2,122 1,196 5,948 98 6,046

2030 1,333 1,518 2,300 1,296 6,448 101 6,549

2040 1,371 1,560 2,365 1,333 6,628 102 6,730

2050 1,409 1,604 2,431 1,370 6,814 103 6,917

2060 1,448 1,649 2,499 1,408 7,005 104 7,109

2070 1,489 1,695 2,569 1,448 7,201 105 7,306

West

2002 300 698 3,463 3,098 7,559 101 7,660

2030 323 752 3,733 3,340 8,149 104 8,253

2040 332 772 3,830 3,426 8,360 105 8,465

2050 340 792 3,930 3,516 8,578 106 8,684

2060 349 813 4,032 3,607 8,801 107 8,908

2070 358 834 4,137 3,701 9,030 108 9,138
Appendix H

Table H-2. (contd)

Sector/Year

16-32 km 32-48 km 48-64 km 64-80 km 0-16 km

(10-20 mi) (20-30 mi) (30-40 mi) (40-50 mi) Subtotal (0-10 mi) Total

NUREG-1817 H-8 April 2006

W-NW

2002 2,012 1,700 4,586 5,946 14,244 6 14,250

2030 2,169 1,833 4,944 6,410 15,355 6 15,361

2040 2,225 1,880 5,072 6,576 15,754 6 15,760

2050 2,283 1,929 5,204 6,747 16,164 6 16,170

2060 2,343 1,979 5,339 6,923 16,584 6 16,590

2070 2,403 2,031 5,478 7,103 17,015 6 17,021

NW

2002 104 240 1,418 7,000 8,762 35 8,797

2030 113 262 1,546 7,630 9,551 35 9,586

2040 117 269 1,592 7,859 9,837 35 9,872

2050 120 278 1,640 8,095 10,132 35 10,167

2060 124 286 1,689 8,338 10,436 35 10,471

2070 128 294 1,740 8,588 10,749 35 10,784

N-NW

2002 700 3,338 8,300 2,069 14,407 0 14,407

2030 768 3,663 9,109 2,271 15,812 0 15,812

2040 793 3,783 9,405 2,345 16,326 0 16,326

2050 819 3,905 9,711 2,421 16,856 0 16,856

2060 846 4,032 10,027 2,499 17,404 0 17,404

2070 873 4,163 10,352 2,581 17,970 0 17,970

Totals

2002 42,400 50,584 103,576 128,558 325,118 7,251 332,369

2030 45,746 54,325 111,279 138,449 349,799 7,776 357,575

2040 46,955 55,673 114,054 142,012 358,694 7,964 366,658

2050 48,197 57,059 116,904 145,671 367,831 8,157 375,988

2060 49,475 58,482 119,831 149,429 377,217 8,355 385,572

2070 50,788 59,944 122,838 153,288 386,858 8,557 395,415

| Sources: SERI 2005a

H.2 Environmental Impacts of Transportation

| Section H.2 discusses the potential environmental impacts of transporting reactor fuel and

| radioactive waste to and from potential ESP sites including North Anna Power Station, Clinton

Nuclear Power Station, Grand Gulf Nuclear Station, and their associated alternative sites. 

| Section H.2.1 briefly discusses the effects of transporting unirradiated fuel to ESP sites, and

| Section H.2.2 discusses the effects of transporting spent fuel from ESP sites to a spent fuel

disposal facility. Section H.2.3 discusses the environmental effects of radioactive waste

shipments.
Appendix H

April 2006 H-9 NUREG-1817

H.2.1 Unirradiated Fuel Shipping

This section addresses the number and characteristics of shipments of unirradiated fuel to ESP

sites relative to the conditions in 10 CFR 51.52. Comparisons are also made against Table S–4

in 10 CFR 51.52(c) and WASH-1238 (AEC 1972), which provided the data that supports

Table S–4. Section H.2.1.1 presents the basic unirradiated fuel shipping requirements for each |

advanced reactor design. These data were extracted from INEEL (2003). Section H.2.1.2

presents the comparisons to 10 CFR 51.52 conditions. |

H.2.1.1 Advanced Reactor Unirradiated Fuel Shipping Data

In WASH-1238 (AEC 1972), a reference boiling water reactor (BWR) and pressurized water

reactor (PWR) were used to formulate the basic numbers of unirradiated fuel shipments required

for initial core loading and refueling. Both reference reactor types had a net electrical output of

1100 MW(e). The reference BWR assumed an initial core loading of 150 metric tons of uranium

(MTU), and the reference PWR assumed a 100 MTU initial loading. Both reactor types resulted |

in 18 truck shipments of unirradiated fuel per reactor for initial core loading. Annual reload |

quantities were assumed to be 30 MTU/yr for both reactor types, which resulted in an additional

six truck shipments per year per reactor. In total, about 252 truck shipments of unirradiated fuel

would be required over a 40-year reactor life, including the initial core and 39 years of reloads, |

for both reactor types.

The initial fuel loading and annual reload quantities for the Advanced Boiling Water Reactor

(ABWR), a 1500-MW(e) reactor, and the Economic Simplified Boiling Water Reactor (ESBWR)

are approximately the same: 156.96 MTU per reactor initial core loading and 32.76 MTU/yr per

reactor reload quantities (INEEL 2003). This equates to about 872 unirradiated fuel assemblies

in the initial core and 213 assemblies per year for refueling. Truck shipment capacities were

stated in INEEL (2003) to be 28 to 30 unirradiated fuel assemblies per truck shipment. 

Assuming 30 fuel assemblies per truck shipment, approximately 30 shipments of unirradiated |

fuel would be required to load the initial core and 6.1 truck shipments per year would be needed |

for refueling. If 28 fuel assemblies per truck shipment are used, the initial core load would |

require about 32 shipments of unirradiated fuel and annual refueling would require about |

6.5 truck shipments per year.

The surrogate AP1000 is an 1150-MW(e) advanced PWR. The initial core load was estimated |

to be 84.5 MTU per reactor, and the annual reload requirement was estimated to be |

24.4 MTU/yr per reactor. The data in INEEL (2003) also indicated that the average uranium

mass in an unirradiated surrogate AP1000 fuel assembly would be 0.583 MTU and that 12 fuel |

assemblies per truck shipment would be transported. Therefore, about 14 truck shipments |

would be needed to supply the initial core and about 3.8 truck shipments per year would be |

needed to support refueling. For a site with two reactors, these estimates would be doubled. |
Appendix H

NUREG-1817 H-10 April 2006

| The ACR-700 is an Advanced CANDU (CANada Deuterium Uranium) Reactor assumed to

| generate 731 MW(e). It was stated in INEEL (2003) that the initial core load for the ACR-700 is

| 61.3 MTU per reactor, and the annual refueling requirement is 33.1 MTU/yr per reactor. Each

fuel assembly contains 18 kg of uranium (INEEL 2003). This corresponds to 3406 fuel

| assemblies in the initial core loading and 1839 fuel assemblies per year for refueling. The range

| of truck shipment capacities given by INEEL (2003) was 180 to 240 fuel assemblies per truck

| shipment. This equates to 15 to 19 truck shipments needed to supply the initial core load and

from 7.7 to 10.2 annual refueling shipments. For a site with two reactors, these estimates would

be doubled.

The International Reactor Innovative and Secure (IRIS) design is a 335-MW(e) advanced PWR. 

It requires an initial core load of 48.67 MTU or 89 fuel assemblies per unit (546.9 kg of uranium

per fuel assembly) (INEEL 2003). For refueling, the IRIS reactor was assumed to require an

additional 6.26 MTU/yr of unirradiated fuel per reactor or about 40 unirradiated fuel assemblies

every 3.5 years. INEEL (2003) indicates that a “typical” site may contain three reactors. 

| Assuming each truck shipment carries eight fuel assemblies, the initial core load would require

| 34 truck shipments per three-reactor site, and annual refueling would require an additional

4.3 truck shipments per year per three-reactor site.

The Gas Turbine–Modular Helium Reactor (GT-MHR) is a gas-cooled reactor that uses a

substantially different fuel design than current and advanced LWRs. The reactor’s thermal

power level is rated at 600 MW(t) per reactor, and the electric generation capacity is rated at

| 285 MW(e) per reactor. A standard GT-MHR site is assumed to be composed of four reactors. 

| INEEL (2003) states that the initial core load for a single reactor would be about 1020 fuel

| assemblies. Annual average reload requirements would be 510 fuel assemblies per reactor. 

| INEEL (2003) also indicates that each truck shipment could carry 80 fuel assemblies, so for all

| four reactors, about 51 truck shipments would be required to transport the initial core load and

| about 20 truck shipments per year would be required for the annual reload requirements.

The Pebble Bed Modular Reactor (PBMR) is a gas-cooled reactor that is rated at 400 MW(t)

(165 MW(e)) per reactor. A typical PBMR site is assumed to consist of eight reactors. The

PBMR uses a substantially different fuel design than a typical LWR. INEEL (2003) states that

| each reactor requires 260,000 fuel spheres for its initial core load; 120,000 fuel spheres per

| reactor are required for annual average reloads. A total of 48,000 fuel spheres is assumed to be

transported in a typical truck shipment. As a result, it would take about 44 shipments of fuel

spheres to transport the initial core load for all eight reactors and about 20 shipments per year to

transport the annual reload quantity for all eight reactors.

| To make comparisons to Table S–4, the environmental impacts were 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
Appendix H

April 2006 H-11 NUREG-1817

adjusted to calculate impacts per site by multiplying the normalized impacts by the ratio of the

total electrical output for the advanced reactor sites to the electrical output of the reference

reactor.

H.2.1.2 Analysis of the Environmental Impacts of Unirradiated Fuel Shipments

As required by 10 CFR 51.52, applicants for a construction permit are required to submit a |

statement that the reactor and the transportation of fuel and waste to and from the reactor meet

all the conditions specified in 10 CFR 51.52(a) or 10 CFR 51.52(b). An ESP is a partial |

construction permit (10 CFR 52.21). The conditions specified in 10 CFR 51.52(a) that apply to |

unirradiated fuel include the following:

(1) The reactor core has a thermal loading less than 3800 MW. [51.52(a)(1)] |

(2) The reactor fuel is in the form of sintered UO2 pellets not exceeding 4 percent uranium-235

by weight, and the pellets are encapsulated in zircaloy rods. [51.52(a)(2)] |

(3) Unirradiated fuel is shipped to the reactor by truck. [51.52(a)(5)] |

(4) The environmental impacts of transportation of fuel and waste are as set forth in Summary

Table S–4 in 10 CFR 51.52(c). [51.52(a)(6)] |

If these conditions are not met, 10 CFR 51.52(b) requires the applicant to provide a full |

description and detailed analysis of the environmental impacts of transporting fuel and waste to |

and from the reactor, including values for the environmental impact under normal conditions of |

transport and the environmental risk from accidents in transport. |

Unirradiated fuel shipment information for the advanced reactors is discussed below for each of

these criteria.

Reactor Core Thermal Loading

The thermal output ratings of the seven advanced reactor types, as given in INEEL (2003), are

as follows:

 C ABWR – 4300 MW(t) (single reactor)

 C ESBWR – 4000 MW(t) (single reactor)

 C Surrogate AP1000 – 3400 MW(t) (single reactor)

 C ACR-700 – 1982 MW(t) per reactor x two reactors per site = 3964 MW(t) per site

 C IRIS – 1000 MW(t) per reactor x three reactors per site = 3000 MW(t) per site

 C GT-MHR – 600 MW(t) per reactor x four reactors per site = 2400 MW(t) per site

 C PBMR – 400 MW(t) per reactor x eight reactors per site = 3200 MW(t) per site.
Appendix H

NUREG-1817 H-12 April 2006

As shown above, single-unit ABWR and ESBWR plants exceed the 3800-MW(t) condition in

10 CFR 51.52(a)(1). In addition, the twin-reactor ACR-700 site exceeds the core thermal power

condition.

Reactor Fuel Form

All of the advanced LWRs (i.e., the ABWR, ESBWR, surrogate AP1000, IRIS, and ACR-700)

use sintered UO2 fuel pellets encapsulated in zircaloy rods. The average enrichment for the

| ACR-700 fuel is about 2 percent, which is well within the 10 CFR 51.52(a)(2) condition. The

average enrichments for the other advanced LWR fuels exceed the 4 percent uranium-235 by

weight condition in 10 CFR 51.52(a)(2).

| The gas-cooled reactors (i.e., the GT-MHR and PBMR) have substantially different fuel forms

| than those described in 10 CFR 51.52(a)(2). The fuel forms for these reactors are coated

uranium oxycarbide fuel kernels (GT-MHR) or coated uranium dioxide fuel kernels (PBMR). The

fuel kernels are coated with layers of pyrolitic carbon and silicone carbide. Thus, these fuel

| forms are not the same as those specified in 10 CFR 51.52(a)(2). Furthermore, the equilibrium

| enrichments for these fuels are 12.9 percent (PBMR) and 19.8 percent (GT-MHR). 

Shipping Mode

| Trucks are used to ship unirradiated fuel to the various sites for all the reactor types

(INEEL 2003).

WASH-1238 and Table S–4 of 10 CFR 51.52(c)

| The condition specified in Table S–4 that applies to shipment of unirradiated fuel limits the

number of shipments of fuel and waste to and from a commercial nuclear power plant to less

than one per day. Table H-3 summarizes the number of truck shipments of unirradiated fuel

| required for each reactor type. The numbers of shipments are normalized to the net electrical

| generation output for the reference reactor in WASH-1238 (AEC 1972) or 880 MW(e)

| (1100-MW(e)) plant operating at 80-percent annual capacity factor.

| As shown in Table H-3, the ACR-700, PBMR, and GT-MHR advanced reactor types exceed the

number of truck shipments 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 number of unirradiated fuel, spent fuel, and radioactive waste shipments

| per day equate to far less than one truck shipment per day for all reactor types. Consequently,

the numbers of shipments for all the advanced reactor types are within the conditions specified

in Table S–4 of 10 CFR 51.52. Table S–4 includes a condition that the truck shipments not
Appendix H

April 2006 H-13 NUREG-1817

Table H-3. Numbers of Truck Shipments of Unirradiated Fuel for Each Advanced Reactor Type

Reactor Type

Number of Shipments per 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

(d,e) 30 6.1 267 1500(f) 0.95 165

Surrogate AP1000 14 3.8 161 1150(f) 0.95 130

ACR-700 30 15.4 628 1462(g) 0.9 420

IRIS 34 4.3 201 1005(h) 0.96 184

GT-MHR 51 20 831 1140(I) 0.88 729

PBMR 44 20 824 1320(j) 0.95 579

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

(b) Total shipments of 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) reactor at 80 percent or net

electrical output of 880 MW(e)).

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

fuel shipment data for these reactors were derived using the upper limits of the ranges.

(f) The ABWR/ESBWR unit includes one reactor at 1500 MW(e), and the surrogate AP1000 unit includes one

reactor at 1150 MW(e).

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

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

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

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

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

exceed 33,100 kg (73,000 lb) as governed by Federal or State gross vehicle weight restrictions. 

All of the advanced reactors were indicated in INEEL (2003) to be capable of meeting this

restriction for unirradiated fuel shipments.

Finally, Table S–4 includes conditions related to radiological doses to transport workers and

members of the public along transport routes. These doses are a function of 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 time and stop time), and the

number of shipments to which the individuals are exposed. The radiological dose impacts of the

transportation of unirradiated fuel were calculated using the RADTRAN 5 computer code

(Neuhauser et al. 2003). The RADTRAN 5 calculations were performed to develop estimates of

the worker and public doses associated with annual unirradiated fuel shipments to the ESP sites.

One of the key assumptions in WASH-1238 (AEC 1972) for the reference LWR unirradiated fuel

shipments is that the radiation dose rate at 1 m (3 ft) from the transport vehicle is about
Appendix H

NUREG-1817 H-14 April 2006

0.001 mSv/hr (0.1 mrem/hr). This assumption was also used in the analysis of advanced reactor

unirradiated fuel shipments. This assumption is reasonable for all the advanced reactor fuel

types because the fuel materials will be low-dose-rate uranium radionuclides and will be

packaged similarly (i.e., inside a metal container that provides little radiation shielding). The

numbers of shipments per year were obtained by dividing the normalized shipments in Table H-3

by 40 years of operation. Other key input parameters used in the radiation dose analysis for

unirradiated fuel are shown in Table H-4.

Table H-4. RADTRAN 5 Input Parameters for Unirradiated Fuel Shipments

Parameter

RADTRAN 5

Input Value Source

Shipping distance, km 3200 AEC (1972)(a)

| Travel fraction – rural 0.90 NRC (1977a)

Travel fraction – suburban 0.05

Travel fraction – urban 0.05

Population density – rural, persons/km2 10 DOE (2002a)

Population density – suburban, persons/km2 349

Population density – urban, persons/km2 2260

Vehicle speed – rural, km/hr 88.49 Based on average speed in rural areas given

Vehicle speed – suburban, km/hr 88.49 in DOE (2002a)

Vehicle speed – urban, km/hr 88.49

Traffic count – rural, vehicles/hr 530 DOE (2002a)

Traffic count – suburban, vehicles/hr 760

Traffic count – urban, vehicles/hr 2400

Dose rate at 1 m from vehicle, mSv/hr 0.001 AEC (1972)

Packaging length, m 7.3 Approximate length of two LWR fuel element

packages placed on end

| Number of truck crew 2 AEC (1972), NRC (1977a), and DOE (2002a)

Stop time, hr/trip 4.5 Based on 0.0014-hour stop time per km

(Hostick et al. 1992)

Population density at stops, persons/km2 64,300 Based on 20 people in annular ring extending

from 1 to 10 m (3.3 to 33 ft) from the vehicle

(a) AEC (1972) provides a range of shipping distances between 40 km (25 mi) and 4800 km (3000 mi) for

unirradiated fuel shipments. A 3200-km (2000-mi) “average” shipping distance was assumed here.

The RADTRAN 5 results for this “generic” unirradiated fuel shipment are as follows:

 C Worker dose: 1.71 x 10-5 person-Sv/shipment (1.71 x 10-3 person-rem/shipment)

| C General public dose (onlookers/persons at stops and sharing the highway): 

6.65 x 10–5 person-Sv/shipment (6.65 x 10-3 | person-rem/shipment)

 C General public dose (along route - persons living near a highway): 1.61 x 10-6 personSv/shipment (1.61 x 10-4 person-rem/shipment).
Appendix H

April 2006 H-15 NUREG-1817

These values were combined with the average annual shipments of unirradiated fuel for each

advanced reactor type (see Table H-3) normalized to the WASH-1238 (AEC 1972) reference

LWR electric output (880 MW(e)) to calculate annual doses to the public and workers. The

results are compared to Table S–4 conditions. The results are shown in Table H-5. As shown,

the calculated radiation doses for shipping unirradiated fuel to advanced reactor sites are within

the conditions shown in Table S–4. |

Table H-5. Radiological Impacts of Transporting Unirradiated Fuel to ESP Sites

Plant Type

Normalized

Average Annual

Shipments

Cumulative Annual Dose, person-Sv/yr(a) per

1100 MW(e)

Workers

Public –

Onlookers

Public – Along

Route

Reference LWR (WASH-1238 (AEC 1972)) 6.3 1.1 x 10-4 4.2 x 10-4 1.0 x 10-5 |

ABWR/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 x 10-2 3.0 x 10-2 3.0 x 10-2 |

(a) Person-Sv = person-sievert; multiply person-Sv/yr times 100 to obtain dose in person-rem/yr. |

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 exposures 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 exposure 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 theory. Simply stated, 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 (ICRP) Publication 60 (ICRP 1991). All the public doses |

presented in Table H-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
Appendix H

NUREG-1817 H-16 April 2006

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

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

| radiation, based on the same risk model.

H.2.1.3 Transportation Accidents

Accidents involving unirradiated fuel shipments are also addressed in Table S–4. Accident risks

are the product of accident frequency times consequence. Accident frequencies are likely to be

lower than they were when WASH-1238 (AEC 1972) was published because traffic accident,

injury, and fatality rates have fallen over the past 30 years. Consequences of accidents that are

severe enough to result in a release of unirradiated fuel particles are not significantly different for

advanced 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 to

advanced LWR sites would be smaller than the WASH-1238 results that formed the basis for

Table S–4.

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

and associated accident frequencies (accidents per year) would follow the same trends as for

| LWRs (i.e., overall reduction relative to the accident rates used in WASH-1238). The

consequences of accidents involving gas-cooled reactor unirradiated fuel, however, are more

uncertain. A literature search was conducted to identify publicly available documents that

describe the effects of accidents (i.e., exposure of unirradiated gas-cooled reactor fuel to

structural and thermal transients). No definitive references were found. Consequently, it was

| 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 judged 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 be required to meet the same performance requirements as

| unirradiated LWR fuel packages including fissile material controls to prevent criticality under

| normal and accident conditions. Consequently, packaging systems for unirradiated gas-cooled

| reactor fuels are expected to provide protection equivalent 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., uranium oxide for the PBMR and uranium oxycarbide for the GT-MHR versus uranium

| oxide for LWRs); thus, the inherent failure resistance provided by unirradiated gas-cooled

| reactor fuels is expected to be similar to that provided by LWR fuels. Based on the assumption

| that unirradiated gas-cooled and LWR fuels and associated packaging systems provide similar

| resistance to various environmental conditions, the staff concluded that the impacts of accidents

involving unirradiated gas-cooled reactor fuel are not expected to be significantly different than

those for unirradiated LWR fuel.
Appendix H

April 2006 H-17 NUREG-1817

H.2.2 Spent Fuel Shipping

This section discusses the impact of transporting irradiated or spent advanced reactor fuel from

ESP sites to a potential high-level waste repository at Yucca Mountain, Nevada. The section is |

divided into two parts. The first part considers incident-free transportation, and the second part

considers transportation accidents.

The 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

onto a modified trailer. These assumptions are consistent with assumptions made in the

evaluation of the environmental impacts of transportation of spent fuel presented in Addendum I

to NUREG-1437 (NRC 1999). As discussed in Addendum I, these assumptions are

conservative because the alternative assumptions involve rail transportation or heavy-haul

trucks, which would reduce the number of spent-fuel shipments.

Environmental impacts of the transportation of spent fuel were calculated using the RADTRAN 5

computer code (Neuhauser et al. 2003). Routing and population data for input to RADTRAN 5

for shipment by truck were obtained from the TRAGIS routing code (Johnson and

Michelhaugh 2000). The population data in the TRAGIS code is based on the 2000

U.S. Census. |

H.2.2.1 Incident-Free Transportation of Spent Fuel

“Incident-free” transportation refers to transportation activities in which the shipments of

radioactive material reach their destination without releasing any radioactive cargo to the

environment. The vast majority of radioactive shipments are expected to reach their destination

without experiencing an accident or incident or releasing any cargo. The “incident-free” impacts

from these normal, routine shipments arise from the low levels of radiation that penetrate the

heavily shielded spent fuel shipping cask. Although Federal regulations in 10 CFR Part 71 and

49 CFR Part 173 impose constraints on radioactive material shipments, some radiation

penetrates the shipping container and exposes nearby persons to low levels of radiation.

Incident-free, legal-weight truck transportation of spent fuel has been evaluated by considering

shipments from 11 representative reactor sites to the proposed high-level waste repository at |

Yucca Mountain, Nevada, (referred to here as the proposed Yucca Mountain repository) for |

disposal. This assumption is conservative because it tends to maximize the shipping distance

from the East Coast and Midwest, where most of the reactors are assumed to be located. 

Therefore, shipment to one or more other potential sites, such as a monitored retrievable |

storage facility, would reduce the impacts. |
Appendix H

(a) Impacts were not calculated for the River Bend site because the analysis is bounded by the impacts

calculated for Grand Gulf. Impacts were not calculated for the Dresden and LaSalle sites because

they are bounded by the Braidwood analysis.

NUREG-1817 H-18 April 2006

Environmental impacts from these shipments will occur to persons residing along the

| transportation corridors between the potential advanced reactor sites and the proposed

repository; to persons in vehicles passing the spent-fuel shipment; to persons at vehicle stops

for refueling, rest, and vehicle inspections; and to transportation crew members. The impacts to

these exposed population groups were quantified using the RADTRAN 5 computer code

(Neuhauser et al. 2003).

| This analysis assumes that all spent nuclear fuel will be transported to the proposed Yucca

| Mountain repository because Congress has directed (Nuclear Waste Policy Act of 1982, as

| amended) the U.S. Department of Energy to study only Yucca Mountain for the proposed

repository. 

| The characteristics of specific shipping routes (e.g., population densities and shipping distances)

influence the normal radiological exposures. To address the differences that arise from the

specific reactor site from which the spent fuel shipment originates, each advanced reactor

design was assumed to be located at all of the primary and alternative ESP sites. These sites

are:

| C Primary Sites

| - North Anna Power Station, Virginia

| - Clinton Nuclear Power Station, Illinois

| - Grand Gulf Nuclear Power Station, Mississippi

 C Alternative Sites(a) |

| - Savannah River Site (SRS), South Carolina

| - Portsmouth Gaseous Diffusion Plant (PGDP), Ohio

| - FitzPatrick Nuclear Power Plant, New York

| - Pilgrim Nuclear Power Station, Massachusetts

| - Zion Nuclear Power Station, Illinois

| - Quad Cities Nuclear Power Station, Iowa

| - Braidwood Nuclear Power Station, Illinois

| - Surry Power Station, Virginia

Input to RADTRAN 5 includes the total shipping distance between the origin and destination

sites and the population distributions along the routes. This information was obtained by running

the TRAGIS computer code (Johnson and Michelhaugh 2000) for the origin-destination

combinations of interest for legal-weight trucks. The resulting route characteristics information is
Appendix H

April 2006 H-19 NUREG-1817

shown in Table H-6. Note that for truck shipments, all the spent fuel is assumed to be shipped

to the proposed Yucca Mountain repository over designated controlled-quantity highway routes. |

The routes used here are the same as those used in the Yucca Mountain Environmental Impact

Statement (DOE 2002b).

Table H-6. Transportation Route Information for Shipments from ESP Sites to the Proposed |

High-Level Waste Repository at Yucca Mountain |

ESP Site

One-Way Shipping Distance, km Population Density, persons/km2 Stop |

Time per

Total Rural Suburban Urban Rural Suburban Urban Trip, hr

Primary Site

North Anna 4409.5 3498 812.4 99.1 11.3 319 2310.6 5

Clinton 3076.3 2626.3 398.3 51.7 9.4 306.1 2372.2 3.5

Grand Gulf(a) 3718.3 3030.4 581.3 106.6 9.2 339.4 2429.4 4

Alternative Site

Savannah River Site 4263 3260 881 122 11 331.5 2311.2 5 |

Portsmouth Gaseous

Diffusion Plant |

3902.2 3166.9 647.2 88.1 10.7 316.4 2339.7 4.5

FitzPatrick 4212.2 3228.6 875.4 108.2 11.4 312.4 2348.7 5

Pilgrim 4682.3 3469.3 1091.7 121.3 11.8 312.3 2377.2 5.5

Zion 3138.9 2629.6 441.3 68 9.5 323.8 2360.3 3.5

Quad Cities 2853.1 2451 352.6 49.5 9.1 310.2 2391.3 3

Braidwood(b) 3034.5 2604.4 378.7 51.4 9.4 308.9 2377.2 3.5

Surry 4555.4 3590.7 863.9 100.8 11.4 317.6 2301.6 5

(a) The River Bend alternative site can be assumed to be bounded by the Grand Gulf values because of the

proximity of the sites.

(b) Dresden and LaSalle can be assumed to be bounded by the Braidwood values because of the proximity of the

sites.

Shipping casks have not been designed for advanced reactor spent fuel. Although some of the

advanced reactor fuel designs are similar to current LWR fuel, no attempt has been made to

optimize the cargo capacities of shipping casks for advanced LWR fuels. For the non-LWR fuel

types (i.e., the GT-MHR and PBMR), there is little information on even a conceptual basis that

would provide a defensible technical basis for shipping-cask capacities. The shipping-cask

capacity data in the Early Site Permit Environmental Report Sections and Supporting

Documentation (INEEL 2003) is summarized as follows: |

 C ABWR – The ABWR fuel is not significantly different from existing LWR fuel designs;

thus, the number of ABWR assemblies that can be transported in a legal-weight truck

shipment (i.e., 23 MT [25-ton] shipping cask) is not expected to be different from current |

cargo capacities.
Appendix H

NUREG-1817 H-20 April 2006

 C ESBWR – The ESBWR fuel is similar to the ABWR fuel.

| C Surrogate AP1000 – The surrogate AP1000 fuel assemblies are similar to currentgeneration PWR fuel. No information was provided in INEEL (2003) on shipping cask

capacities for surrogate AP1000 spent nuclear fuel.

 C ACR-700 – The ACR-700 fuel is somewhat different from the current and advanced LWR

| fuel designs. System Energy Resources, Inc. (SERI) estimated that an ACR-700 rail

cask would hold about 10 MTU of spent fuel, similar to the current cask designs. This

value is nearly identical to the cargo capacities of current rail cask designs; thus, it was

assumed that the truck cask capacity for ACR-700 and current-generation LWRs would

also be about the same (i.e., 1.8 MTU/shipment).

 C IRIS – The IRIS fuel is similar to current-generation PWR fuel. No information was

provided in INEEL (2003) on shipping-cask capacities for IRIS spent nuclear fuel.

 C GT-MHR – The GT-MHR fuel is a spherical coated-particle fuel with a uranium

oxycarbide fuel kernel loaded into graphite fuel assemblies. This fuel concept is

significantly different from current and advanced LWR fuels (sintered UO2 pellets loaded

into zircaloy tubes). According to INEEL (2003), six spent fuel assemblies containing

0.023 MTU of spent fuel is assumed to be transported in a legal weight truck cask.

 C PBMR – The PBMR fuel is also a spherical coated-particle fuel with uranium oxide fuel

kernels. INEEL (2003) estimated that 0.495 MTU of spent PBMR fuel can be transported

in a single legal-weight truck shipment.

| These shipping cask capacities are approximations based on current shipping cask designs. 

Actual shipping cask capacities in the future may be significantly different. Applicants must

| account for changes in shipping cask capacities in applications at the construction permit or

combined operating license stage.

Incident-free radiation doses are a function of many variables. The most important of these

variables are presented in Table H-7. Most of these variables, which are extracted from the

literature, are considered to be “standard” values used in many RADTRAN 5 applications,

| including environmental impact statements and regulatory analyses.
Appendix H

April 2006 H-21 NUREG-1817

Table H-7. RADTRAN 5 Incident-Free Exposure Parameters

Parameter

RADTRAN 5

Input Value Source

Vehicle speed – rural, km/hr 88.49 Based on average speed in rural areas given in

DOE (2002a). Because m ost travel is on interstate

highways, the same vehicle speed is assumed in

rural, suburban, and urban areas. No speed

reductions were assum ed for travel at rush hour.

Vehicle speed – suburban, km/hr 88.49

Vehicle speed – urban, km/hr 88.49

Traffic count – rural, vehicles/hr 530 DOE (2002a)

Traffic count – suburban, vehicles/hr 760

Traffic count – urban, vehicles/hr 2400

Dose rate at 1 m from vehicle,

 mSv/hr

0.14 Approximate dose rate at 1 m (3 ft) that is

equivalent to maximum dose rate allowed by the |

U.S. Department of Transportation and NRC |

regulations (i.e., 0.1 mSv/hr at 2 m (~7 ft) from the

side of a transport vehicle) (DOE 2002b)

Packaging dim ensions, m Length – 5.2

Diam eter – 1.0

DOE (2002b)

Number of truck crew 2 (AEC 1972; NRC 1977a; DOE 2002a) |

Stop tim e, hr/trip Route-specific See Table H-6.

Population density at stops,

 persons/km2

30,000 Sprung et al. (2000)

Min/max radii of annular area around

 vehicle at stops, m

1 to 10 Sprung et al. (2000)

Shielding factor applied to annular

 area surrounding vehicle at stops

1

(no shielding)

Sprung et al. (2000)

Population density surrounding truck

 stops, persons/km2

340 Sprung et al. (2000)

Min/max radius of annular area

 surrounding truck stop, m

10 to 800 Sprung et al. (2000)

Shielding factor applied to annular

 area surrounding truck stop

0.2 Sprung et al. (2000)

For purposes of this Section H.2 analysis, the transportation crew for spent fuel shipments |

delivered by truck is assumed to consist of two drivers. Escorts were considered, but they were |

not included because their distance from the shipping cask would reduce the dose rates to levels

well below the dose rates experienced by the drivers. Stop times were assumed to accrue at the

rate of 30 minutes per 4-hour driving time. TRAGIS outputs were used to determine the number

of stops for each origin-destination.

Doses to the public at truck stops have been significant contributors to the doses calculated in

previous RADTRAN 5 analyses. For this Section H.2 analysis, stop doses are the sum of the |
Appendix H

NUREG-1817 H-22 April 2006

doses to individuals located in two annular rings centered at the stopped vehicle, as illustrated in

Figure H-1. The inner ring represents persons who may be at the truck stop at the same time as

a spent fuel shipment and extends 1 to 10 m from the edge of the vehicle. The outer ring

represents persons who reside near a truck stop and extends from 10 to 800 m from the vehicle. 

This scheme is the same as that used in Sprung et al. (2000).

Figure H-1. Illustration of Truck Stop Model (Sprung et al. 2000)

Population densities and shielding factors were also taken from Sprung et al. (2000) and were

based on the observations of Griego et al. (1996).

The results of these routine (incident-free) exposure calculations are shown in Table H-8 for

| spent fuel shipments from all 11 primary and alternative sites to the proposed Yucca Mountain

| repository. Population dose estimates are given for workers (i.e., truck crew members),
Appendix H

April 2006 H-23 NUREG-1817

onlookers (doses to persons at truck stops and persons and on highways exposed to the spent

fuel shipments), and along the route (persons living near the highway).

Table H-8. Routine (Incident-Free) Radiation Doses to Transport Workers and the Public from

Shipping Spent Fuel from Potential ESP Sites to the Proposed High-Level Waste |

Repository at Yucca Mountain |

Reactor Site

Population Dose, person-Sv/shipment(a)

Crew Onlookers Along Route

Braidwood(b) 7.1 x 10-4 2.4 x 10-3 4.4 x 10-5

Clinton 7.2 x 10-4 2.5 x 10-3 4.5 x 10-5

FitzPatrick 9.8 x 10-4 3.5 x 10-3 9.5 x 10-5

Grand Gulf(c) 8.7 x 10-4 2.8 x 10-3 7.0 x 10-5

North Anna 1.0 x 10-3 3.5 x 10-3 9.2 x 10-5

Pilgrim 1.1 x 10-3 3.9 x 10-3 1.2 x 10-4

Portsm outh 9.1 x 10-4 3.2 x 10-3 7.3 x 10-5

Quad Cities 6.7 x 10-4 2.1 x 10-3 4.1 x 10-5

Savannah River 9.9 x 10-4 3.5 x 10-3 1.0 x 10-4

Surry 1.1 x 10-3 3.5 x 10-3 9.7 x 10-5 |

Zion 7.3 x 10-4 2.5 x 10-3 5.2 x 10-5

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

(b) The River Bend alternative site can be assumed to be bounded by the Grand Gulf values because of the

proximity of the sites.

(c) Dresden and LaSalle can be assumed to be bounded by the Braidwood values because of the proximity of

the sites.

This discussion addresses whether or not the environmental effects of incident-free advanced

reactor spent fuel shipments are within the guidelines established in Table S–4. The bounding

cumulative doses to the exposed population given in Table S–4 are:

 C Transport workers 0.04 person-Sv (4 person-rem)

per reference reactor year.

 C General public (onlookers and along route) 0.03 person-Sv (3 person-rem) |

per reference reactor year.

Calculation of the cumulative doses entailed converting the per-shipment risks given in

Table H-8 to estimates of environmental effects per reference reactor year of operation. The

per-shipment results, which are independent of reactor type (i.e., the doses are dependent on

the assumed external radiation dose rate emitted from the cask, which is fixed at the regulatory

maximum limit for all of the advanced reactor types), are given in terms of the population dose

per shipment of spent fuel. To develop estimates of the annual environmental impacts, the

following assumptions were made:
Appendix H

NUREG-1817 H-24 April 2006

 C The basis for the annual number of shipments of spent fuel from the reference LWR in

WASH-1238 (AEC 1972) will be used. In WASH-1238, it was assumed that

| 60 shipments per year would be made, each shipment carrying 0.5 MTU of spent fuel. 

This equates to shipping 30 MTU of spent fuel per year. This is equivalent to the annual

refueling requirements for the reference LWR. It was assumed that the other reactor

types would also ship spent fuel at a rate equal to their annual refueling requirements.

 C Shipping cask capacities that were used to calculate annual spent fuel shipments for the

| advanced LWRs were assumed to be the same as for the reference LWR

| (i.e., approximately 0.5 MTU per truck shipment).

 C The annual numbers of spent fuel shipments from the advanced gas-cooled reactors

| were taken directly from INEEL (2003). These estimates were 34 shipments per year

| from a GT-MHR site and 12 shipments per year from the PBMR site.

Table H-9 provides the estimated annual population doses from routine (incident-free)

| transportation of spent fuel from ESP sites to the proposed Yucca Mountain repository. The

| results in Table H-9 have been normalized to the WASH-1238 (AEC 1972) net electrical

generation (i.e., 880 MW(e)). 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

| exposure 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, the 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 over-estimates 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 (ICRP) Publication 60 (ICRP 1991). All the population

doses presented in Table H-9 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 occur

| annually in the same population from exposure to natural sources of radiation.
Appendix H

April 2006 H-25 NUREG-1817

As shown in Table H-9, some of the estimated population doses are higher than the Table S–4

conditions. 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 used in this assessment than were used in WASH-1238 (AEC 1972).

WASH-1238 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 2900 km (1800 mi) to

4700 km (2900 mi). The higher numbers of shipments are based on spent fuel shipping-casks

designed to transport short-cooled fuel (150 days out of the reactor). It was assumed in this

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 designs are based on

longer-cooled spent fuel (5 years out of reactor) and have larger capacities than those used in

this assessment. DOE (2002b) 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 will reduce the number of spent fuel shipments and the associated

environmental impacts. If the population doses are adjusted for the shipping distance (a factor

of 2 to 3) and shipping cask capacity (a factor of 4), the routine population doses from spent fuel

shipments from all reactor types and all sites fall within the Table S–4 conditions.

Most of the stops made for actual spent fuel shipments are short duration stops

(i.e., 10 minutes) for brief visual inspections of the cargo (e.g., checking the cask tie-downs).

These stops typically occur in areas devoid of people, such as overpasses or freeway ramps in |

unpopulated areas. Therefore, doses to residents surrounding these types of stops are |

negligible. In DOE (2002b), close-proximity exposures (i.e., from 1 to 15.8 m from the cask)

were not assumed to occur at the short-duration inspection stops. In this analysis, for the

purpose of developing bounding estimates of environmental effects, close-proximity (1 to 10 m

from cask) exposures at all truck stops were included in the RADTRAN 5 calculations. Because

the numbers of stops in this analysis are effectively doubled relative to DOE (2002b), truck stop

doses are also doubled. The doses to residents would also be lower; however, because doses

to residents are two to three orders of magnitude (i.e., a factor of 100 to 1000) less than the

calculated close-proximity doses, this reduction does not affect the total stop dose.

The number of exposed persons at stops is higher in this Section H.2 analysis by about a factor |

of 1.5 relative to DOE (2002b) assumptions (6.9 persons in DOE 2002b versus 10 persons

assumed in this analysis). Thus, the bounding doses calculated in this analysis are also a factor

of 1.5 (10 divided by 6.9) greater than those given in DOE (2002b). 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 for food and refueling observed by Griego et al. (1996) are

on the order of 18 minutes. This amounts to another factor of 1.5 increase in stop doses

calculated here relative to DOE (2002b).
Table H-9. Routine (Incident-Free) Population Doses from Spent Fuel Transportation, Normalized to Refer

LWR Net Electrical Generation

Reactor Type Reference LWR (WASH-1238) ABWR/ESBWR Surrogate AP1000 

No. Shipments

per year 60 41 40 

Environmental Effects, person-Sv per reference reactor year(a)

Reactor Site Crew Onlookers

Along

Route Crew Onlookers

Along

Route Crew Onlookers

Along

Route Crew 

Braidwood(b) 4.2 x 10-2 1.5 x 10-1 2.6 x 10-3 2.9 x 10-2 1.0 x 10-1 1.8 x 10–3 2.8 x 10-2 9.7 x 10-2 1.7 x 10-3 6.3 x 10-2 

Clinton 4.3 x 10-2 1.5 x 10-1 2.7 x 10-3 2.9 x 10-2 1.0 x 10-1 1.8 x 10-3 2.8 x 10-2 9.7 x 10-2 1.8 x 10-3 6.4 x 10-2 FitzPatrick 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 Grand Gulf(c) 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 North Anna 6.2 x 10-2 2.1 x 10-1 5.5 x 10-3 4.2 x 10-2 1.4 x 10-1 3.7 x 10-3 4.1 x 10-2 1.4 x 10-1 3.6 x 10-3 9.2 x 10-2 Pilgrim 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 

|

Portsmouth 5.5 x 10-2 1.9 x 10-1 4.4 x 10-3 3.7 x 10-2 1.3 x 10-1 3.0 x 10-3 3.6 x 10-2 1.2 x 10-1 2.9 x 10-3 8.1 x 10-2 Quad Cities 4.0 x 10-2 1.3 x 10-1 2.4 x 10-3 2.7 x 10-2 8.6 x 10-2 1.7 x 10-3 2.6 x 10-2 8.4 x 10-2 1.6 x 10-3 6.0 x 10-2 Savannah River 6.0 x 10-2 2.1 x 10-1 6.0 x 10-3 4.0 x 10-2 1.4 x 10-1 4.1 x 10-3 3.9 x 10-2 1.4 x 10-1 4.0 x 10-3 8.9 x 10-2 Surry 6.4 x 10-2 2.1 x 10-1 5.8 x 10-3 4.3 x 10-2 1.4 x 10-1 3.9 x 10-3 4.2 x 10-2 1.4 x 10-1 3.8 x 10-3 9.5 x 10-2 

|

Zion 4.4 x 10-2 1.5 x 10-1 3.1 x 10-3 3.0 x 10-2 1.0 x 10-1 2.1 x 10-3 2.9 x 10-2 9.7 x 10-2 2.0 x 10-3 6.5 x 10-2 

April 2006 H-26 NUREG-1817
Table H-9. (contd)

Reactor Type IRIS GT-MHR PBMR

No. Shipments 

per year 35 34 12

Environmental Effects, person-rem per reference reactor year(a)

Reactor Site Crew Onlookers Along Route Crew Onlookers Along Route Crew Onlookers 

Braidwood 2.5 x 10-2 8.5 x 10-2 1.5 x 10-3 2.4 x 10-2 8.2 x 10-2 1.5 x 10-3 7.9 x 10-3 2.7 x 10-2 

Clinton 2.5 x 10-2 8.5 x 10-2 1.6 x 10-3 2.4 x 10-2 8.2 x 10-2 1.5 x 10-3 8.0 x 10-3 2.8 x 10-2 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 Grand Gulf 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 North Anna 3.6 x 10-2 1.2 x 10-1 3.2 x 10-3 3.4 x 10-2 1.2 x 10-1 3.1 x 10-3 1.2 x 10-2 4.0 x 10-2 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 Portsmouth 3.1 x 10-2 1.1 x 10-1 2.5 x 10-3 3.0 x 10-2 1.1 x 10-1 2.4 x 10-3 1.0 x 10-2 3.6 x 10-2 Quad Cities 2.3 x 10-2 7.4 x 10-2 1.4 x 10-3 2.2 x 10-2 7.1 x 10-2 1.4 x 10-3 7.5 x 10-3 2.4 x 10-2 Savannah River 3.4 x 10-2 1.2 x 10-1 3.5 x 10-3 3.3 x 10-2 1.2 x 10-1 3.3 x 10-3 1.1 x 10-2 3.9 x 10-2 Surry 3.7 x 10-2 1.2 x 10-1 3.3 x 10-3 3.5 x 10-2 1.2 x 10-1 3.2 x 10-3 1.2 x 10-2 4.0 x 10-2 Zion 2.5 x 10-2 8.5 x 10-2 1.8 x 10-3 2.4 x 10-2 8.2 x 10-2 1.7 x 10-3 8.2 x 10-3 2.8 x 10-2 (a) Multiply person-Sv/yr by 100 to obtain doses in person-rem/yr. (b) The River Bend alternative site can be assumed to be bounded by the Grand Gulf values because of the proximity of the (c) Dresden and LaSalle can be assumed to be bounded by the Braidwood values because of the proximity of the sites. 

NUREG-1817 H-27 April 2006
Appendix H

NUREG-1817 H-28 April 2006

Based on these observations, the staff concluded that the stop model used in this study

overestimates public doses at stops by approximately a factor of four (factor of two for

close-proximity exposure time at stops, a factor of 1.5 for average stop time at food and

refueling stops, and a factor of 1.5 for the number of people in proximity to the shipping cask).

Coupled with the factor of two reduction in shipping cask dose rates that result from fuel aging,

the doses to onlookers at stops could be reduced to about one-eighth of the doses shown in

Table H-9 [1/(2 x 1.5 x 1.5 x 2) . 0.12] to reflect more realistic truck shipping conditions. Based

| on the previous discussion, use of more realistic dose rates, shipping cask capacities, and truck

stop model assumptions in the RADTRAN 5 calculations could substantially reduce the

environmental effects presented in Table H-9.

Table H-10 provides a comparison between the radiological incident-free doses calculated in

| NUREG-0170 (NRC 1977a) and those calculated here. The table also summarizes the key

| incident-free input parameters used in NUREG-0170 and in this study. Comparisons are also

made between the doses for spent fuel shipments in NUREG-0170 and doses calculated for a

| shipment from the Quad Cities, Iowa, to the proposed Yucca Mountain repository because the

shipping distances are comparable (2530 km in NUREG-0170 versus 2853 km for Quad Cities

to Yucca Mountain). As shown in the table, many parameters have changed over the years and

the technical bases for them have improved. For example, the work of Griego et al. (1996) has

improved the basis for assumptions about stop times and persons exposed at truck stops, and

the TRAGIS computer code has improved the basis for shipping distances and population

distributions along highway routes.

The incident-free impacts at truck stops shown in the table have been adjusted, as discussed

above, to reflect more realistic conditions than assumed in the bounding analysis. Adjustments

were not made to the onlookers, along route, and crew doses shown in Table H-9. As shown,

| the adjusted doses in Table H-10 for spent fuel shipments from the Quad Cities to the proposed

| Yucca Mountain repository are about a factor of two lower than the per-shipment doses from

NUREG-0170 when the doses to and doses associated with in-transit storage from

| NUREG-0170 are excluded. Storage doses were excluded from this Section H.2 analysis

because spent fuel shipments proceed directly from the reactor site to Yucca Mountain with no

intermediate storage involved. Handler doses were excluded from this analysis because doses

| to workers who load the spent fuel cask at reactors and unload them at the proposed repository

| are treated as facility doses, not transportation doses.
Appendix H

April 2006 H-29 NUREG-1817

Table H-10. Comparison of Incident-Free Doses from NUREG-0170 (NRC 1977a) Spent Fuel |

Shipments and Spent Fuel Shipment from Quad-Cities to the Proposed High-Level |

Waste Repository at Yucca Mountain |

Incident-Free Exposure Parameter |

NUREG-0170

(NRC 1977a)

This Study (Quad Cities to

Yucca Mountain)(a)

One-way shipping distance, km 2530 2853

Travel fraction

 Urban 0.05 0.02

 Suburban 0.05 0.12

 Rural 0.9 0.86

Population density along highway, persons per km2

 Urban 3861 2391.3

 Suburban 719 310.2

 Rural 6 9.1

Speed, km/hr

 Urban 24 88

 Suburban 40 88

 Rural 88 88

Traffic count, vehicles/hr

 Urban 2800 2400

 Suburban 780 760

 Rural 470 530

Shipment dose Rate, mSv/hr at 2m 0.1 0.1

Crew dose rate, mSv/hr 0.02 Calculated (7.4 m from package)

Stop tim e, hr per trip

 Urban 2 3 hours per trip (30 minutes per

 Suburban 5 4 hours driving time)

 Rural 1

Population density at stops (per km2)

 Urban 3861 Distribution: 1 to 10 m - 30,000;

10 to 800 m - 340 (see

Figure G-1)

 Suburban 719

 Rural 6

Person-Sv/shipment

Crew 1.2 x 10-3 4.8 x 10-4

Off-link 1.5 x 10-4 3.1 x 10-4

On-link 7.4 x 10-5 1.7 x 10-4

Stops 1.9 x 10-4 1.7 x 10-4(b)

Total 1.6 x 10-3 8.5 x 10-4

Handlers + Storage 2.1 x 10-3 Not calculated

Grand Total 3.7 x 10-3 8.5 x 10-4

(a) Tables H-7 and H-9 provide the basis for these input parameters. |

(b) Stop doses have been adjusted as described in the text to reflect more realistic assumptions than were used in

the bounding analysis (Table H-9).
Appendix H

NUREG-1817 H-30 April 2006

H.2.2.2 Transportation Accident Impacts

| RADTRAN 5 assesses accident risk by calculating a risk value, which is the product of

| probabilities and the consequences of accidents. 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 (e.g., accidents in

| which the shipping container is exposed to severe mechanical and thermal conditions).

Radionuclide inventories are important parameters in the calculation of accident risks. The

| radionuclide inventories used in this analysis were taken directly from the Early Site Permit

| Environmental Report Sections and Supporting Documentation (INEEL 2003). The 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 identifies the radionuclides that will contribute more than

99.999 percent of the dose from inhalation.

A sum-of-fractions approach was used for this screening. First, the inventory of each

radionuclide was multiplied by its respective inhalation dose conversion factor, taken from

| Federal Guidance Report 13 (EPA 2002). These values were then summed. Then, each

inventory-conversion factor product was divided by the sum of the products to obtain the fraction

of the total inhalation dose for each radionuclide. The resulting fractions were then sorted from

largest to smallest, their cumulative contributions were calculated, and those that contributed to

| 99.999 percent of the inhalation-dose potential were selected. Two gases, krypton-85 and

iodine-129, were added to the list because they are more easily released than the solid and

semi-volatile species contained in the fuel.

The inventories of radionuclides used in this study are shown in Table H-11. 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 INEEL (2003) did not provide radionuclide source terms for radioactive

| material deposited on the external surfaces of LWR spent fuel rods, which is commonly referred

| to as “crud.” In addition, data on activation products was provided for only the ABWR. The

| ABWR spent fuel transportation risks were calculated assuming the entire Co-60 inventory is in

| the form of crud. This is very conservative as the source term used here 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 will need to be

| examined at the construction permit or combined operating license stage.
Appendix H

April 2006 H-31 NUREG-1817

Table H-11 shows that the dominant radionuclides are approximately the same regardless of |

fuel type. The table does not show radionuclide inventory data for the ACR-700 and IRIS |

advanced reactors, as those were not given in INEEL (2003). Nor were they provided in |

WASH-1238 (AEC 1972) for the reference LWR. Consequently, accident risks were not |

quantified for these reactor types. |

Table H-11. Radionuclide Inventories Used in the Transportation Accident Risk Calculations for

Each Advanced Reactor Type

Radionuclide

ABWR and ESBWR

Inventory, Bq/MTU(a)

Surrogate AP1000

Inventory, Bq/MTU

GT-MHR

Inventory, Bq/MTU

PBMR

Inventory, Bq/MTU

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

I-129 1.55 x 1091.55 x 1091.55 x 1091.55 x 109|

Kr-85 3.29 x 1014 3.29 x 1014 3.29 x 1014 3.29 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) To convert Bq/MTU to Ci/MTU, divide the value by 3.7 x 1010.

(b) Co-60 is an activation product. Only the ABWR/ESBWR submittal in INEEL (2003) provided inventory data for |

activation products.
Appendix H

NUREG-1817 H-32 April 2006

Robust shipping casks are used to transport spent fuel because of the heavy radiation shielding

and accident resistance required by 10 CFR Part 71. Spent fuel shipping casks must be

certified Type B packaging systems, which means they must withstand a series of severe

hypothetical accident conditions with essentially no loss of containment or shielding capability. 

| These casks are also designed with fissile material controls to ensure that the spent fuel

| remains subcritical under both normal and accident conditions. The tests include a 9-m (30-ft)

free drop onto an unyielding surface, a drop onto a puncture probe, an exposure to an engulfing

800°C fire for 30 minutes, and an underwater immersion. According to Sprung et al. (2000), the

probability of encountering accident conditions more severe than these tests that could lead to

| shipping cask failure are less than 0.01 percent of all accidents (i.e., more than 99.99 percent of

all accidents would not result in a release of radioactive material from the shipping cask). It was

assumed that shipping casks for advanced reactor spent fuels will provide equivalent mechanical

and thermal protection of the spent fuel cargo.

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

estimates of the annual accident risks associated with spent fuel shipments from each potential

ESP site. As was done for routine exposures, it was assumed that the numbers of shipments of

spent fuel per year are equivalent to the annual discharge quantities: 32.76 MTU/yr for the

| ABWR and ESBWR; 24.4 MTU/yr for a single-reactor surrogate AP1000 site; 6.8 MTU/yr for the

four-reactor GT MHR site; and 8.3 MTU/yr for the eight-reactor PBMR site. These data were

taken from INEEL (2003) and have not been normalized to the reference LWR net electrical

generation.

Route-specific accident rates (accidents per km) were derived for the RADTRAN 5 accident risk

analysis. The approach used to develop accident rates for spent fuel shipments is as follows. 

The TRAGIS data provide estimates of the distance traveled in each state along a route and the

type of highway (interstate, state highway, or other). Saricks and Tompkins (1999) provide

accident rates for each state that are a function of highway type. The approach taken to

estimate route-specific accident rates was to multiply the state-level accident or fatality rates by

the distances traveled in each state on the corresponding highway type and then sum over all

the states on each route. For example, for interstate highways, the interstate distances and

interstate accident rates were used. For non-interstate highway travel, either the “Primary” or

| “Other” accident rates given by Saricks and Tompkins (1999) were used. This approach allowed

computation of route-specific accident rates.

Transportation accident risk analysis in RADTRAN 5 is performed using an accident severity and

package release model. The user can define up to 30 severity categories, with each category

increasing in magnitude. Severity categories are related to fire, puncture, crush, and immersion

| environments created in vehicular accidents. For this analysis, the 19 severity categories

| defined by Sprung et al. (2000) were adopted.
Appendix H

April 2006 H-33 NUREG-1817

Each severity category has an assigned conditional probability (or the probability, given an

accident occurs, that it will be of the specified severity). The accident scenarios are further

defined by allowing the user to input release fractions and aerosol and respirable fractions for

each severity category. These fractions are a function of the physical-chemical properties of the

materials being transported as well as the mechanical and thermal accident conditions that

define the severity categories. The severity and release fractions used here are presented in

Table H-12.

Table H-12. Severity and Release Fractions Used to Model Spent Fuel Transportation

Accidents (Sprung et al. 2000)

Severity

Category |

Severity

Fraction(b)

Release Fractions(a)

Gas Cesium Ruthenium Particulates

Corrosion

Products |

1 1.53 x 10-8 0.8 2.4 x 10-8 6.0 x 10-7 6.0 x 10-7 2.0 x 10-3

2 5.88 x 10-5 0.14 4.1 x 10-9 1.0 x 10-7 1.0 x 10-7 1.4 x 10-3

3 1.81 x 10-6 0.18 5.4 x 10-9 1.3 x 10-7 1.3 x 10-7 1.8 x 10-3

4 7.49 x 10-8 0.84 3.6 x 10-5 3.8 x 10-6 3.8 x 10-6 3.2 x 10-3

5 4.65 x 10-7 0.43 1.3 x 10-8 3.2 x 10-7 3.2 x 10-7 1.8 x 10-3

6 3.31 x 10-9 0.49 1.5 x 10-8 3.7 x 10-7 3.7 x 10-7 2.1 x 10-3

7 0 0.85 2.7 x 10-5 2.1 x 10-6 2.1 x 10-6 3.1 x 10-3

8 1.13 x 10-8 0.82 2.4 x 10-8 6.1 x 10-7 6.1 x 10-7 2.0 x 10-2

9 8.03 x 10-11 0.89 2.7 x 10-8 6.7 x 10-7 6.7 x 10-7 2.2 x 10-3

10 0 0.91 5.9 x 10-6 6.8 x 10-7 6.8 x 10-7 2.5 x 10-3

11 1.44 x 10-10 0.82 2.4 x 10-8 6.1 x 10-7 6.1 x 10-7 2.0 x 10-3

12 1.02 x 10-12 0.89 2.7 x 10-8 6.7 x 10-7 6.7 x 10-7 2.2 x 10-3

13 0 0.91 5.9 x 10-6 6.8 x 10-7 6.8 x 10-7 2.5 x 10-3

14 7.49 x 10-11 0.84 9.6 x 10-5 8.4 x 10-5 1.8 x 10-5 6.4 x 10-3

15 0 0.85 5.5 x 10-5 5.0 x 10-5 9.0 x 10-6 5.9 x 10-3

16 0 0.91 5.9 x 10-6 6.4 x 10-6 6.8 x 10-7 3.3 x 10-3

17 0 0.91 5.9 x 10-6 6.4 x 10-6 6.8 x 10-7 3.3 x 10-3

18 5.86 x 10-6 0.84 1.7 x 10-5 6.7 x 10-8 6.7 x 10-8 2.5 x 10-3

19 0.99993 0 0 0 0 0

(a) RADTRAN 5 also models the fraction of the released particulate material that is small enough to be dispersible |

in prevailing wind conditions and the fraction that is respirable. For this analysis, these parameters were set to

1.0 (i.e., 100 percent dispersible and 100 percent respirable).

(b) Severity fractions are the conditional probabilities, given the occurrence of an accident, that the mechanical and

thermal conditions experienced by a spent fuel shipping cask are within the conditions defined by the Severity

Category. See Sprung et al. (2000) for detailed information about the derivation of these data. Generic |

steel-depleted uranium-steel cask designs were assumed for the severity fractions. |

The severity categories and release fractions published by Sprung et al. (2000) were designed |

specifically to address accidents involving current generation LWR fuel and the current

generation of spent fuel shipping casks. While some of the advanced reactor fuel designs are

similar to current-generation reactor fuel designs (e.g., the ABWR, ESBWR, Surrogate AP1000, |

ACR-700, and IRIS), others are significantly different, including the GT-MHR and PBMR. 
Appendix H

(a) The 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.

NUREG-1817 H-34 April 2006

Extrapolating the current generation of LWR fuel and shipping casks to advanced LWR fuels

| and shipping casks is expected to be relatively straightforward because the fuel form, cladding,

and physical and mechanical properties are similar. Furthermore, substantial experimental data

exist to develop technically defensible release fractions for various radionuclide groups (e.g.,

gases, semi-volatiles such as cesium and ruthenium, and particulates). However, because

detailed experimental studies of releases from GT-MHR and PBMR fuels have not been this

approach is bounding. However, gas-cooled reactors operate at much higher temperatures than

LWRs; thus, high-temperature conditions anticipated in transportation accident fires are

expected to have less effect on radionuclide releases than they would for LWR fuels. 

Consequently, smaller release fractions are anticipated for advanced gas-cooled reactor fuels

| than for LWR fuels subjected to thermal transients.

For accidents that result in a release of radioactive material, RADTRAN 5 assumes the material

is dispersed into the environment according to standard Gaussian diffusion models. The code

allows the user to choose two different methods for modeling the atmospheric transport of

radionuclides after a potential accident. The user can input either Pasquill atmospheric-stability

| category data or averaged time-integrated concentrations. In this Section H.2 analysis, the

default standard cloud option (using time-integrated concentrations) was used.

| RADTRAN 5 was used to calculate the population dose from the released radioactive material

for four of five(a) | possible exposure pathways:

| C External dose from exposure to the passing cloud of radioactive material (cloudshine).

| C External dose from radionuclides deposited on the ground by the passing plume

| (groundshine). The Section H.2 analysis included the radiation exposures 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.

| C Internal dose from inhalation of airborne radioactive contaminants (inhalation).

| C Internal dose from radioactive materials that were deposited on the ground and then

| resuspended (resuspension). The Section H.2 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.

A sixth pathway, external doses arising from increased radiation fields surrounding a shipping

cask with damaged shielding, was considered but not included in the analysis. It is possible that
Appendix H

April 2006 H-35 NUREG-1817

shielding materials incorporated into the cask structures could become damaged as a result of

an accident. For example, casks with lead shielding could undergo a slumping phenomenon in

which impact or fire causes gaps to form in the lead. Radiation would penetrate through the

gaps in the shielding at higher intensities, leading to higher radiation dose rates. These events, |

which are commonly referred to as “loss of shielding events,” were not included in this |

assessment because their contribution to spent fuel transportation risks is much smaller than the

dispersal accident risks.

Standard radionuclide uptake and dosimetry models are incorporated into RADTRAN 5. The

computer code combines the accident consequences and frequencies of each severity category,

sums up the severity categories, and then integrates across all the shipments. Accident-risk

impacts are provided in the form of a collective population dose (person-rem over the entire |

shipping campaign).

The shipping distances and population distribution information for the routes used for the

evaluation of the impacts of incident-free transportation (see Table H-6) were also used to

calculate transportation impacts. Representative shipping casks described above were

assumed.

Table H-13 presents unit (per MTU) accident risks associated with transportation of spent fuel

from each potential ESP site to the proposed Yucca Mountain repository. |

Projected annual accident risks, normalized to the WASH-1238 (AEC 1972) reference LWR net

electrical generation (i.e., 880 MW(e)) are presented in Table H-13. As expected, accident risks

are highest for the longest shipments. Also, consistent with past spent fuel transportation risk

assessments, the routine impacts are several orders of magnitude greater than accident

impacts.

Considering the small magnitude of the risks presented in Table H-12 and the conservative |

computational methods and data used to address uncertainties, the overall transportation |

accident risks associated with ABWR, ESBWR, Surrogate AP1000, GT-MHR, and PBMR spent |

fuel shipments are judged to be small. Although likely to also be small, accident risks associated

with IRIS and ACR-700 spent fuel shipments could not be analyzed because of the lack of

radionuclide source-term data. Additional analyses are necessary to quantify the impacts of |

IRIS and ACR-700 spent fuel shipments.
Appendix H

NUREG-1817 H-36 April 2006

Table H-13. Unit Spent Fuel Transportation Accident Risks for Advanced Reactors

Site

Advanced Reactor Type

ABWR/

ESBWR

Surrogate

AP1000 GT-MHR PBMR

Population Dose, person-Sv/MTU(a)

Braidwood(b) 1.0 x 10-7 1.0 x 10-8 1.5 x 10-8 2.5 x 10-8

Clinton 1.1 x 10-7 1.0 x 10-8 1.5 x 10-8 2.6 x 10-8

FitzPatrick 1.9 x 10-7 1.7 x 10-8 2.5 x 10-8 4.3 x 10-8

Grand Gulf(c) 2.0 x 10-7 1.9 x 10-8 2.8 x 10-8 4.7 x 10-8

North Anna 2.3 x 10-7 2.1 x 10-8 3.2 x 10-8 5.4 x 10-8

Pilgrim 4.0 x 10-7 3.7 x 10-8 5.5 x 10-8 9.3 x 10-8

| Portsm outh 2.3 x 10-7 2.1 x 10-8 3.1 x 10-8 5.2 x 10-8

Quad Cities 1.0 x 10-7 9.4 x 10-9 1.4 x 10-8 1.4 x 10-8

Savannah River 2.3 x 10-7 2.4 x 10-8 3.6 x 10-8 6.1 x 10-8

| Surry 2.4 x 10-7 2.2 x 10-8 3.3 x 10-8 5.6 x 10-8

Zion 1.5 x 10-7 1.4 x 10-8 2.1 x 10-8 3.5 x 10-8

(a) To convert to person-rem, multiply person-Sv by 100.

| (b) Dresden and LaSalle can be assumed to be bounded by the Braidwood values because of the proximity of the

sites.

| (c) The River Bend alternative site can be assumed to bounded by the Grand Gulf values because of the proximity

of the sites.

Table H-14 presents the environmental consequences of transportation accidents when shipping

spent fuel from the proposed ESP sites 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. The table presents estimates

of population dose (person-Sv/reference 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). 

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 low dose rates. However, radiation protection experts

conservatively assume that any amount of radiation exposure 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
Appendix H

April 2006 H-37 NUREG-1817

Table H-14. Annual Spent Fuel Transportation Accident Impacts for Advanced Reactors,

Normalized to Reference LWR Net Electrical Generation

MTU/reference reactor year

Advanced Reactor Type

ABWR/ESBWR

Surrogate

AP1000 GT-MHR PBMR

20.3 19.7 6.0 5.8

Population Dose, person-Sv per reference reactor year(a) |

Braidwood(b) 2.1 x 10-6 2.0 x 10-7 8.9 x 10-8 1.5 x 10-7

Clinton 2.3 x 10-5 2.0 x 10-7 9.1 x 10-8 1.5 x 10-7 |

FitzPatrick 3.8 x 10-6 3.3 x 10-7 1.5 x 10-7 2.5 x 10-7

Grand Gulf(c) 4.1 x 10-6 3.7 x 10-7 1.7 x 10-7 2.7 x 10-7

North Anna 4.7 x 10-6 4.2 x 10-7 1.9 x 10-7 3.1 x 10-7

Pilgrim 8.1 x 10-6 7.2 x 10-7 3.3 x 10-7 5.4 x 10-7 |

Portsm outh 4.6 x 10-6 4.0 x 10-7 1.8 x 10-7 3.0 x 10-7

Quad Cities 2.1 x 10-6 1.8 x 10-7 8.4 x 10-8 8.1 x 10-8 |

Savannah River 5.3 x 10-6 4.7 x 10-7 2.2 x 10-7 3.5 x 10-7

Surry 4.9 x 10-6 4.3 x 10-7 2.0 x 10-7 3.2 x 10-7 |

Zion 3.0 x 10-6 2.7 x 10-7 1.2 x 10-7 2.0 x 10-7

(a) Multiply person-Sv/reference reactor year by 100 to obtain doses in person-rem/reference reactor year. |

(b) Dresden and LaSalle can be assumed to be bounded by the Braidwood values because of the proximity of the

sites.

(c) The River Bend alternative site can be assumed to be bounded by the Grand Gulf values because of the

proximity of the sites.

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, the 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

over-estimates 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 ICRP

Publication 60 (ICRP 1991). All the population doses presented in Table H-14 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 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 using the same risk model. |
Appendix H

NUREG-1817 H-38 April 2006

H.2.3 Shipment of Radioactive Waste

This section discusses the environmental effects of transporting radioactive waste from

advanced reactor sites. The environmental conditions listed in 10 CFR 51.52 that apply to

shipments of radioactive waste are as follows:

| C Radioactive waste (except spent fuel) is packaged and in a solid form [51.52(a)(4)]

| C Radioactive waste (except spent fuel) is shipped from the reactor by truck or rail

| [51.52(a)(5)].

INEEL (2003) indicates that all of the advanced reactors will transport their radioactive waste by

truck. Furthermore, INEEL (2003) indicates that all of the advanced reactors plan to solidify and

package their radioactive waste. In addition, all of the advanced reactors will be subject to NRC

| (10 CFR Part 71) and U.S. Department of Transportation regulations for the shipment of

| radioactive material (49 CFR Parts 171, 172, 173, 178).

Table S–4 also specifies the following conditions that apply to shipments of radioactive waste:

| C Weight – less than 33,100 kg (73,000 lb) per truck or 100 tons per cask per rail car

| C Traffic density – less than one truck shipment per day or three rail cars per month.

The advanced reactors are assumed to be capable of shipping their radioactive wastes in

compliance with Federal or State weight restrictions. With respect to the traffic density, all of the

| advanced reactor vendors provided radioactive waste generation estimates. Table H-15

provides these estimates, in addition to the radioactive waste generation estimates for the

reference LWR in WASH-1238 (AEC 1972).

As shown in the table, only the PBMR generates a larger volume of radioactive waste than the

reference LWR in WASH-1238. However, the GT-MHR and PBMR information in INEEL (2003)

| assumed these advanced reactors would ship wastes using two different packaging systems: 

one that hauls 28 m3/shipment (1000 ft3 per shipment) and one that hauls 5.7 m3/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/year per 1100 MW(e) (880 net MW(e)) for the GT-MHR and seven shipments/year

per 1100 MW(e) for the PBMR. These estimates are well below the reference LWR

(42 shipments per 1100 MW(e)). In any event, all the estimates are well below the one truck

shipment per day condition given in 10 CFR 51.52, Table S–4. Doubling the shipment estimates

to account for empty return shipments is still well below the one shipment per day condition.
Appendix H

April 2006 H-39 NUREG-1817

Table H-15. 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(c) 62 27

ESBWR 100 m3/yr

per unit

100 1500(c) 62 27

Surrogate

AP1000

55 m3/yr per

unit

56 1150(c) 45 20

ACR-700 47.5 m3/yr per

unit

95 1462(d) 64 28

IRIS 25 m3/yr

per unit

74 (3 units) 1005(e) 67 29

GT-MHR 98 m3/yr (4-unit

plant)

98 (4 units) 1140(f) 86 37(h)

PBMR 100 drums/yr

per unit

168 (8 units) 1320(g) 118 51(h)

(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). |

(c) The ABWR and ESBWR units include one reactor at 1500 MW(e) and the surrogate AP1000 site includes one |

reactor at 1150 MW(e).

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

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

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

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

(h) INEEL (2003) states 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). |

H.3 Support Information for Radiological Dose Assessment |

|

The staff performed an independent assessment of the radiological impacts of normal operation |

for a new nuclear unit at the Grand Gulf ESP site. Results of this assessment are presented in |

this appendix and are compared to SERI’s results in Section 5.9, “Radiological Impacts of |

Normal Operation,” of Revision 2 of the Environmental Review in SERI’s application |
Appendix H

(a) The maxim ally exposed individual probably does not really exist. It is an imaginary person used to

| ensure that the dose criterial set forth in NRC regulations are met by nuclear power facilities. At this

| facility, the maximally exposed individual is assumed to eat fish and shellfish caught near the point of

| discharge and use the shoreline for activities such as sunbathing or fishing.

NUREG-1817 H-40 April 2006

| (SERI 2005a). This section contains information on (1) dose estimates to the public from liquid

| effluents, (2) dose estimates to the public from gaseous effluents, and (3) dose estimates to the

| biota from both liquid and gaseous effluents.

|

| For comparative purposes with SERI’s estimates, all doses and amounts of radioactive material

| are reported in millirem (mrem) and curies (Ci), respectively.

|

| H.3.1 Dose Estimates to the Public from Liquid Effluents

|

| The staff used the LADTAP II code (Strenge et al. 1986) and input parameters supplied by SERI

| in its environmental report, Revision 2 (SERI 2005a) to estimate doses to the general population

and the maximally exposed individual(a) | from the liquid effluent pathway

|

| H.3.1.1 Scope

|

| The important pathways for determining the dose to the population and to the maximally

| exposed individual from liquid effluents include:

| C Eating fish or invertebrates that are caught in the Mississippi River near the point of

| discharge. The population doses are based on the commercial fish and invertebrate

| catches taken from the Grand Gulf environmental report (MP&L 1973). The annual

| consumption rate of fish and shellfish by the maximally exposed individual is taken from

estimates provided in Regulatory Guide 1.109 (NRC 1977b)

| C External exposure from the surface of contaminated water or from shoreline sediment as

| a result of using the shoreline for activities such as sunbathing or fishing

|

| There are only three public water supply systems in the state of Mississippi that use surface

| water as a source, and none of these are located within 80 km (50 mi) of the Grand Gulf ESP

| site. There are no downstream intakes that use the Mississippi River as a potable water supply

| within 160 km (100 mi) of the Grand Gulf ESP site. For this reason, ingestion of water is not

| considered as a pathway. 

|

| Swimming and recreational boating in the Mississippi River near the GGNS site is very limited,

| and it was assumed for this analysis that no swimming or recreational boating occurs.

|
Appendix H

April 2006 H-41 NUREG-1817

Doses to the maximally exposed individual were calculated for the following: |

 C Total body - Dose was the total for all pathways (i.e., fish and invertebrate consumption |

and shoreline usage) with the highest value for either the adult, teen, child, or infant |

compared to the design objective of 0.03 mSv/yr (3 mrem/yr) per reactor in Title 10 of the |

Code of Federal Regulations (CFR), Part 50, Appendix I. |

|

 C Organ dose - Dose was the total for each organ for all pathways (i.e., fish and |

invertebrate consumption and shoreline usage) with the highest value for either the adult, |

teen, child, or infant compared to the design objective of 0.1 mSv/yr (10 mrem/yr) per |

reactor in 10 CFR Part 50, Appendix I. |

|

The NRC staff reviewed the input parameters used by SERI for appropriateness. Default values |

for input parameters from Regulatory Guide 1.109 were used when site-specific parameters |

were not available. |

|

H.3.1.2 Resources Used

|

The staff used a personal computer version of the LADTAP II code entitled NRCDOSE Version |

2.3.5 (Bland 2000), obtained through the Oak Ridge Radiation Safety Information Computational |

Center (RSICC) to calculate doses to the public from liquid effluents. |

H.3.1.3 Input Parameters

|

Table H-16 provides a listing of the major parameters used in calculating dose to the public from |

liquid effluent releases during normal operation. The values used by SERI and the staff for each |

parameter are listed along with comments regarding the appropriateness of the value. All of the |

input parameters were similar and appropriate. |
Appendix H

NUREG-1817 H-42 April 2006

| Table H-16. Parameters Used in Calculating Dose to the Public from Liquid Effluents

| Parameter SERI Value Staff Value

| Comments

| (Appropriateness

of Value)

Source term (Ci/yr) Table 3.0-8 of SERI (2005a), modified

as discussed in “Comments” column

Table 3.0-8 of SERI (2005a), modified

as discussed in “Comments” column

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

|

| Iodine-131

Iodine-132

Iodine-133

Iodine-134

Iodine-135

Tritium

Carbon-14

Sodium-24

Phosphorus-32

Chromium-51

Manganese-54

Manganese-56

Cobalt-57

Cobalt-58

Cobalt-60

Iron-55

Iron-59

Nickel-63

Copper-64

Zinc-65

Bromine-84

Rubidium-88

Rubidium-89

Strontium-89

Strontium-90

Yttrium-90

Strontium-91

Yttrium-91

Yttrium-91m

Strontium-92

Yttrium-92

Yttrium-93

Zirconium-95

Niobium-95

Molybdenum-99

Technetium-99m

Ruthenium-103

Ruthenium-106

Silver-110m

2.826 x 10-2

5.200 x 10-3

2.000 x 10-2

3.400 x 10-3

1.503 x 10-2

6.200 x 10+3

8.800 x 10-4

5.622 x 10-3

3.600 x 10-4

1.541 x 10-2

5.200 x 10-3

7.622 x 10-3

1.438 x 10-4

6.720 x 10-3

1.822 x 10-2

1.162 x 10-2

4.000 x 10-4

2.800 x 10-4

1.503 x 10-2

8.200 x 10-4

4.000 x 10-5

5.400 x 10-4

8.811 x 10-5

2.200 x 10-4

7.027 x 10-5

6.216 x 10-6

1.800 x 10-3

2.200 x 10-4

2.000 x 10-5

1.600 x 10-3

1.200 x 10-3

1.800 x 10-3

2.080 x 10-3

3.820 x 10-3

1.659 x 10-3

1.600 x 10-3

9.860 x 10-3

1.470 x 10-1

2.100 x 10-3 

Iodine-131

Iodine-132

Iodine-133

Iodine-134

Iodine-135

Tritium

Carbon-14

Sodium-24

Phosphorus-32

Chromium-51

Manganese-54

Manganese-56

Cobalt-57

Cobalt-58

Cobalt-60

Iron-55

Iron-59

Nickel-63

Copper-64

Zinc-65

Bromine-84

Rubidium-88

Rubidium-89

Strontium-89

Strontium-90

Yttrium-90

Srtrontium91

Yttrium-91

Yttrium-91m

Strontium-92

Yttrium-92

Yttrium-93

Zirconium-95

Niobium-95

Molybdenum-99

Technetium-99m

Ruthenium-103

Ruthenium-106

Silver-110m

2.83 x 10

-2

5.20 x 10

-3

2.00 x 10

-2

3.40 x 10

-3

1.50 x 10

-2 |

6.20 x 10

+3

8.80 x 10

-4

5.62 x 10

-3

3.60 x 10

-4

1.54 x 10

-2

5.20 x 10

-3

7.62 x 10

-3

1.44 x 10

-4

6.72 x 10

-3

1.82 x 10

-2

1.16 x 10

-2

4.00 x 10

-4

2.80 x 10

-4

1.50 x 10

-2

8.20 x 10

-4

4.00 x 10

-5

5.40 x 10

-4

8.81 x 10

-5

2.20 x 10

-4

7.03 x 10

-5

6.22 x 10

-6

1.80 x 10

-3

2.20 x 10

-4

2.00 x 10

-5

1.60 x 10

-3

1.20 x 10

-3

1.80 x 10

-3

2.08 x 10

-3

3.82 x 10

-3

1.66 x 10

-3

1.60 x 10

-3

9.86 x 10

-3 |

1.47 x 10

-1 |

2.10 x 10

-3

| The staff’s values

| were rounded to

| three significant

| digits.

Rhodium-106 

(1.47 x 10-1 curies), 

| Rhodium-103m

(9.86 x 10-3 curies),

Silver-110

(2.80 x 10-4 curies),

| Barium-137m

(2.49 x 10-2 curies),

were not included in

the calculation

because they are

not accepted by the

LADTAP II code. 

Their contribution to

the dose is relatively

small.
Appendix H

April 2006 H-43 NUREG-1817

Table H-16. (contd)

Parameter SERI Value Staff Value |

Comments |

(Appropriateness

of Value)

Antimony-124 |

Tellurium-129

Tellurium-129m

Tellurium-131

Tellurium-131m

Tellurium-132

Cesium-134

Cesium-136

Cesium-137

Cesium-138

Barium-140

Lanthanum-140

Cerium-141

Cerium-143

Cerium-144

Praseodymium-143

Praseodymium-144

Tungsten-187

Neptunium-239

1.358 x 10-3

3.000 x 10-4

2.400 x 10-4

6.000 x 10-5

1.800 x 10-4

4.800 x 10-4

1.986 x 10-2

1.260 x 10-3

2.664 x 10-2

3.800 x 10-4

1.104 x 10-2

1.486 x 10-2

2.400 x 10-4

3.800 x 10-4

6.320 x 10-3

2.600 x 10-4

6.320 x 10-3

2.600 x 10-4

6.216 x 10-3

Antimony-124

Tellurium-129

Tellurium-129m

Tellurium-131

Tellurium-131m

Tellurium-132

Cesium-134

Cesium-136

Cesium-137

Cesium-138

Barium-140

Barium-140

Cerium-141

Cerium-143

Cerium-144

Praseodymium-143

Praseodymium-144

Tungsten-187

Neptumium-239 |

1.35 x 10

-3

3.00 x 10

-4

2.40 x 10

-4

6.00 x 10

-5

1.80 x 10

-4

4.80 x 10

-4

1.99 x 10

-2

1.26 x 10

-3

2.66 x 10

-2

3.80 x 10

-4

1.10 x 10

-2

1.49 x 10

-2

2.40 x 10

-4

3.80 x 10

-4

6.32 x 10

-3

2.60 x 10

-4

6.32 x 10

-3

2.60 x 10

-4

6.22 x 10

-3

|

|

|

|

|

|

|

|

|

|

|

|

|

|

Discharge flow rate |

cfs (gpm) |

0.078 (35) 0.078 (35) Site-specific value |

from Table 3.0-1

(SERI 2005a) |

Source term |

multiplier |

1 1 Site specific value

from SERI (2005b) 

Site type Fresh water Fresh water Site specific value |

from SERI (2005b) |

Reconcentration |

model |

No internal reconcentration model

employed

No internal reconcentration model

employed

80-kilometer |

(50-mile) |

population |

3.95 x 10

+5 3.95 x 10+5 Value from |

Tables 2.5-1 and |

2.5-6 (SERI 2005a)

Shore width factor 0.2 0.2 Site-specific value |

based on |

Regulatory Guide |

1.109 (NRC 1977b) |
Appendix H

Table H-16. (contd)

Parameter SERI Value Staff Value |

Comments |

(Appropriateness

of Value)

NUREG-1817 H-44 April 2006

| Dilution factors for

| aquatic food and

| shoreline

|

|

|

|

|

|

| 730 730 Based on a cooling

| tower blowdown rate

| of 808 L/s (12,800

| gpm) (Table 5.4-1 of

| SERI 2005a) and a

dilution factor of 2 in

the Mississippi River

(SERI 2006).

| Consumption and

usage factors for

adults, teens,

| children, and infants

Values from Table 5.4-2 of

environmental report (SERI 2005a)

Values from Table 5.4-2 of

environmental report (SERI 2005a)

| Default values from

| Regulatory Guide

| 1.109

| H.3.1.4 Comparison of Results

| Table H-17 compares the results obtained by SERI for the maximum individual and total

| population dose with those obtained by calculations performed by the staff. The dose results are

essentially the same. 

| Table H-17. Comparison of Dose Estimates to the Public from Liquid Effluent Release per Unit

Type of Dose

SERI’s Environmental

Report

| (SERI 2005a) Staff’s Calculation Percent Difference

Maximum Individual Dose (mrem/yr)(a) |

Total Body (mrem/yr)(b) | 2.17 2.17 0

Organ Dose (bone)(c) |

| (mrem/yr)

| 4.10 4.09 <1

Total Population Dose (person-rem/yr)(d) |

| Total Body 2.06 2.06 0

| Maximum Organ (liver ) 3.32 3.32 0

(a) mrem = millirem; divide mrem/yr by 100 to obtain millisievert/yr

(b) An adult was found to receive the maximum individual total body dose.

(c) A child was found to receive the maximum individual organ dose.

(d) Divide man-rem/yr by 100 to obtain person-sievert/yr.
Appendix H

April 2006 H-45 NUREG-1817

H.3.2 Dose Estimates to the Public from Gaseous Effluents |

The staff used the GASPAR II code (Strenge et al. 1987) and input parameters supplied by |

SERI in its environmental report, Revision 2 (SERI 2005a) to estimate doses to the general |

population within an 80-km (50-mi) radius of the Grand Gulf ESP site and to the maximally |

exposed individual from the gaseous effluent pathways. |

H.3.2.1 Scope |

The staff and SERI calculated annual radiation exposures for the population within a 80-km |

(50-mi) radius of the site and for hypothetical individuals of various ages using the GASPAR II |

code and assuming the following pathways: |

 C Direct radiation from immersion in the gaseous effluent cloud and from particulates |

deposited on the ground |

 C Inhalation of gases and particulates |

 C Ingestion of milk contaminated through the grass-cow-milk pathway |

 C Ingestion of vegetables contaminated by particulates |

 C Ingestion of meat from animals grazing on contaminated pasture. |

Three types of doses were calculated by the staff and compared with SERI’s calculations. |

 C Doses to an individual located at the exclusion area boundary of 0.93 km (0.58 mi) north |

of the site as a result of gamma air dose, beta air dose, total body dose and skin dose |

 C Doses to hypothetical individuals (maximally exposed individual) of various ages that are |

exposed to gaseous radioactive effluents via the pathways listed above |

 C Doses to the population residing within an 80-km (50-mi) radius of the site. |

The NRC staff reviewed the input parameters used by SERI for appropriateness. Default values |

for input parameters from Regulatory Guide 1.109 (NRC1977b) were used when site-specific |

input parameters were not available. The staff concluded that all the input parameters used by |

SERI were appropriate. These parameters were used by the staff in its independent |

calculations. |
Appendix H

NUREG-1817 H-46 April 2006

| H.3.2.2 Resources Used

| The staff used a personal computer version of GASPAR II code entitled NRCDOSE,

| Version 2.3.5 (Bland 2000), obtained through the Oak Ridge RSICC to calculate doses to the

| public from gaseous effluents. 

| H.3.2.3 Input Parameters

| Table H-18 provides a list of the major parameters used in calculating dose to the public from

| gaseous effluent releases during normal operation. The values used by SERI and the staff for

| each parameter are listed along with comments regarding the appropriateness of the value. 

| Table H-18. Parameters Used in Calculating Dose to the Public from Gaseous Effluent

| Releases

| Parameter

SERI Value 

(Table 3.0-7 in SERI 2005a) Staff Value

| Comments 

| (Appropriateness of

Value)

| Source term for

| calculating dose to

| the maximally

| exposed individual

| (curies/year) and

| population within 50-

mile radius

|

|

Krypton-83m 1.68 x 10

-3 Krypton-83m 1.68 x 10-3 The source term is the

bounding plant

paramenter envelope. 

The values are

appropriate.

Krypton-90

(6.49 x 10-4 curies), 

Xenon-139

(8.11 x 10-4 curies), 

| Ruthenium-103m

(2.22 x 10-4 curies)

Rhodium-106

(3.78 x 10-5 curies),

were not included in the

calculation because they

are not accepted by the

GASPAR code. Their

contribution to the total

dose is small.

Krypton-85m 7.20 x 10

+1 Krypton-85m 7.20 x 10+1

Krypton-85 8.20 x 10

+3 Krypton-85 8.20 x 10+3

Krypton-87 5.03 x 10

+1 Krypton-87 5.03 x 10+1

Krypton-88 9.20 x 10

+1 Krypton-88 9.20 x 10+1

Krypton-89 4.81 x 10

+2 Krypton-89 4.81 x 10+2

Xenon-131m 3.60 x 10

+3 Xenon-131m 3.60 x 10+3

Xenon-133m 1.74 x 10

+2 Xenon-133m 1.74 x 10+2

Xenon-133 9.20 x 10

+3 Xenon-133 9.20 x 10+3

Xenon-135m 8.11 x 10

+2 Xenon-135m 8.11 x 10+2

Xenon-135 9.19 x 10

+2 Xenon-135 9.19 x 10+2

| Xenon-137 1.03 x 10

+3 Xenon-137 1.03 x 10+3

| Xenon-138 8.65 x 10

+2 Xenon-138 8.65 x 10+2

| Iodine-131 5.19 x 10

-1 Iodine-131 5.19 x 10-1

| Iodine-132 4.38 x 10

+0 Iodine-132 4.38 x 10+0

| Iodine-133 3.41 x 10

+0 Iodine-133 3.41 x 10+0

| Iodine-134 7.57 x 10

+0 Iodine-134 7.57 x 10+0

| Iodine-135 4.81 x 10

+0 Iodine-135 4.81 x 10+0

| Carbon-14 2.19 x 10

+1 Carbon-14 2.19 x 10+1

| Tritium 7.06 x 10

+3 Tritium 7.06 x 10+3

| Sodium-24 8.11 x 10

-3 Sodium-24 8.11 x 10-3

| Phosphorus-32 1.84 x 10

-3 Phosphorus-32 1.84 x 10-3

| Argon-41 1.02 x 10

+2 Argon-41 1.02 x 10+2

| Chromium-51 7.03 x 10

-2 Chromium-51 7.03 x 10-2

| Manganese-54 1.08 x 10

-2 Manganese-54 1.08 x 10-2

| Manganese-56 7.03 x 10

-3 Manganese-56 7.03 x 10-3
Appendix H

April 2006 H-47 NUREG-1817

Table H-18. (contd) |

|

Parameter |

SERI Value |

(Table 3.0-7 in SERI 2005a) Staff Value

Comments |

(Appropriateness of |

Value)

Iron-55 1.30 x 10 |

-2 Iron-55 1.30 x 10-2

Cobalt-57 2.46 x 10 |

-5 Cobalt-57 2.46 x 10-5

Cobalt-58 6.90 x 10 |

-2 Cobalt-58 6.90 x 10-2

Iron-59 1.62 x 10 |

-3 Iron-59 1.62 x 10-3

Cobalt-60 2.61 x 10 |

-2 Cobalt-60 2.61 x 10-2

Nickel-63 1.30 x 10 |

-5 Nickel-63 1.30 x 10-5

Copper-64 2.00 x 10 |

-2 Copper-64 2.00 x 10-2

Zinc-65 2.22 x 10

-2 Zinc-65 2.22 x 10-2 |

Rubidium-89 8.65 x 10

-5 Rubidium-89 8.65 x 10-5 |

Strontium-89 1.14 x 10

-2 Strontium-89 1.14 x 10-2 |

Strontium-90 3.60 x 10

-3 Strontium-90 3.60 x 10-3 |

Yttrium-90 9.19 x 10

-5 Yttrium-90 9.19 x 10-5 |

Strontium-91 2.00 x 10

-3 Strontium-91 2.00 x 10-3 |

Strontium-92 1.57 x 10

-3 Strontium-92 1.57 x 10-3 |

Yttrium-91 4.81 x 10 |

-4 Yttrium-91 4.81 x 10-4

Yttrium-92 1.24 x 10 |

-3 Yttrium-92 1.24 x 10-3

Yttrium-93 2.22 x 10 |

-3 Yttrium-93 2.22 x 10-3

Zirconium-95 3.19 x 10 |

-3 Zirconium-95 3.19 x 10-3

Niobium-95 1.68 x 10 |

-2 Niobium-95 1.68 x 10-2

Molybdenum-99 1.19 x 10 |

-1 Molybdenum-99 1.19 x 10-1

Technetium-99m 5.95 x 10 |

-4 Technetium-99m 5.95 x 10-4

Ruthenium-103 7.03 x 10 |

-3 Ruthenium-103 7.03 x 10-3

Ruthenium-106 2.34 x 10 |

-4 Ruthenium-106 2.34 x 10-4

Silver-110m 4.00 x 10 |

-6 Silver-110m 4.00 x 10-6

Antimony-124 3.62 x 10

-4 Antimony-124 3.62 x 10-4 |

Antimony-125 1.83 x 10

-4 Antimony-125 1.83 x 10-4 |

Tellurium-129m 4.38 x 10

-4 Tellurium-129m 4.38 x 10-4 |

Tellurium-131m 1.51 x 10

-4 Tellurium-131m 1.51 x 10-4 |

Tellurium-132 3.78 x 10

-5 Tellurium-132 3.78 x 10-5 |

Cesium-134 1.24 x 10

-2 Cesium-134 1.24 x 10-2 |

Cesium-136 1.19 x 10

-3 Cesium-136 1.19 x 10-3 |

Cesium-137 1.89 x 10

-2 Cesium-137 1.89 x 10-2 |

Cesium-138 3.41 x 10

-4 Cesium-138 3.41 x 10-4 |

Barium-140 5.41 x 10

-2 Barium-140 5.41 x 10-2 |

Lanthanum-140 3.62 x 10

-3 Lanthanum-140 3.62 x 10-3 |

Cerium-141 1.84 x 10

-2 Cerium-141 1.84 x 10-2 |

Cerium-144 3.78 x 10

-5 Cerium-144 3.78 x 10-5 |

Praseodymium-144 3.78 x 10

-5 Praseodymium-144 3.78 x 10-5 |

Tungsten-187 3.78 x 10

-4 Tungsten-187 3.78 x 10-4 |

Neptunium-239 2.38 x 10

-2 Neptunium-239 2.38 x 10-2 |

Population

Distribution

Used data from SERI’s supporting

documentation (equivalent to data

found in Tables 2.5-1 and 2.5-6 of

SERI (2005a) for the year 2070)

Used data from SERI’s supporting |

documentation (equivalent to data |

found in Tables 2.5-1 and 2.5-6 of |

SERI (2005a) for the year 2070) |

Site-specific data -

appropriate for use

Atmospheric

dispersion factors

Used data from SERI’s supporting

documentation (equivalent to

Used data from SERI’s supporting

documentation (equivalent to

Site-specific data - |

appropriate for use |
Appendix H

Table H-18. (contd) |

|

Parameter |

SERI Value |

(Table 3.0-7 in SERI 2005a) Staff Value

Comments |

(Appropriateness of |

Value)

NUREG-1817 H-48 April 2006

(sec/m3| ) Table 2.7-118 of SERI (2005a)) Table 2.7-118 of SERI (2005a))

Ground deposition

factors (m— 2)

Used data from SERI’s supporting

documentation (equivalent to

Table 2.7-120 of SERI (2005a))

Used data from SERI’s supporting

documentation (equivalent to

| Table 2.7-120 of SERI (2005a))

| Site-specific data -

| appropriate for use

Milk production rate

within 80 km (50 mi)

(L/yr)

Used data from SERI’s supporting

documentation (equivalent to

Table 5.4-5 of SERI (2005a))

| Used data from SERI’s supporting

| documentation (equivalent to

| Table 5.4-5 of SERI (2005a))

| Site-specific data -

| appropriate for use

Meat production rate

within 80 km (50 mi)

(kg/yr)

Used data from SERI’s supporting

documentation (equivalent to

Table 5.4-6 of SERI (2005a))

Used data from SERI’s supporting

documentation (equivalent to

| Table 5.4-6 of SERI (2005a))

| Site-specific data -

| appropriate for use

| Vegetable production

| rate within 80 km

(50 mi) (kg/yr)

Used data from SERI’s supporting

documentation (equivalent to

| Table 5.4-7 of SERI (2005a))

Used data from SERI’s supporting

documentation (equivalent to

Table 5.4-7 of SERI (2005a))

Site-specific data -

appropriate for use

| Pathway receptor

| locations (direction,

| distance and

| atmospheric

| dispersion factors)

Used data from SERI’s supporting

documentation (equivalent to

Table 2.7-117 of SERI (2005a))

Used data from SERI’s supporting

documentation (equivalent to

Table 2.7-117 of SERI (2005a))

Site-specific data -

appropriate for use

| Nearest site 

boundary

| Nearest vegetable 

garden

| Nearest home

| Nearest milk cow

| Nearest meat cow

| Consumption factors

| for leafy vegetables,

| meat, milk, and

| vegetable/fruit

Table 5.4-4 of SERI (2005a) Table 5.4-4 of SERI (2005a) Appropriate for use -

NRC 1977b

| Fraction of year that

| leafy green

| vegetables are grown

1 1 Appropriate for use

| Fraction of year that

| cows are on pasture

1 1 Appropriate for use

| Fraction of the max

| individual’s vegetable

| intake from their own

| garden

0.76 0.76 Appropriate for use
Appendix H

Table H-18. (contd) |

|

Parameter |

SERI Value |

(Table 3.0-7 in SERI 2005a) Staff Value

Comments |

(Appropriateness of |

Value)

April 2006 H-49 NUREG-1817

Fraction of milk cow |

intake that is from |

pasture while on |

pasture |

1 1 Appropriate for use

Average absolute |

humidity over the |

growing season |

8 8 Appropriate for use -

Default value of

GASPAR II code -

appropriate for use |

Fraction of year goats |

are on pasture |

1 1 Site-specific data -

appropriate for use

Fraction of goat intake |

that is from pasture |

while on pasture |

1 1 Site-specific data -

appropriate for use

Fraction of year beef- |

cattle are on pasture |

1 1 Site-specific data -

appropriate for use

Fraction of beef-cattle |

intake that is from |

pasture while on |

pasture |

1 1 Site-specific data -

appropriate for use

H.3.2.4 Comparison of Dose Estimates to the Public from Gaseous Effluent Releases |

Table H-19 compares results obtained by SERI with those performed by the staff for doses to |

the maximally exposed individual, primarily at the exclusion area boundary. Doses calculated |

were similar.

Table H-19. Comparison of Dose Estimates to the Maximally Exposed Individual from |

Gaseous Pathway Releases

Type of Dose |

SERI’s Environmental |

Report (SERI 2005a) |

Staff’s

Calculation

Percent |

Difference

Gamma air dose at exclusion area boundary (mrad)(a) 1.80 1.80 0 |

Beta air dose at exclusion area boundary (mrad)(a) 3.48 3.48 0 |

Total body dose at exclusion area boundary (plume, |

ground, and inhalation) - (Teen) (mrem)(b) |

1.62 1.69 4.3 |

Skin dose at exclusion area boundary - (Teen) (mrem)(b) 4.42 4.42 0 |

Vegetable consumption at nearest garden |

(Child, thyroid) (mrem)(b) |

6.70 6.70 0 |

(a) mrad = millirad; divide mrad by 100 to obtain milligray/yr. |

(b) mrem = millirem; divide mrem by 100 to obtain millisievert/yr. |
Appendix H

NUREG-1817 H-50 April 2006

| Table H-20 provides the doses to the maximally exposed individual calculated by SERI and the

| staff. Doses to the maximally exposed individual were calculated at the nearest residence,

| nearest garden, nearest site boundary, nearest meat cow, and nearest milk cow. The doses

| calculated by the NRC staff were the same as those calculated by SERI. Thus, only one

number is given for each entry in the table.

| Table H-20. SERI (2005a) and Staff Dose Estimates to the Maximally Exposed Individual from

Gaseous Effluent Releases from Operation of One New Nuclear Unit

| Location Pathway

| Total Body Dose

(mrem/yr)(a)

Skin Dose

(mrem/yr)(a)(b)

| Thyroid Dose

(mrem/yr)(a) |

Nearest Residence(c) |

| (NNE, 1.02 km

(0.64 mi))

|

|

|

|

|

| Plume Exposure 0.63 2.09 0.63

| Inhalation

| Adult 0.171 - 0.69

| Teen 0.173 - 0.85

| Child 0.153 - 0.995

| Infant 0.088 - 0.855

Nearest Garden(c) |

| (ENE, 1.01 km

(0.63 mi))

|

|

|

Vegetable

Consumption

|

| Adult 0.387 - 2.87

| Teen 0.491 - 0.36

| Child 0.901 - 0.67

| Nearest Site

Boundary(d) |

(N, 0.93 km (0.58 mi))

|

|

|

|

| Plume Exposure 1.18 3.88 1.18

| ___Inhalation___ -

| Adult 0.318 - 1.28

| Teen 0.321 - 1.58

| Child 0.285 - 1.84

Infant 0.164 - 1.58

Nearest Milk Cow(c) |

| (S SW 16 km (10 mi))

|

|

|

|

| Cow Milk -

| Adult 0.00565 - 0.055

| Teen 0.00833 - 0.0865

| Child 0.0159 - 0.172

| Infant 0.0287 - 0.409
Appendix H

April 2006 H-51 NUREG-1817

Table H-20. (contd) |

Location Pathway |

Total Body Dose |

(mrem/yr)(a)

Skin Dose

(mrem/yr)(a)(b)

Thyroid Dose |

(mrem/yr)(a) |

Nearest Meat Cow(c) (S, |

6.4 km (4.0 mi))

|

|

|

Meat Consumption |

Adult 0.00653 - 0.0144 |

Teen 0.00472 - 0.0104 |

Child 0.00758 - 0.0163 |

(a) mrem = milirem; divide mrem/yr by 100 to obtain millisievert/yr. |

(b) Skin dose is only applicable to plume exposure. |

(c) “Nearest” refers to the location at which the highest radiation dose to an individual from the |

applicable pathways has been estimated. |

(d) “Nearest” refers to that site boundary location at which the highest radiation doses from gaseous |

emissions have been estimated to occur. |

|

H.3.2.5 Comparison of Results - Population Doses

Table H-21 compares the SERI population dose estimates taken from Table 5.4-13 of SERI

(2005a) with the staff’s estimate. Calculated doses are the same.

Table H-21. Comparison of Population Doses from Gaseous Effluent Releases |

Pathway SERI’s Environmental |

Report (SERI 2005a)

Staff’s Calculation Percent Difference |

Total Body (person-rem/yr)(a) |

Plume 0.157 0.157 0 |

Ground 0.0546 0.0546 0 |

Inhalation 0.418 0.418 0 |

Vegetable ingestion 0.152 0.152 0 |

Cow-milk ingestion 0.215 0.215 0 |

Meat ingestion 0.184 0.184 0 |

Total 1.18 1.18 0 |

Thyroid (Worst Case Organ) (person-rem/yr)(a) |

Plume 0.157 0.157 0 |

Ground 0.0546 0.0546 0 |

Inhalation 1.23 1.23 0 |

Vegetable ingestion 0.154 0.154 0 |

Cow-milk ingestion 0.89 0.890 0 |

Meat ingestion 0.248 0.248 0 |

Total 2.73 2.73 0 |

(a) Divide person-rem /yr by 100 to obtain person-sievert/yr. |
Appendix H

NUREG-1817 H-52 April 2006

H.3.3 Dose Estimates to the Biota from Liquid and Gaseous Effluents

To estimate doses to the biota from the liquid and gaseous effluent pathways, the staff used the

LADTAP II code (Strenge et al. 1986) and the GASPAR II code (Strenge et al. 1987) and input

| parameters supplied by SERI as part of its environmental report (SERI 2005a).

H.3.3.1 Scope

| Doses from the liquid pathways to both terrestrial and aquatic biota were calculated using the

| LADTAP II code. Aquatic biota include fish, invertebrates, and algae. Terrestrial biota include

| muskrat, raccoon, heron, and duck. The LADTAP II code calculates the biota dose from the

| liquid effluent pathway by calculating an internal dose component and an external dose

| component and summing them for a total dose. The NRC staff reviewed the input parameters

| used by SERI for appropriateness. Default values from Regulatory Guide 1.109 (NRC 1977b)

| were used when input parameters were not available. The staff used the same parameters in

its independent calculations using LADTAP.

| Terrestrial biota could also be exposed via the gaseous effluent pathway. An estimate of these

| values was made by using the dose from exposure to the ground and multiplying it by a factor

| that adjusts for the size of the animal and their distance from the ground. This is added to the

dose from the plume to obtain the external dose. Internal dose to terrestrial biota is based on

the total body inhalation dose for the maximally exposed individual (infant) at the site boundary

| calculated by GASPAR II. The total body inhalation dose (rather than organ specific doses) is

| used since the biota doses are assessed on a total body basis.

| H.3.3.2 Resources Used

| To calculate the doses to the public from liquid releases, the staff used a computer code entitled,

| NRCDOSE, version 2.3.5 (Bland 2000) which is a version of the LADTAP II code and the

| GASPAR II code, obtained through the Oak Ridge RSICC.

| H.3.3.3 Input Parameters

| The LADTAP II parameters are specified in Table H-16 and include the source term, discharge

| flow rate, reconcentration model, effluent discharge rate to the Mississippi River, impoundment

| total volume, and shore width factor. Parameters unique to the biota dose calculation were

| taken from Table 5.4-14 (terrestrial biota parameters including food intake, body mass and

| effective body radius) and Table 5.4-15 of the environmental report (SERI 2005a) (shoreline

| exposure and swimming exposure estimates). These parameters were default values used in

| the LADTAP II code (Strenge et al. 1986) and are appropriate values to use in calculating

| biota dose.
Appendix H

April 2006 H-53 NUREG-1817

H.3.3.4 Comparison of Results |

Table H-22 compares the dose results obtained by SERI (2005a, Table 5.4-16) with those |

performed by the staff for liquid effluents. The dose estimates were the same. |

Table H-22. Comparison of Dose Estimates to Biota from Liquid Effluents |

Biota Type of Dose 

SERI’s

Environmental

Report (SERI

2005a) (mrad/yr)(a)

Staff’s Calculation |

(mrad/yr)(a) Percent Difference |

Fish Internal 14.2 14.2 0 |

External 11.2 11.2 0 |

Invertebrates Internal 143 143 0 |

External 22.3 22.3 0 |

Algae Internal 148 148 0 |

External 0.05 0.05 0 |

Muskrat Internal 73.8 73.8 0 |

External 7.45 7.45 0 |

Raccoon Internal 13.4 13.4 0 |

External 5.57 5.57 0 |

Heron Internal 186 186 0 |

External 7.44 7.44 0 |

Duck Internal 69.9 69.9 0 |

External 11.2 11.2 0 |

(a) mrad = millirad; divide mrad/yr by 100 to obtain milligray/yr.

Table H-23 compares the dose results obtained by SERI (2005a, Table 5.4-16) with those |

performed by the staff for gaseous effluents. These dose estimates were similar. |

Table H-23. Comparison of Dose Estimates to Biota from Gaseous Effluents |

Biota Type of Dose |

SERI’s

Environmental

Report (SERI

2005a) (mrad/yr)(a)

Staff’s Calculation

(mrad/yr)(a) Percent Difference |

Muskrat Internal 0.164 0.164 0 |

External 2.03 2.02 < 1 |

Raccoon Internal 0.164 0.164 0 |

External 1.82 1.81 < 1 |

Heron Internal 0.164 0.164 0 |

External 1.69 1.68 <1 |

Duck Internal 0.164 0.164 0 |

External 2.14 2.13 < 1 |

(a) mrad = millirad; divide mrad/yr by 100 to obtain milligray/yr.
Appendix H

NUREG-1817 H-54 April 2006

H.4 References

| 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 71. Code of Federal Regulations,. Title 10, Energy, “Packaging and Transportation

| of Radioactive Material.”

| 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 Packagings.”

| 49 CFR Part 178. Code of Federal Regulations, Title 49, Transportation, Part 178,

| “Specifications for Packagings.”

| Bland, J.S. 2000. NRC DOSE for Windows. Radiation Safety Information Computational

| Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

| Griego, N.R., J.D. Smith, and K.S. Neuhauser. 1996. “Investigation of RADTRAN Stop Model

| Input Parameters for Truck Stops,” in Conference Proceedings Waste Management 96,

| CONF-960212-44. Tucson, Arizona.

Hostick, C.J., J.C. Lavender, and B.H. Wakeman. 1992. Time/Motion Observations and Dose

Analysis of Reactor Loading, Transportation, and Dry Unloading of an Overweight Truck Spent

Fuel Shipment. PNL-7206, Pacific Northwest Laboratory, Richland, Washington.

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. 
Appendix H

April 2006 H-55 NUREG-1817

International Commission on Radiological Protection (ICRP). 1991. 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. Accessed on the Internet February 7, 2005 at |

http://www.ornl.gov/~webworks/cpr/v823/rpt/106749.pdf. |

|

Mississippi Power and Light Company (MP&L). 1973. Grand Gulf Nuclear Station Units 1 and 2 |

Final Environmental Report (FER), as amended through Amendment No. 8. Jackson, |

Mississippi. |

|

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. Accessed on the

Internet February 7, 2005 at http://infoserve.sandia.gov/sand_doc/2003/032354.pdf. |

Nuclear Waste Policy Act of 1982, as amended. 42 USC 10101, et seq. |

Saricks, C.L. and M.M. Tompkins. 1999. State-Level Accident Rates of Surface Freight

Transportation: A Reexamination. ANL/ESD/TM-150, Argonne National Laboratory,

Argonne, Illinois. 

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.

Strenge, D.L., R.A. Peloquin, and G. Whelan. 1986. LADTAP II - Technical Reference and |

User Guide. NUREG/CR-4013, Pacific Northwest Laboratory, Richland, Washington. |

Strenge, D.L., T. Bander, and J.K. Soldat. 1987. GASPAR II - Technical Reference and User |

Guide. NUREG/CR-4653, Pacific Northwest Laboratory, Richland, Washington. |

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. |
Appendix H

NUREG-1817 H-56 April 2006

| System Energy Resources, Inc. (SERI). 2005b. Letter dated November 4, 2005 from G.A Zinke

| to Nuclear Regulatory Commission, “Response to Request for Additional Information Regarding

the Grand Gulf Early Site Permit Draft Environmental Impact Statement.” Available at

| http://www.nrc.gov/reading-rm/adams.html, Accession No. ML060190548.

| System Energy Resources, Inc. (SERI). 2006. Letter dated January 23, 2006 from G.A. Zinke

| to Nuclear Regulatory Commission, “Response to Request for Additional Information Regarding

| the Grand Gulf Early Site Permit Draft Environmental Impact Statement (TAC MC 1379).” 

| Available at http://www.nrc.gov/reading-rm/adams.html, Accession No. ML060250265.

U.S. Atomic Energy Commission (AEC). 1972. Environmental Survey of Transportation of

Radioactive Materials to and from Nuclear Power Plants. WASH-1238, Washington, D.C.

U.S. Department of Energy (DOE). 2002a. A Resource Handbook on DOE Transportation Risk

Assessment. DOE/EM/NTP/HB-01, Washington, D.C.

U.S. Department of Energy (DOE). 2002b. Final Environmental Impact Statement for a

Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste

at Yucca Mountain, Nye County, Nevada. DOE/EIS-0250, Office of Civilian Radioactive Waste

Management, Washington, D.C.

U.S. Environmental Protection Agency (EPA). 2002. Cancer Risk Coefficients for

Environmental Exposure to Radionuclides. Federal Guidance Report No. 13, Washington, D.C.

| U.S. Nuclear Regulatory Commission (NRC). 1977a. Final Environmental Statement on

Transportation of Radioactive Material by Air and Other Modes. NUREG-0170, Vol.1,

Washington, D.C.

| U.S. Nuclear Regulatory Commission (NRC). 1977b. Regulatory Guide 1.109 - Calculation of

| Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating

| Compliance with 10 CFR Part 50, Appendix I. Revision 1, Washington, D.C.

U.S. Nuclear Regulatory Commission (NRC). 1999. Generic Environmental Impact Statement

for License Renewal of Nuclear Plants, Main Report, Section 6.3 - Transportation, Table 9.1

Summary of Findings on NEPA Issues for License Renewal of Nuclear Power Plants.

| NUREG-1437, Volume 1, Addendum 1, Washington, D.C.
Appendix I

Plant Parameter Envelope Values

April 2006 I-1 NUREG-1817

Appendix I

Plant Parameter Envelope Values

This appendix contains the System Energy Resources, Inc. plant parameter envelope for the |

proposed Grand Gulf early site permit site as submitted in System Energy Resources, Inc. |

environmental report (SERI 2005) as Tables 3.0-1 to 3.0-9 and reproduced here as Table I-1. |

Table I-1. PPE for the Grand Gulf Early Site Permit Facility
Appendix I

NUREG-1817 I-2 April 2006
Appendix I

April 2006 I-3 NUREG-1817
Appendix I

NUREG-1817 I-4 April 2006
Appendix I

April 2006 I-5 NUREG-1817
Appendix I

NUREG-1817 I-6 April 2006
Appendix I

April 2006 I-7 NUREG-1817
Appendix I

NUREG-1817 I-8 April 2006
Appendix I

April 2006 I-9 NUREG-1817
Appendix I

NUREG-1817 I-10 April 2006
Appendix I

April 2006 I-11 NUREG-1817
Appendix I

NUREG-1817 I-12 April 2006
Appendix I

April 2006 I-13 NUREG-1817
Appendix I

NUREG-1817 I-14 April 2006
Appendix I

April 2006 I-15 NUREG-1817
Appendix I

NUREG-1817 I-16 April 2006
Appendix J

System Energy Resources, Inc. Commitments |

and NRC Staff Assumptions Relevant to the Analysis of Impact |

(a) The listings are not intended to be a complete list of the commitments described in the SERI

environm ental report.

April 2006 J-1 NUREG-1817

Appendix J

System Energy Resources, Inc. Commitments and |

NRC Staff Assumptions Relevant to the Analysis of Impact

Throughout the environmental report supporting the Grand Gulf Early Site Permit (ESP)

application (SERI 2005), System Energy Resources, Inc. (SERI) provides: |

(1) Commitments to address certain issues in the design, construction, and operation of the

facility

(2) Statements of planned compliance with current laws, regulations, and requirements

(3) Commitments to future activities and actions that it will take should it be granted an ESP |

and decide to apply for a construction permit (CP) or combined license (COL) |

(4) Descriptions of SERI’s estimate of the environmental impacts resulting from the

construction and operation of a new nuclear unit or units on the Grand Gulf ESP site

(5) Descriptions of SERI’s estimates of future activities and actions of others and the likely

environmental impacts of those activities and actions that would be expected should it be |

granted an ESP and decide to apply for a CP or COL. |

Those statements are discussed throughout this environmental impact statement (EIS) and are

listed in this Appendix.(a) Some of those statements considered by the staff in determining the

level of impacts to a resource are related to matters that are within SERI’s control. Table J-1

lists those matters that were considered in the staff’s evaluation of the environmental impacts

related to the construction and operation of a new nuclear unit or units at the Grand Gulf ESP

site. The table shows the technical area where the matter is addressed in the EIS, and SERI’s

statement that addresses the matter. Table J-2 lists assumptions related to likely activities and |

actions of others that were considered by the staff.

In some cases, the same statement or similar statements are made in more than one place in

the environmental report. Where statements contain essentially the same information, the

location of the more comprehensive statements are listed first in the table, and the text provided

is the text from that location. 
Appendix J

NUREG-1817 J-2 April 2006

| Table J-1. Statements Made in the SERI Environmental Report and in Response to NRC

| Staff Requests for Additional Information Related to Future Actions and

| Activities by SERI and the Impacts of Those Activities Considered in the Staff’s

Analysis

| Technical Area Environmental Report or RAI Statement

| Land Use The Universal Transverse Mercator Grid Coordinates for the location of the

new reactor(s) on the site are approximately N3,543,261 meters and

E684,018 meters.

| Land Use

|

| There is no rail service in the vicinity of the Grand Gulf Nuclear Station

(GGNS) site and there are no active railroad tracks that traverse the GGNS

site or the vicinity surrounding the site.

| Land Use Entergy Operations allows access to parts of the plant site property for

recreational purposes. The site is posted to ensure awareness of access

restrictions by individuals.

| Land Use There is no activity at the GGNS plant site to explore for, drill for, or otherwise

extract minerals. Past unsuccessful exploratory activities on or near the

GGNS plant site and the geological character of the subsurface structure in the

vicinity of the GGNS plant site indicate that comm ercial mineral production

appears unlikely in the foreseeable future.

| Land Use Information from the Claiborne County Extension office at the present tim e

indicated that there are approximately 300 to 400 head of cattle within a 6 m ile

radius of the site, and m ost of the cattle are located southwest of the plant.

There are no milk cows or swine within Claiborne County.

| Land Use Dredging would be required to form the embayment. The embayment bottom

would be at approxim ately elevation 15 ft msl.

| Land Use There are three transmission lines associated with GGNS: (1) the BaxterW ilson line, a 22-mile single-circuit 500 kV transmission line connecting GGNS

to the Baxter-Wilson EHV Substation near Vicksburg, Mississippi; (2) the

Franklin line, a 43.6-mile single-circuit 500 kV transmission line connecting the

GGNS switchyard to the Franklin EHV Substation; and, (3) the Port Gibson

line, a 5.5-mile single-circuit 115 kV transmission line connecting the GGNS

switchyard to the Port Gibson Substation. The electrical power generated by

| GGNS Unit 1 is transmitted by interconnection with 500 kV transmission

| facilities that were in existence when Unit 1 was constructed.
Appendix J

April 2006 J-3 NUREG-1817

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

Land Use The power transmission and distribution (T&D) system existing at the time of

the new facility startup and operation will be relied upon to distribute the

electricity generated by a new facility at Grand Gulf. In support of site selection

evaluation work (environmental report, Section 9.3), a sensitivity analysis of the |

T&D system was performed to assess transmission injection capability for the

new potential electrical power generation at GGNS. This study concluded that

the existing T&D system is adequate for at least an additional 1311 MW (e) of

generating capacity, provided that certain modifications were accomplished. 

Land Use W hen the specific facility design, the expected electrical output, the need for |

power, and primary market location(s) are established, the adequacy of the

existing (at that time) T&D system to support the new facility will be

determined. If, at that time, additional changes to the T&D system were

warranted, the associated environmental impacts would be evaluated.

Land Use An estimated 400 acres of the 2100-acre GGNS site would be affected by

construction of a new facility.

Land Use Of the approxim ately 400 acres estimated for the construction of a new facility, |

approximately 120 acres overlap currently developed or previously altered

areas. It is estimated that approximately 125 acres would contain permanent

structures (primarily a power block area, cooling tower area, and bottom land

pipeline and intake areas.

Land Use The barge slip constructed for GGNS Unit 1 would be used to offload large |

equipment and materials for the construction of a new facility transported by

river.

Land Use There would be some impact from excavation and construction of the intake |

structure along the river bank in the flood plain areas, but the impact is

expected to be sm all and temporary. Additionally, trenching from the intake to

the proposed power block location on the bluffs east of the river would be

required to lay supply and discharge piping from the new facility. Most of the

floodplain areas are also classified as wetlands.
Appendix J

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

NUREG-1817 J-4 April 2006

| Land Use Makeup water (cooling tower makeup and other raw water needs) for a new

facility would be supplied primarily from the Mississippi River via an

embayment, and associated intake structure, located on the east bank of the

river and on the north side of the existing barge slip. Dredging would be

required to form the embayment on the Mississippi River... Riprap, or other

appropriate means, would be used to stabilize the banks of the embayment

and the river shoreline around the embayment during and following

construction. These construction activities would be done in com pliance with

Corps of Engineer requirements...

| Land Use The proposed outfall, located above normal river water level, would include a

concrete drainage course to the river sim ilar to that for the GGNS Unit 1

discharge structure.

| Land Use It is anticipated that the existing road system would be adequate for

construction of a new facility, and new road construction would not be

necessary.

| Land Use Use of Ham ilton and Gin Lakes for recreational fishing may be temporarily

restricted during construction as a safety measure to protect members of the

public from hazards related to the use of heavy construction equipm ent. 

Therefore, the impact to recreational users of these lakes would be m inim al.

| Land Use Approximately 145 acres of upland forest and approximately 105 acres of

upland fields would be affected by the construction of a new facility

(Figure 2.4-3). This represents approximately 35% and 66% of these habitat

types within the GGNS site, respectively... Approximately 100 acres of the

upland area of the GGNS site would be permanently altered (i.e., for

structures, parking lots, etc.) for a new facility. The remaining acreage

disturbed by construction would be revegetated or reseeded and allowed to

develop back into a stable ecological community.

| Land Use Approximately 30 acres of bottomland palustrine, forested, seasonally flooded

wetland would be disturbed during the construction of a new facility

(Figure 2.4-3). This is approximately 3% of this habitat type within the GGNS

site property. The rem ainder of the area required for construction would be in

areas previously disturbed for the construction of GGNS Unit 1 (e.g., heavy

haul road, barge slip area).

| Land Use Additional re-routing of onsite drainages and construction of additional

sediment retention basins would likely be required to support construction of a

new facility. A buffer zone of native vegetation could be maintained between

the construction areas and the lakes.
Appendix J

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

April 2006 J-5 NUREG-1817

Land Use Operation of a new facility is not expected to produce any additional significant |

impacts to land use on the site nor in the vicinity of the GGNS site... These

recreational areas may experience increased visitation due to the operational

work force at a new facility. No other impacts to these recreational facilities

would be expected.

Land Use The bounding estimate of salt deposition from the operation of cooling towers |

would be approximately 8 lbm/100-acre-month ... This amount of deposition

would not be expected to cause damage to vegetation in the vicinity of the

GGNS site. Therefore, no significant impact to land use from cooling tower

drift is expected on the site. And, based on proposed cooling tower(s) distance

from the site boundary, and the prevailing wind direction, none is expected

beyond the site boundaries.

Land Use The rail line, which extended from Vicksburg to the site and beyond, and the |

spur constructed to the site to support GGNS Unit 1 construction, have since

been abandoned. To support transport of heavy materials and equipment to

the site, new rail service will likely be required. This may involve reconstruction

of rail tracks along the former rights of way, or construction of new rail lines.

Land Use |

(response to 

RAI 4.1-1)

The statement quoted from the Environmental Report is intended to identify |

that the precise m ethods of construction material transportation to the site

have not been determined or projected and that rail service is not immediately

available at the site. The Environmental Report does not propose, project or

evaluate possible changes to rail service. Many variables could affect potential

future construction m aterial transportation modes, including the degree to

which modular construction methods are to be used. Although not evaluated

for the ESP, the Environmental Report does not preclude future consideration

and evaluation of rail service.

Land Use

(response to RAI

2.5.3-1)

NRC staff considered the aerial photos of the GGNS construction in the |

evaluation of ESP facility construction impacts. |

Land Use Additional analysis would be necessary to confirm whether, beyond the |

addition of 1311 MW (e), any supporting T&D system upgrades or changes

would be required, and what the associated operational environmental impacts

would be. This additional analysis was not pursued at ESP.
Appendix J

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

NUREG-1817 J-6 April 2006

| Land Use

|

| The occurrence of icing conditions even in the vicinity of the linear mechanical

| draft cooling towers (LMDCTs) is expected to be rare since the water

deposition rate is small and prolonged periods with below freezing

temperatures are infrequent. Because any icing would be confined within the

site property boundary, no adverse impact on surrounding public lands or

roadways would occur.

| Land Use

|

|

| Based on the results of the evaluation performed for this application, the

| guidance provided in NUREG-1555, and the results of the Cooling Tower Drift

Program performed for the existing GGNS facility, no adverse impact on the

surrounding vegetation from salt deposition due to the operation of the NHS

cooling towers for the new facility is anticipated.

| Land Use The majority of in-migrants and their families would be expected to settle in

developed, more populous areas, or their suburbs, such as Vicksburg (Warren

County), Natchez (Adams County), and Clinton/Jackson (Hinds County), which

have a combined year 2000 population of over 300,000 people.

| Land Use

|

| The tem porary outage staff typically stays in area hotels or recreational vehicle

courts dispersed throughout the region; therefore, no single com munity would

be overburdened by the influx of temporary workers. It is expected that the

increased frequency of the temporary outage staff would not significantly

impact the region.

| Land Use

|

| Em pirical case studies of seven operating nuclear power plants indicated in all

instances that the in-migration of plant personnel had small impacts on

housing. In addition, the workers would not move exclusively to one

community but rather would be expected to m ake residences in the relatively

large area formed by surrounding comm unities.

| Land Use It is possible that the influx of site workers would increase demand for and

stimulate the development of some commercial businesses (e.g., gasoline and

autom otive service stations, restaurants, etc). However, these services would

likely be confined to existing commuter routes, and would not represent a

major land use change for the region.

Meteorology and Air

Quality

| The normal plant heat sink (NHS) that will be used to dissipate heat from the

turbine cycle for the new facility will utilize cooling towers to dissipate the heat

directly to the atmosphere.

Meteorology and Air

Quality

The cycles of concentration for the NHS circulating water is expected to be a

maximum of 4, which will result in the concentrations in the circulating water

being 4 times that of river water.
Appendix J

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

April 2006 J-7 NUREG-1817

Meteorology and Air |

Quality

Seasonal and Annual Cooling Tower Impact (SACTI) model predicts that the |

majority of the fogging due to the operation of the LMDCTs will be confined to

within about 1⁄2 m ile (800 m) to the south to southeast of the towers with

occasional fogging (approximately 2 hrs/yr) up to about 3⁄4 mile (1200 m) to the

south to southeast of the towers (this area is entirely within the property

boundary of the site). Therefore, it is predicted that the operation of the

LMDCTs will result in limited increased fogging at the site.

Meteorology and Air

Quality

The towers will use drift eliminators to minimize the amount of water lost from

the towers via drift.

Meteorology and Air

Quality

Gaseous emissions will be within regulatory guidelines set by Federal and

State agencies.

Meteorology and Air

Quality

The meteorological monitoring program will be the same throughout the preconstruction and operational phases of the project. The monitoring program

will simply be a continuation of the ongoing meteorological monitoring program

for the GGNS Unit 1 facility.

Ecology It will be required to coordinate with the Corps of Engineers and/or other

appropriate regulatory agencies and obtain permits for construction of the

embayment and intake structure when the final design of the intake structure

and its exact location are defined. The design and placement of the

embayment and intake structure will be in accordance with the Corps

guidance, MDEQ and EPA requirements, and good engineering practice.

Ecology The normal heat sink circulating water system for the new facility will be a

closed-cycle type system using either hyperbolic natural draft cooling towers or

mechanical draft cooling towers.

Ecology The design and placement of the em bayment and intake structure will be in

accordance with the Corps of Engineers guidance, MDEQ and EPA

requirements, and good engineering practice.

Ecology The Corps of Engineers has completed revetments along the east and west

river banks... It is expected that these measures will stabilize the Mississippi

River shoreline near the site.

Ecology This portion of the switchyard would be used, with modifications.

Ecology Plant makeup (cooling tower makeup and other raw water needs) for a new

facility would be supplied from the Mississippi River via an intake structure

located on the east bank of the river.
Appendix J

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

NUREG-1817 J-8 April 2006

Ecology The Corps of Engineers continues to evaluate the need for additional shoreline

work, and would be expected to make improvements as considered

appropriate. However, those actions would not be expected to impact site

suitability.

Ecology Makeup to the normal heat sink cooling towers, balance of plant cooling

systems (e.g., plant service water), and other raw water makeup needs for a

new facility would be supplied by an intake structure located on the east bank

of the Mississippi River.

Ecology The new facility owner would be required to coordinate with the Corps of

Engineers and obtain permits from appropriate regulatory agencies for

construction of the embayment and intake structure when the final design of

the embayment and intake structure and its exact location are defined.

Ecology Eagles nesting on site would be largely protected from shooting, development

and habitat alteration, and other human disturbance that usually accounts for

mortality and reduced breeding success elsewhere.

Ecology Other than the installation of additional revetments along the east bank, no

significant changes to the river channel or banks which would be expected to

alter the ecological characteristics of this riparian habitat have occurred.

Ecology Makeup water to the cooling tower(s) and supply or makeup water for the SWS

will be withdrawn directly from the Mississippi River through an intake structure

on the river shore.

Ecology The power transmission and distribution (T&D) system existing at the time of

the new facility startup and operation will be relied upon to distribute the

electricity generated by a new facility at Grand Gulf.

Ecology W hen the specific facility design, the expected electrical output, the need for

power, and primary market location(s) are established, the adequacy of the

existing (at that time) T&D system to support the new facility will be

determined.

Ecology Construction activities to be conducted within a floodplain on the site would be

the water intake structure and embayment along with other items that are a

part of that water intake facility. This water intake will be located at or near the

existing barge slip area.

Ecology Once the facility design is finalized, appropriate analyses of transmission and

distribution system adequacy will be made.
Appendix J

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

April 2006 J-9 NUREG-1817

Ecology Traffic on Grand Gulf Road will increase substantially during the peak

construction period, and will be at its peak during the morning and evening

shift changes. Noise in the general area will increase from this increased

traffic but the increases will be temporary, and will only occur as indicated

twice per day, during the week.

Ecology The new facility will require a small amount of water withdrawal relative to

normal river flow; makeup flow requirem ents are estimated at approximately

85,000 gpm.

Ecology

|

There is little potential that operation of the cooling system intake for a new

facility at the Grand Gulf ESP site will impact any such areas (wildlife). |

Ecology The Normal Plant Heat Sink (NHS) that will be used to dissipate heat from the

turbine cycle for the new facility will utilize cooling towers to dissipate the heat

directly to the atmosphere.

Ecology The heat dissipation system for the NHS for the new facility will use either

natural draft cooling towers or linear mechanical draft cooling towers. |

Ecology Two types of cooling systems will be considered for a new facility at the Grand |

Gulf ESP site: natural draft cooling towers and mechanical draft cooling |

towers.

Ecology Environmental measurements and monitoring of terrestrial and aquatic ecology |

at the GGNS site will be divided into four phases: |

 • Pre-application (CP or COL) Monitoring • Site Preparation and Construction |

Monitoring • Pre-operational Monitoring • Operational Monitoring

Ecology The Grand Gulf ESP site will not be substantially different from the acceptable |

environmental impacts identified for the previously analyzed sites. |

Ecology (Coal) Additional ecological impact will occur due to land use related to mining

of coal and limestone. Substantially greater impacts expected, relative to that

required for uranium mining and reprocessing. 

Ecology (Combined Cycle Natural Gas) Additional ecological impact will occur due to |

land use related to gas wells and collection stations; expected to be

proportionally higher than that related to uranium m ining and reprocessing.

W ater Use and |

Quality |

Plant makeup (cooling tower makeup and other raw water needs) for a new

facility would be supplied from the Mississippi River via an intake structure

located on the east bank of the river.
Appendix J

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

NUREG-1817 J-10 April 2006

| W ater Use and

| Quality 

| Emergency cooling water (ultimate heat sink) for a new facility would be

provided from closed-cooling system s which utilizes enclosed basins with

mechanical draft cooling towers, or similar heat removal mechanisms, and

would not be reliant on the source of water from the river intake, with the

possible exception of normal make-up.

Socioeconomics Em ergency planning responsibilities are assigned to a number of departm ents

and agencies. Federal, state and local officials will implem ent appropriate

protective actions in case of an em ergency.

Socioeconomics A highway construction plan to extend the present path of Highway 18 is in the

early planning stages. This proposed extension will connect Highway 18 to

Grand Gulf Road, providing additional access to the GGNS site.

| Socioeconomics

|

|

| Depending on the type of plant (merchant plant which would be unregulated, or

a regulated – by the Public Service Commissions of Mississippi and Louisiana

plant), the tax structure may be similar to the GGNS Unit 1 (for a regulated

plant), or be some mutually agreeable amount for an unregulated merchant

plant.

| Socioeconomics The actual mode of shipment [of irradiated fuel] will be determined by DOE and

may include either rail or truck shipments.

Socioeconomics

|

|

|

Construction of the cooling towers will have minimal impact on the

surroundings. Construction noise levels during construction of a new facility at

the Grand Gulf ESP site will have minimal impacts on the surrounding

populace.

Socioeconomics Complying with applicable OSHA noise regulations will ensure that the impact

on construction workers is considered to be small.

| Socioeconomics

|

|

|

| A construction noise abatement and protection program will provide required

mitigative measures for noise which may, on a short term basis, exceed

| guidance [65dB(A)]. Excessively loud construction activities would be done

| during daylight hours if necessary.

| Socioeconomics Traffic on Grand Gulf Road will increase substantially during the peak

construction period, and will be at its peak during the morning and evening

shift changes. Noise in the general area will increase from this increased

traffic but the increases will be temporary, and will only occur as indicated

twice per day, during the week.

Socioeconomics Many of the short-term em ployees will likely travel to the area unaccompanied

by family mem bers.
Appendix J

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

April 2006 J-11 NUREG-1817

Socioeconomics Rural setting of the site and the premise that the majority of the work force will

emanate from the surrounding more populated areas and comm unities away

from the site, it is likely a large portion of these new business and jobs would

be tem porary.

Socioeconomics U.S. 61 S is two-lane improved roadway - will be 4-lane, divided freeway within

2 years like U.S. 61 N from Port Gibson

Socioeconomics SACTI model predicts that the majority of the fogging due to the operation of |

the LMDCTs will be confined to within about 1⁄2 m ile (800 m) to the south to

southeast of the towers with occasional fogging (approximately 2 hrs/yr) up to

about 3⁄4 mile (1200 m) to occasional fogging (approximately 2 hrs/yr) up to

about 3⁄4 mile (1200 m ) to the south to southeast of the towers (this area is

entirely within the property boundary of the site). Therefore, it is predicted that

the operation of the LMDCTs will result in limited increased fogging at the site.

Socioeconomics W hile the proposed project’s workforce and construction time period are

greater than that of the gas plant, the impacts will be short term and m itigated

by dispersion over several relatively populous counties and improved

transportation routes.

Socioeconomics Facility workforce will add to road network traffic load with an associated |

increase in traffic accidents. Road improvements and flexible work schedules

will mitigate this impact to a certain extent.

Socioeconomics Several road improvement and construction projects have been accomplished |

or planned for GGNS area. These projects will help ameliorate traffic

problem s associated with the proposed new facility.

Human Health Liquid radwaste system design will be such that water which is discharged to

the environment shall result in radioactive releases which conform to the “as

low as reasonably achievable” requirements of 10 CFR 50.34a.

Human Health Gaseous radwaste system design, including ventilation systems exhaust

systems, will be such that radioactive gases which are discharged to the

environment from these systems shall result in radioactive releases which

conform to the “as low as reasonably achievable” requirements of

10 CFR 50.34a.

Human Health The LW R technologies being considered will solidify and package their

radioactive waste.

Human Health In all likelihood, the decay time will be at least ten years and probably even

longer.
Appendix J

Table J-1. (contd)

Technical Area Environmental Report or RAI Statement |

NUREG-1817 J-12 April 2006

| Human Health The actual mode of shipment of spent fuel will be determined by DOE and may

include either rail or truck shipments.

Human Health The gas-cooled technologies being considered will solidify and package their

radioactive waste.

| Human Health The gas-cooled reactor technologies will make far fewer shipments. The

GT-MHR will need only 6 shipm ents while the PBMR will require 9 shipments

annually.

Human Health In the case of decay heat, both of the gas-cooled reactor technologies will

generate fewer watts per MTU at time of shipment, and fewer kW per truck

cask at time of shipment. The fuel inventory will be discussed as part of the

remaining two characteristics that were exceeded: actinide inventory and

krypton-85 inventory.

| Human Health Location of a new facility will be several hundred feet or more away from the

protected area boundary, and about 1000 feet from the Unit 1 Turbine Building,

the radiation levels due to nitrogen-16 skyshine are expected to be essentially

background levels, similar skyshine are expected to be essentially background

levels, similar to those readings obtained at TLDs located on the

west/northwest side of the plant protected area boundary.

Human Health These areas are several hundred feet from the protected area boundary, which

will result in a substantial reduction in the dose rate due to distance from the

source of the radiation.

Human Health It is expected that the dose rates in these two constructions areas will be at or

very near background levels.

Human Health

|

The doses they receive from background radiation will be more significant than

| nitrogen-16 skyshine doses.

Human Health Implementation of a radiation environmental monitoring program for the new

facility, compliance with requirements for maintaining dose ALARA, and

attention to design of plant shielding to ensure dose is ALARA, will result in

doses to the public and to construction workers due to direct radiation being

minim al.
Appendix J

April 2006 J-13 NUREG-1817

Table J-2. Key Assumptions Used by the NRC Staff in Assessing Environmental Impacts at the |

Grand Gulf Early Site Permit Site

|

Technical Area Assumption EIS Section |

Land Use The Grand Gulf ESP site will be wholly contained within |

the Grand Gulf site. |

5.1 |

Land Use The construction footprint will align with environmental |

report Figure 2.1-2.

5.1.1 |

|

Land Use Land-use impacts of any potential transm ission line right- |

of-way upgrade or expansion request will be assessed by |

the appropriate authority. State or local agency citing |

procedures will be followed once right-of-way routing is

determined.

5.1.2 |

|

|

|

|

Land Use Existing transm ission line rights-of-way are 61 m (200 ft) |

in width.

5.1.2

|

Land Use Transmission line upgrades would utilize only existing |

500-kV transmission lines and rights-of-way. The 115-kV |

line is used to supply power to the site from offsite.

5.1.2 |

|

|

Land Use No significant agriculture, crops, or dairy production are or |

will be located at or immediately near the Grand Gulf site.

5.1.1

|

Land Use No third-party mining activities would be possible at the |

ESP site.

5.1.1

|

Land Use |

|

|

Planned maintenance and refueling outages would be |

staggered such that only the GGNS Unit 1 or the proposed

Grand Gulf ESP facility would be in outage at one time.

5.1.1 |

|

Land Use Salt drift from any cooling tower design would be localized |

and well below NRC guidance thresholds.

5.1.1

Land Use Induced housing effects of construction and operations |

would be dispersed across urbanized areas of

southwestern and central Mississippi.

5.1.1

Land Use |

|

|

|

The applicant would follow best managem ent practices |

and would abide by all relevant regulations pertaining to

ground-disturbing activities, such as forest and wetlands

protection.

5.1.1, 5.1.2
Appendix J

NUREG-1817 J-14 April 2006

Table J-2. (contd)

| Technical Area Assumption EIS Section

| Meteorology and Air

| Quality

|

Meteorological data from the site presented in various

tables in the environmental report and request for

additional information responses are reasonably

representative of the site (except for wind data). Only the

wind data for 2001 to 2003 are assumed to be

representative.

| 2.3.3

| Meteorology and Air

| Quality

Air emissions from the Grand Gulf ESP facility would be

bounded by those listed in the environm ental report .

| 5.2.2

| Meteorology and Air

Quality

The applicant would use dust control measures during

construction and operation.

4.2.1

| Meteorology and Air

| Quality

If air quality impacts related to transportation occur during

construction, the applicant would implement best

managem ent practices to minimize the impacts.

4.2.2

| Meteorology and Air

| Quality

Various measures outlined in the environm ental report 

would be followed to limit air quality impacts of

construction.

| 4.2.1

| Meteorology and Air

Quality

Cooling towers would have drift eliminators that are

comparable in effectiveness to the drift elim inators in

current generation cooling towers.

5.2.1

| Ecology

|

|

|

Upland and bottomland areas of the proposed Grand Gulf

ESP site that would be disturbed by construction would

undergo a botanical survey prior to initiating such

activities.

| 4.4.1.4

|

|

| Ecology

|

|

|

|

|

A recent description will be provided of the aquatic biota

that are in the vicinity of the ESP site and the transmission

line rights-of-way prior to or during the CP or COL stage. 

The description will be consistent with NUREG 1555,

Environmental Standard Review Plan, Chapter 2.4.2.

| 4.4.2, 4.4.3,1,

| 5.4.2, 5.4.3.1, 7.4

|

|

|

| Ecology

|

|

The proposed intake system will have screens with a size

such that the average intake velocity through the screen

would be less than or equal to 0.15 m/s (0.5 ft/s).

| 5.4.2.1

|

|
Appendix J

Table J-2. (contd)

| Technical Area Assumption EIS Section

April 2006 J-15 NUREG-1817

Socioeconomics Per the discussion in the environmental report, the staff |

assumed that 50 percent of the workforce at the Grand

Gulf ESP site would come from the 80-km zone

surrounding the plant, with almost all imm igrating

personnel and families living in Vicksburg, suburban

Jackson, and Natchez. The staff also did the impact

analysis under the alternative assumption that personnel

and families would be distributed the same as the current

plant-related population for GGNS.

4.5.2, 4.5.3.1, |

4.5.4.3, 4.5.4.4,

5.5.2, 5.5.3.1,

5.5.4.1, 5.5.4.3,

5.5.4.4, 5.5.4.5

Socioeconomics For the Grand Gulf ESP site, the staff identified two ways |

in which a new nuclear plant might be treated for property

tax purposes under Mississippi tax law, which was |

assumed to remain the same in the future. If the plant |

were a m erchant plant, it might be taxed as an ordinary

taxable business asset, taxable by Claiborne County. The

other possibility is that the state of Mississippi might

decide to tax the asset instead, and provide some share of

the funds back to the county and to the city of Port Gibson. 

The staff did the analysis both ways.

4.5.3.2, 5.5.3.2,

2.8.2.3

|

|

Socioeconomics |

|

|

|

|

The staff relied on SERI’s statement in a reply to a request

for additional information that it had no plans to restore the

former rail spur to the Grand Gulf ESP site. This implies

that large items and bulk materials would come in by

barge or truck. SERI also said that a rail spur could not be

precluded.

2.2.1, 2.8.2.2, |

4.5.4.1

Socioeconomics The staff assum ed that if very large groups of families with |

school-age children moved into Claiborne County, the

state of Mississippi would provide some impact assistance

to the local school system.

4.5.4.5, 5.5.4.5 |

|

|

|

Environmental |

Justice |

|

|

There are no unidentified and significant pre-existing

health conditions or resource dependencies among

minority and low-income populations in the region of the

Grand Gulf ESP site.

4.7, 5.7

Environmental |

Justice |

|

|

The relative geographical locations of concentrations of

minority and low-income individuals in the region of the

Grand Gulf ESP site as shown in the 2000 U.S. Census

are valid at time of CP or COL application.

4.7, 5.7
Appendix J

Table J-2. (contd)

| Technical Area Assumption EIS Section

NUREG-1817 J-16 April 2006

Cultural Resources

|

Cultural resource surveys will be conducted if areas

identified in Figure 4-1 in the EIS are selected for

construction.

4.6

| Cultural Resources Appropriate cultural resource surveys would be conducted

prior to construction of new transmission lines.

4.6

Cultural Resources Cultural resource-specific written directions will be

included in SERI’s Excavation and Backfill W ork

Procedures prior to construction and operation.

4.6, 5.6

| Human Health New transmission lines would be built to current industry

and regulatory standards.

5.8.3

| Human Health

|

|

Appropriate State and local requirements would be

considered when assessing the occupational hazard and

health risks associated with construction.

| 4.8.1

| Human Health

|

|

|

The staff assumed adherence to NRC, Occupational

Safety and Health Administration, and State safety

standards, practices, and procedures for operation of new

nuclear units.

| 5.8.5

| Human Health New unit or units are constructed at the location identified

in the ER.

4.9

| Human Health Assumptions listed on pages 6-41 and 6-42. 6.2.4

| Accidents Population growth in the vicinity of the site would not alter

the population distribution in the region.

5.10.2